ML031920234
| ML031920234 | |
| Person / Time | |
|---|---|
| Site: | Susquehanna  |
| Issue date: | 06/24/2003 |
| From: | Susquehanna |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| EP-PS-111, Rev 3 | |
| Download: ML031920234 (61) | |
Text
Jun. 24, 2003 Page 1 of 1 MANUAL HARD COPY DISTRIBUTION DOCUMENT TRANSMITTAL 2003-29902 U NFORMATIOI/
N e: SE M EMPL#:28401 CA#:0363 h 54-3 TRANSMITTAL INFORMATION:
TO: .-- -- 06/24/2003 LOCATION: DOCUMENT CONTROL DESK FROM: NUCLEAR RECORDS DOCUMENT CONTROL CENTER (NUCSA-2)
THE FOLLOWING CHANGES HAVE OCCURRED TO THE HARDCOPY OR ELECTRONIC MANUAL ASSIGNED TO YOU:
111 - 111 - TSC LEAD ENGINEER REMOVE MANUAL TABLE OF CONTENTS DATE: 04/16/2003 ADD MANUAL TABLE OF CONTENTS DATE: 06/23/2003 CATEGORY: PROCEDURES TYPE: EP ID: EP-PS-111 ADD: PCAF 2003-1449 REV: N/A UPDATES FOR HARD COPY MANUALS WILL BE DISTRIBUTED WITHIN 5 DAYS IN ACCORDANCE WITH DEPARTMENT PROCEDURES. PLEASE MAKE ALL CHANGES AND ACKNOWLEDGE COMPLETE IN YOUR NIMS INBOX UPON RECEIPT OF HARD COPY. FOR ELECTRONIC MANUAL USERS, ELECTRONICALLY REVIEW THE APPROPRIATE DOCUMENTS AND ACKNOWLEDGE COMPLETE IN YOUR NIMS INBOX.
4A&O5
),. PROCEDURE CHANGE PROCESS FORM
- 1. PCAF NO. - 2. PAGE 1 OF 3 3. PROC. NO. EP-PS-1 11 REV. 3
- 4. FORMS REVISED - R_, - _ R_, - R , - -R__ - R_, -- R
- 6. REQUESTED CHANGE PERIODIC REVIEW E NO 3 YES INCORPORATE PCAFS 3 NO E YES # # _ _
REVISION ] PCAF DELETION [ (CHECK ONE ONLY)
- 7.
SUMMARY
OF / REASON FOR CHANGE Revised Expiration date. No changes required. '
'~~ k*A~~ ~~~~~~
Continued i
- 8. DETERMINE COMMITTEE REVIEW REQUIREMENTS (Refer to Section 6.1.4)
PORC REVIEW REQ'D? , O E3 YES 9. PORC MTG#
BLOCKS 1I THRU 6 ARE ON PAGE 2 OFFORM
- 17. T.C. Dalpiaz ! 3227 1 06/16/2003 18. COMMUNICATION OF CHANGE REQUIRED?
PREPARER ETN DATE NO E YES (TYPE)
(Print or Type) ,-
/ [1.wZ SIGNATURE iA t' - ATTESTS THAT RESPONSIBLE SUPERViSOR HAS 19.* 0 t/ 5 a CONDUCTED QADR AND TECHNICAL REVIEW UNLESS OTHERWISE DOCUMENTED INBLOCK 16 OR ATTACHED REVIEW FORMS.
PESPON'IBLE ~~~~1BLESUPER~~~ISOW SUPERVISOR j
/DATE 'BY CROSS DISCIPLINE SIGNATURE INBLOCK 16 (IF REVIEW ATTACHED HAS ORREQUIRED) BEENFORMS.
REVIEW COMPLETED
- 20. (w 6hIr7f03
( OMAPROVAL DATE
- 21. RESPONSIBLE APPROVER ENTER N/A IF FUM HAS APPROVAL AUTHORITY INITIALS DATE JUN 182 3 FORM NDAP-QA-0002-8, Rev. 8, Page 1 of 2 (Electronic Form)
DOCUMENT CONTROL SERVICES - SSES
PROCEDURE CHANGE PROCESS FORM
- 1. PCAF NO.,n 3-tq# 2. PAGE 2 OF 3 3. PROC. NO. EP-PS-111 REV. 3
- 11. This question documents the outcome of the 50.59 and 72.48 Review required by NDAP-QA-0726. Either 1a, b, c or d must be checked YES" and the appropriate form attached or referenced.
- a. This change is an Administrative Correction for which 50.59 and 72.48 are not [ YES 3 N/A applicable.
- b. This change is a change to any surveillance, maintenance or administrative s YES a N/A procedure for which 50.59 and 72.48 are not applicable.
- c. This change is bounded by a 50.59/72.48 Screen/Evaluation, therefore, no new ] YES 3 N/A 50.59172.48 Evaluation is required.
Screen/Evaluation No.
- d. 50.59 and/or 72.48 are applicable to this change and a 50.59/72.48 f] YES 0 NIA Screen/Evaluation is attached. 1
- 13. Should this change be reviewed for potential effects on Training Needs or Material? YES 3 NO If YES, enter an Action Item @ NIMS/Action/Gen Work Mech/PICN
- 14. Is a Surveillance Procedure Review Checklist required per NDAP-QA-0722? [ YES 0 NO
- 15. Is a Special, Infrequent or Complex Test/Evolution Analysis Form required per a YES 0 NO NDAP-QA-0320? (SICT/E form does not need to be attached.)
- 16. Reviews may be documented below or by attaching Document Review Forms NDAP-QA-0101-1.
REVIEWED BY WITH DATE REVIEW NO COMMENTS QADR TECHNICAL REVIEW REACTOR ENGINEERINGINUCLEAR FUELS
- IST **
OPERATIONS NUCLEAR SYSTEMS ENGINEERING NUCLEAR MODIFICATIONS MAINTENANCE HEALTH PHYSICS NUCLEAR TECHNOLOGY CHEMISTRY OTHER 10 CFR 50.540 J r IO3 g Z4
- Required for changes that affect, or have potential for affecting core reactivity, nuclear fuel, core power level indication or impact the thermal power heat balance. ()
Required for changes to Section Xi Inservice Test Acceptance Criteria.
FORM NDAP-QA-0002-8, Rev. 8, Page 2 of 2 (Electronic Form)
. A, due0 as~
,@-b 2 X a _.r PROCEDURE COVER SHEET *
- 1 PPL SUSQUEHANNA, LLC NUCLEAR DEPARTMENT PROCEDURE TSC Lead Engineer:Emergency Plan Position Specific Instruction EP-PS-1 11 Revision 3 Page 1 of 3 QUALITY CLASSIFICATION: APPROVAL CLASSIFICATION:
El QA Program 0 Non-QA Program E Plant E] Non-Plant 0 Instruction EFFECTIVE DATE:
PERIODIC REVIEW FREQUENCY: 2 years PERIODIC REVIEW DUE DATE: 6vY 02 tz°5 RECOMMENDED REVIEWS:
All Procedure Owner: Nuclear Emergency Planning Responsible Supervisor. Manager-Station Engineering Responsible FUM: Supervisor Nuclear Emergency Planning Responsible Approver VP-Nuclear Operations FORM NDAP-QA-0002-1, Rev. 3, Page 1 of 1
( I Tab 2 EP-PS-1 11-2 EMERGENCY ORGANIZATION CONTROL ROOM EMERGENCY DIRECTOR (SHIFT MANAGER) ll lf l l EMERGENCY PLAN SHIFT TECHNICAL NRC OPERATORS COMMUNICATOR ADVISOR (STA) COMMUNICATOR(S)
EP-AD-000-406, Revision 16, Page 1 of 3
( (
Tab 2 EP-PS-1 11-2 TSC ORGANIZATION EMERGENCY DIRECTOR*
Proposed NRC Rev. 5.02 Susquehanna Steam Electric Station Units I and 2 Support Personnel (9)
Emergency Plan Ref. Sec. 6.2.12 TSC ORGANIZATION 9 report 0 60 minutes FIGURE 6.2 Support Personnel Ref. Sec. 6.2.12 report 0 90 minutes Designates minimum requirements In accordance with Table 6.1 for 60-minute response.
Individuals may be located in the OSC, TSC, or Field.
- Designates positions required for TSC activation.
EP-AD-000-406, Revision 16, Page 2 of 3
(
Tab 2 EP-PS-1 11-2 EOF ORGANIZATION Proposed NRC Rev. 5-02 Susquehanna Steam Electric Station Units 1 and 2 Emergency Plan EOF ORGANIZATION Designates minimum requirements In accordance with Table 6.1 for 90 minute response.
FIGURE 6.3
- Designates positions required for EOF activation.
EP-AD-000-406, Revision 16, Page 3 of 3
Tab 6 EP-PS-1 11-6 EMERGENCY CLASSIFICATION CHECK 0 1.0 TIMING OF CLASSIFICATION 01 1.1 UNUSUAL EVENT An UNUSUAL EVENT shall be declared within 15 minutes of having information necessary to make a declaration.
03 1.2 ALERT An ALERT shall be declared within 15 minutes of having Iniformation necessary to make a declaration.
0: 1.3 SITE AREA EMERGENCY A SITE AREA EMERGENCY shall be declared within 15 minutes of having information necessary to make a declaration.
0: 1.4 GENERAL EMERGENCY A GENERAL EMERGENCY shall be declared within 15 minutes of having information necessary to make a declaration.
EP-AD-000-200, Revision 19, Page 1 of 38
Tab 6 EP-PS-1 11-6 CLASSIFICATION OF EMERGENCY CONDITIONS USE OF EMERGENCY CLASSIFICATION MATRIX NOTE: CONFIRM THAT INDICATORS AND/OR ALARMS REFLECT ACTUAL CONDITIONS PRIOR TO TAKING ACTION BASED ON THE INDICATOR OR ALARM.
The matrix is worded in a manner that assumes parameter values indicated are the actual conditions present in the plant.
The matrix is designed to make it possible to precisely classify an abnormal occurrence into the proper emergency classification based on detailed Emergency Action Level (EAL) descriptions.
It is impossible to anticipate every abnormal occurrence. Therefore, before classifying any abnormal occurrence based on the EALs in the matrix, one should verify that the general conditions prevalent in-plant and offsite meet the general class description of the emergency classification. In addition, prior to classification, one should be aware of the ramifications in-plant and particularly offsite of that classification. Special consideration of offsite consequences should be made prior to declaring a GENERAL EMERGENCY.
EP-AD-000-200, Revision 19, Page 2 of 38
Tab 6 EP-PS-1 11-6 CLASS DESCRIPTIONS UNUSUAL EVENT - Events that are occurring or have occurred which indicate a potential degradation of the level of safety of the plant.
No releases of radioactive material requiring offsite response or monitoring are expected unless further degradation of safety systems occurs.
ALERT - Events that are occurring or have occurred which involve an actual or potential substantial degradation of the level of safety of the plant. Any releases are expected to be limited to small fractions of the EPA Protective Action Guideline exposure levels.
SITE AREA EMERGENCY - Events that are occurring or have occurred which involve actual or imminent major failures of plant functions needed for protection of the public. Any releases are not expected to exceed EPA Protective Action Guideline exposure levels except inside the emergency planning boundary.
GENERAL EMERGENCY - Events that are occurring or have occurred which involve actual or imminent substantial core degradation or melting with potential for loss of containment integrity. Expectation is that releases will exceed EPA Protective Action Guideline exposure levels beyond the emergency planning boundary.
EP-AD-00O-200, Revision 19, Page 3 of 38
Tab 6 EP-PS-1 11-6 CATEGORY INDEX TO THE MATRIX FOR THE CLASSIFICATION OF EMERGENCY CONDITIONS TABLE OF CONTENTS CATEGORY EVENT PAGE I AIRCRAFTITRAIN ACTIVITY ............................................. 5 2 CONTROL ROOM EVACUATION ............... .............................. 6 3 FUEL CLADDING DEGRADATION ................ ............................. 7 4 GENERAL ............................................. 10 5 INJURED/CONTAMINATED PERSONNEL ............................................. 11 6 IN-PLANT HIGH RADIATION ............................................. 12 7 LOSS OF AC POWER ............................................. 13 8 LOSS OF CONTROL ROOM ALARMS AND ANNUNCIATORS ................... 14 9 LOSS OF DC POWER ............................................. 15 10 LOSS OF DECAY HEAT REMOVAL CAPABILITY .................................... ;.16 11 LOSS OF REACTIVITY CONTROL ................................... 17 12 LOSS OF REACTOR VESSEL INVENTORY ................................... 19 13 NATURAL PHENOMENA ................................... 21 14 ONSITE FIRE/EXPLOSION ................................... 23 15 RADIOLOGICAL EFFLUENT ................................... 25 16 SECURITY EVENT ............ 29 17 SPENT FUEL RELATED INCIDENT ................................ 31 18 STEAM LINE BREAK ................................ 33 19 TOXIC/FLAMMABLE GASES ................................ 36 20 TECHNICAL SPECIFICATION SAFETY LIMIT................................. 37 21 DRY FUEL STORAGE ................................ 38 EP-AD-000-200, Revision 19, Page 4 of 38
Tab 6 EP-PS-1 11-6 I - AIRCRAFT/TRAIN ACTIVITY UNUSUAL EVENT EAL# 1.1 Aircraft crash or train derailment onsite as indicated by:
Visual observation or notification received by control room operator.
ALERT EAL# 1.2 Aircraft or missile strikes a station structure as indicated by:
Direct observation or notification received by control room operator.
SITE AREA EMERGENCY EAL# 1.3 Severe damage to safe shutdown equipment from aircraft crash or missile impact when not in cold shutdown, determined by:
(A and B and C)
A. Direct observation or notification received by control room operator.
and B. Shift Supervisor evaluation.
and C. Reactor Coolant temperature greater than 200OF as indicated on Panel 1C651 (2C651).
GENERAL EMERGENCY EAL# 1.4 None.
EP-AD-000-200, Revision 19, Page 5 of 38
Tab 6 EP-PS-1 11-6 2- CONTROL ROOM EVACUATION UNUSUAL EVENT EAL# 2.1 None.
ALERT EAL# 2.2 Control Room evacuation as indicated by:
(A and B)
A. Initiation of control room evacuation procedures.
and B. Establishment of control of shutdown systems from local stations.
SITE AREA EMERGENCY EAL# 2.3 Delayed Control Room Evacuation as indicated by:
(A and B)
A. Initiation of control room evacuation procedures.
and B. Shutdown systems control at local stations not established within 15 minutes.
GENERAL EMERGENCY EAL# 2.4 None.
EP-AD-000-200, Revision 19, Page 6 of 38
Tab 6 EP-PS-1 11-6 3 - FUEL CLADDING DEGRADATION UNUSUAL EVENT EAL# 3.1 Core degradation as indicated by:
(A or B)
A. Valid Off-gas Pre-treatment Monitor high radiation alarm annunciation on Panel 1C651 (2C651) or indication on Panel I C600 (2C600).
or B. Reactor coolant activity, determined by sample analysis greater than or equal to 2 iCi/cc of 1-131 equivalent.
ALERT EAL# 3.2 Severe fuel cladding degradation as indicated by:
(A or B or C or D)
A. Valid Off-gas Pre-treatment monitor High-High radiation alarm annunciation on Panel 1C651 (2C651) or indication on Panel 1C600 (2C600).
or B. Valid Reactor coolant activity greater than 300 iACi/cc of equivalent 1-131, as determined by sample analysis.
or C. Valid Main Steam Line High radiation trip annunciation or indication on Panel 1C651 (2C651).
or D. Valid containment post accident monitor indication on Panel 1C601 (2C601) greater than 200 R/hr. (An 8R/hr correction factor must be added manually to the indication to offset a downscale error if primary containment temperature exceeds 225 degrees Fahrenheit. Reference EC-079-0521.)
(CONTINUED ON NEXT PAGE)
EP-AD-000-200, Revision 19, Page 7 of 38
Tab 6 EP-PS-1 11-6 3 - FUEL CLADDING DEGRADATION (continued)
SITE AREA EMERGENCY EAL# 3.3 Severely degraded core as indicated by:
(A or B)
A. Reactor coolant activity greater than 1,000 gCi/cc of equivalent 1-131 as determined by sample analysis.
or B. Valid containment post accident monitor indication on Panel C601 (2C601) greater than 400 R/hr. (An 8 R/hr correction factor must be added manually to the indication to offset a downscale error if primary containment temperature exceeds 225 degrees Fahrenheit. Reference EC-079-0521.)
(CONTINUED ON NEXT PAGE)
EP-AD-000-200, Revision 19, Page 8 of 38
Tab 6 EP-PS-1 11-6 3 - FUEL CLADDING DEGRADATION (continued)
GENERAL EMERGENCY EAL# 3.4.a Fuel cladding degradation. Loss of 2 out of 3 fission product barriers (fuel cladding and reactor coolant pressure boundary) with potential loss of the third barrier (primary containment) as indicated by:
(A or B)
A. (1 and 2)
- 1. Valid containment post accident monitor indication on Panel C601 (2C601) greater than 400 R/hr. (An 8 R/hr correction factor must be added manually to the indication to offset a downscale error if primary containment temperature exceeds 225 degrees Fahrenheit. Reference EC-079-0521.)
and
- 2. (a or b or c)
- a. Containment pressure greater than 40.4 PSIG, indicated on Panel 1C601 (2C601).
or
- b. A visual inspection of the containment indicates a potential for loss of containment (e.g. anchorage or penetration failure, a crack in containment concrete at tendon).
or
- c. Other indications of potential or actual loss of primary containment. -
or B. (1 and 2)
- 1. Reactor coolant activity greater than 1,000 ,uCicc of- equivalent 1-131 as determined by sample analysis.
and
- 2. Actual or potential failure of reactor coolant isolation valves to isolate a coolant leak outside containment as determined by valve position indication on Panel 1C601 (2C601) or visual inspection.
OR EAL# 3.4.b Core melt as indicated by:
(A and B)
A. Valid containment post accident monitor indication on Panel 1C601 (2C601) greater than 2000 Rhr. (An 8 R/hr correction factor must be added manually to the indication to offset a downscale error if primary containment temperature exceeds 225 degrees Fahrenheit. Reference EC-079-0521.)
and B. Containment high pressure indication or annunciation on Panel 1C601 (2C601).
EP-AD-000-200, Revision 19, Page 9 of 38
Tab 6 EP-PS-1 11-6 4-GENERAL UNUSUAL EVENT EAL# 4.1 Plant conditions exist that warrant increased awareness on the part of plant operating staff or state and/or local offsite authorities as indicated by.
Events that are occurring or have occurred which indicate a potential degradation of the level of safety of the plant. No releases of radioactive material requiring offsite response or monitoring are expected unless further degradation of safety systems occurs.
ALERT EAL# 4.2 Other plant conditions exist that warrant precautionary activation of PPL, State, County, and local emergency centers as indicated by:
Events that are occurring or have occurred which involve an actual or potential substantial degradation of the level of safety of the plant. Any releases are expected to be limited to small fractions of the EPA Protective Action Guideline exposure levels.
SITE AREA EMERGENCY EAL# 4.3 Other plant conditions exist that warrant activation of emergency centers and monitoring teams or a precautionary notification to the public near the site as indicated by:
Events that are occurring or have occurred which involve actual or imminent major failures of plant functions needed for protection of the public. Any releases are not expected to exceed EPA Protective Action Guideline exposure levels except inside the emergency planning boundary.
GENERAL EMERGENCY EAL# 4.4 Other plant conditions exist, from whatever, source, that make release of large amounts of radioactivity in a short time period available as indicated by:
Events that are occurring or have occurred which involve actual or imminent substantial core degradation or melting with potential for loss of containment integrity. Expectation is that releases will exceed EPA Protective Action Guideline exposure levels beyond the emergency planning boundary.
EP-AD-000-200, Revision 19, Page 10 of 38
Tab 6 EP-PS-1 11-6 5.- INJUREDICONTAMINATED PERSONNEL UNUSUAL EVENT EAL# 5.1 Transportation of externally contaminated injured individual from site to offsite medical facility as deemed appropriate by Shift Supervisor.
ALERT EAL# 5.2 None.
SITE AREA EMERGENCY EAL# 5.3 None.
GENERAL EMERGENCY EAL# 5.4 None.
EP-AD-000-200, Revision 19, Page 11 of 38
Tab 6 EP-PS-1 11-6 6 - IN-PLANT HIGH RADIATION UNUSUAL EVENT EAL# 6.1 Unanticipated or unplanned concentrations of airborne activity exist in normally accessible areas, which are not due to planned maintenance activities, as indicated by:
Concentrations exceed 500 times the DAC values of IOCFR20 Appendix B, Table I values for a single isotope, or for multiple isotopes where CA + C + CC CN > 500 DACA DAC. DACC DACH ALERT EAL# 6.2 Unexpected in-plant high radiation levels or airborne contamination which indicates a severe degradation in the control of radioactive material as indicated by:
Area Radiation Monitor reading 1000 times normal annunciation on Panel 1C601 (2C601) or indication on Panel I C600 (2C600).
SITE AREA EMERGENCY EAL# 6.3 None.
GENERAL EMERGENCY EAL# 6.4 None.
EP-AD-000-200, Revision 19, Page 12 of 38
Tab 6 EP-PS-1 11-6 7 - LOSS OF AC POWER UNUSUAL EVENT EAL# 7.1 Loss of offsite power or loss of all onsiteAC power supplies as indicated by:
(A or B)
A. Loss of power to Startup Transformer 10 and 20 annunciation or indication on Panel OC653.
or B. Failure of all diesel generators to start or synchronize to the emergency buses by indication or annunciation on Panel OC653.
ALERT EAL# 7.2 Loss of all offsite power and all onsite AC power supplies as indicated by:
(A and B)
A. Loss of power to Startup Transformer 10 and 20 annunciation or indication on Panel OC653.
and B. Failure of all diesel generators to start or synchronize to the emergency buses by
-annunciation or indication on Panel OC653.
SITE AREA EMERGENCY EAL# 7.3 Loss of all offsite power and loss of all onsite AC power supplies for greater than 15 minutes as indicated by:
(A and B and C)
A. Loss of offsite power.
and B. Failure of all diesel generators to startup or synchronize to the emergency buses by indication or annunciation on OC653.
and C. The above conditions exist for greater than 15 minutes.
GENERAL EMERGENCY EAL# 7.4 None.
EP-AD-000-200, Revision 19, Page 13 of 38
Tab 6 EP-PS-1 11-6 8 - LOSS OF CONTROL ROOM ALARMS AND ANNUNCIATORS UNUSUAL EVENT EAL# 8.1 None.
ALERT EAL# 8.2 Loss of all control room annunciators as indicated by:
In the opinion of the Shift Supervisor, all Control Room annunciators and the Plant Process Computer are lost, or insufficient annunciators are available to safely operate the unit(s) without supplemental observation of plant systems.
SITE AREA EMERGENCY EAL# 8.3 All annunciators lost and plant transient initiated while annunciators are lost as indicated by:
(A and B)
A. In the opinion of the Shift Supervisor, all Control Room annunciators and the Plant Process Computer are lost, or insufficient annunciators are available to safely operate the unit(s) without supplemental observation of plant systems.
and B. (1or2or3or4)
- 1. Low-Low reactor water level indication on Panel C651-(2C651) followed by ECCS initiation on Panel 1C601 (2C601).
or
- 2. Reactor coolant temperature change greater than 1000 F per hour indication on recorder TR-1 R006 on Panel 1C007 (2C007) (Reactor Building elevation 683').
or
- 3. High reactor pressure indication on Panel 1C651 (2C651) and followed by scram indication on Panel 1C651 (2C651).
or
- 4. Any indication that transient has occurred or is in progress.
GENERAL EMERGENCY EAL# 8.4 None.
EP-AD-000-200, Revision 19, Page 14 of 38
Tab 6 EP-PS-1 11-6 9 - LOSS OF DC POWER UNUSUAL EVENT EAL# 9.1 None.
ALERT EAL# 9.2 Loss of onsite vital DC power as indicated by-(A and B)
A. Less than 210 volts on the 250 VDC main distribution Panel buses, D652 (2D652) and 1D662 (2D662) as indicated by trouble alarms on Panel 1C651 (2C651).
and B. Less than 105 volts on the 125 VDC main distribution buses 1D612 (2D612), D622 (2D622), D632 (2D632), and 1D642 (2D642) as indicated by trouble alarms on Panel 1C651 (2C651).
NOTE: Buses are not tripped on undervoltage condition.
SITE AREA EMERGENCY EAL# 9.3 Loss of all vital onsite DC power sustained for greater than 15 minutes as indicated by:
(A and B and C)
A. Less than 210 volts on the 250 VDC main distribution Panel buses, I D652 (2D652) and 1D662 (2D662) as indicated by trouble alarms on Panel 1C651 (2C651).
and B. Less than 105 volts on the 125 VDC main distribution buses 1D612 (2D612), 1D622 (2D622), D632 (2D632), and 1D642 (2D642) as indicated by trouble alarms on Panel 1C651 (2C651).
and C. The above condition exists for greater than 15 minutes.
NOTE: Buses are not tripped on undervoltage condition.
GENERAL EMERGENCY EAL# 9.4 None.
EP-AD-000-200, Revision 19, Page 15 of 38
Tab 6 EP-PS-1 11-6 10 - LOSS OF DECAY HEAT REMOVAL CAPABILITY UNUSUAL EVENT EAL# 10.1 None.
ALERT EAL# 10.2 Inability to remove decay heat while in plant condition 4, inability to maintain the plant in cold shutdown as indicated by:
Inability to maintain reactor coolant temperature less than 2000 F with the reactor mode switch in shutdown; exception is when testing per Special Test Exception TS 3.10.1 which allows maximum temperature of 212 0F.
SITE AREA EMERGENCY EAL# 10.3 Inability to remove decay heat while the plant is shutdown as indicated by:
(A and B and C)
A. Reactor Mode switch in shutdown.
and B. Reactor Coolant System temperature greater than 200OF and rising.
and C. Suppression Pool temperature greater than 120OF and rising.
GENERAL EMERGENCY EAL# 10.4 Inability to remove decay heat while the plant is shutdown with possible release of large amounts of radioactivity as indicated by:
(A and B and C)
A. Reactor mode switch in shutdown.
and B. Reactor coolant system temperature greater than 200OF and rising.
and C. Suppression pool temperature greater than 2900 F indicated on the computer output (MAT 12,13,14,15 or 16).
EP-AD-000-200, Revision 19, Page 16 of 38
Tab 6 EP-PS-1 11-6 11 - LOSS OF REACTIVITY CONTROL UNUSUAL EVENT EAL# 11.1 Inadvertent Criticality as indicated by:
Unexpected increasing neutron flux indication on Panel 1C651 (2C651).
ALERT EAL# 11.2 Failure of the Reactor Protection System or the Alternate Rod nsertion System to initiate and complete a scram that brings the reactor subcritical as indicated by:
(A or B) and (C and D and E)
A. Trip of at least one sub-channel in each trip system (RPS A and RPS B) as indicated by annunciators and trip status lights on Panel 1C651(2C651).
or B. Trip of both trip systems (ARI A and ARI B) as indicated by annunciators on Panel 1C601 (2C601).
and C. Failure of control rods to insert, confirmed by the full core display indication on Panel 1C651(20651) or process computer indications.
and D. Failure to bring the reactor subcritical confirmed by neutron count rate on the neutron monitoring indication on Panel 1C651 (2C651).
and E. Reactor power >5% as indicated on Panel 1C651(2C651).
(CONTINUED ON NEXT PAGE)
EP-AD-000-200, Revision 19, Page 17 of 38
Tab 6 EP-PS-1 11-6 11 - LOSS OF REACTIVITY CONTROL (continued)
SITE AREA EMERGENCY EAL# 11.3 Loss of functions needed to bring the reactor subcritical and loss of ability to bring the reactor to cold shutdown as indicated by:
(A and B and C and D)
A. Inability to insert sufficient control rods to bring the reactor subcritical as indicated by count rate on the neutron monitoring instrumentation on Panel 1C651 (2C651).
and B. (1 or2)
Failure of both loops of standby liquid control to inject into the vessel indicated by:
- 1. Low pump discharge pressure indication on Panel 1C601_C601).
or
- 2. Low flow indication on Panel 1C601 (2C601).
and C. Reactor coolant temperature greater than 2000F, indicated on Panel 1C651 (2C651).
and D. Reactor power >5% indicated on Panel 1C651 (2C651).
GENERAL EMERGENCY EAL# 11.4 Loss of functions needed to bring the reactor subcritical and transient in progress that makes release of large amounts of radioactivity in a short period possible as indicated by:
(A or B)and (C and D)
A. Trip of at least one sub-channel in each trip system (RPS A and RPS B), indicated by annunciation or trip status lights on Panel 1C651 (2C651).
or B. Trip of both systems (ARI A and ARI B) as indicated by annunciators on Panel 1C601 (2C601).
and C. Loss of SLC system capability to inject, indicated by instrumentation on Panel 1C601 (2C601).
and D. Reactor power greater than 25% of rated, indicated on Panel 1C651 (2C651).
EP-AD-000-200, Revision 19, Page 18 of 38
Tab 6 EP-PS-1 11-6 12 - LOSS OF REACTOR VESSEL INVENTORY UNUSUAL EVENT EAL# 12.1 Valid initiation of an Emergency Core Cooling System (ECCS) System as indicated by:
(A or B)
A. Initiation of an ECCS System and low, low, low reactor water level (-129) annunciation or indication on Panel 1C651 (2C651).
or B. Initiation of an ECCS System and High Drywell Pressure annunciation or indication on Panel 1C601 (2C601).
ALERT EAL# 12.2 Reactor coolant system leak rate greater than 50 gpm as indicated by:
(A or B)
A. Drywell floor drain sump A or B Hi-Hi alarm on Panel 1C601 (2C601) and 2 or more drywell floor drain pumps continuously running as indicated on Panel 1C601 (2C601).
or B. Other estimates of Reactor coolant system leakage indicating greater than 50 gpm.
SITE AREA EMERGENCY EAL# 12.3 Known loss of coolant accident greater than make-up capacity as indicated by:
Water level below (and failure to return to) top of active fuel for greater than three minutes as indicated on fuel zone level indicator on Panel 1C601 (2C601).
(CONTINUED ON NEXT PAGE)
EP-AD-000-200, Revision 19, Page 19 of 38
Tab 6 EP-PS-1 11-6 12 - LOSS OF REACTOR VESSEL INVENTORY (continued)
GENERAL EMERGENCY EAL# 12.4.a Loss of coolant accident with possibility of imminent release of large amounts of radioactivity as indicated by:
Water level below (and failure to return to) top of active fuel for greater than 20 minutes as indicated on fuel zone level indicator on Panel 1C601 (2C601).
OR EAL# 12.4.b Loss of Reactor Vessel inventory. Loss of 2 out of 3 fission product barriers (fuel cladding & reactor coolant pressure boundary) with potential loss of the third barrier (primary containment), as indicated by:
(A or B)
A. (1 and 2 and 3)
- 1. High drywell pressure annunciation or indication on Panel 1C601 (2C601).
and
- 2. (a or b or c)
- a. Containment pressure exceeds 40.4 PSIG as indicated on Panel 1C601 (2C601).
or
- b. A visual inspection of the containment indicates a potential or actual loss of containment (e.g. anchorage or penetration failure).
or
- c. Containment isolation valve(s) fail to close as indicated by valve position indication on Panel 1C601 (2C601).
and
- 3. Reactor Vessel level drops below (and fails to return to) top of active fuel for greater than three minutes as indicated on fuel zone level indicator on Panel 1C601 (2C601).
or B. (1 and 2)
- 1. Failure of reactor pressure vessel isolation valves to isolate coolant break outside containment as indicated by valve position indication on Panel 1C601 (2C601) or visual inspection.
and
- 2. Reactor vessel level drops below (and fails to return to) top of active fuel for greater than three minutes as indicated on fuel zone level indicator on Panel 1C601 (2C601).
EP-AD-000-200, Revision 19, Page 20 of 38
Tab 6 EP-PS-1 11-6 13 - NATURAL PHENOMENA UNUSUAL EVENT EAL# 13.1 Natural phenomenon occurrence as indicated by:
(A or B or C)
A. Tomado impact on site.
or B. Hurricane impact on site.
or C. Earthquake detected by seismic instrumentation systems on Panel 0C696.
ALERT EAL# 13.2 Natural Phenomenon Occurrence as indicated by:
(A or B or C)
A. Tomado with reported wind velocities greater than 200 mph impacting on site.*
or B. Reported hurricane or sustained winds greater than 70 mph.*
or C. Earthquake at greater than operating basis earthquake (OBE) levels as indicated on Panel 0C696.
- Telephone numbers for the National Weather Bureau are located in the Emergency Telephone Directory.
(CONTINUED ON NEXT PAGE)
EP-AD-000-200, Revision 19, Page 21 of 38
Tab 6 EP-PS-1 11-6 13 - NATURAL PHENOMENA (continued)
SITE AREA EMERGENCY EAL# 13.3 Severe natural phenomenon occurrence, with plant not in cold shutdown, as indicated by:
(A and 3)
A. Reactor Coolant Temperature greater than 200OF as indicated on Panel C651 (2C651).
and B. (1 or2 or3)
- 1. Reported hurricane or sustained winds greater than 80 mph.*
or
- 2. Earthquake with greater than Safe Shutdown Earthquake (SSE) levels as indicated on Panel OC696.
or
- 3. Tornado with reported wind velocities greater than 220 mph impacting on site.*
GENERAL EMERGENCY EAL# 13.4 None.
- Telephone numbers for the National Weather Bureau are located in the Emergency Telephone Directory.
EP-AD-000-200, Revision 19, Page 22 of 38
Tab 6 EP-PS-1 11-6 14 - ONSITE FIRE/EXPLOSION UNUSUAL EVENT EAL# 14.1 Significant fire within the plant as indicated by:
(A and B)
A. Activation of fire brigade by Shift Supervisor.
and B. Duration of fire longer than 15 minutes after time of notification.
OR Explosion inside security protected area, with no significant damage to station facilities, as indicated by:
Visual observation or notification received by control room operator and Shift Supervisor evaluation.
ALERT EAL# 14.2 On-site Fire/Explosion as indicated by:
(A or B)
A. Fire lasting more than 15 minutes and fire is in the vicinity of equipment required for safe shutdown of the plant and the fire is damaging or is threatening to damage the equipment due to heat, smoke, flame, or other hazard.
or B. (1 and 2)
Explosion damage to facility-affecting plant operation as determined by:
- 1. Direct observation or notification received by control room operator.
and
- 2. Shift Supervisor observation.
(CONTINUED ON NEXT PAGE)
EP-AD-000-200, Revision 19, Page 23 of 38
Tab 6 EP-PS-1 11-6 14 - ONSITE FIRE/EXPLOSION (continued)
SITE AREA EMERGENCY EAL# 14.3 Damage to safe shutdown equipment due to fire or explosion has occurred when plant is not in cold shutdown, and damage is causing or threatens malfunction of equipment required for safe shutdown of the plant as determined by:
(A and B and C)
A. Direct observation or notification received by control room operator.
and B. Shift Supervisor evaluation.
and C. Reactor Coolant Temperature greater than 200OF as indicated on Panel 1C651 (2C651).
GENERAL EMERGENCY EAL# 14.4 None.
EP-AD-00O-200, Revision 19, Page 24 of 38
Tab 6 EP-PS-1 11-6 15 - RADIOLOGICAL EFFLUENT UNUSUAL EVENT EAL# 15.1 Any unplanned release of gaseous or liquid radioactivity to the environment that exceeds 2 times the Technical Requirements Manual limits for 60 minutes or longer.
EAL# 15.1(1 or 2 or 3)
- 1. Valid Noble Gas vent stack monitor reading(s) that exceeds a total site release rate of 2.OE+6 p.Ci/min and that is sustained for 60 minutes or longer.
- 2. Confirmed sample analyses for airborne releases indicates total site release rates at the site boundary with a release duration of 60 minutes or longer resulting in dose rates of:
a) Noble gases >1000 mrenyear whole body, or b) Noble gases >6000 mremlyear skin, or c) 1-131, 1-133, H-3, and particulates with half lives >8 days >3000 mrem/year to any organ (inhalation pathways only).
- 3. Confirmed sample analyses for liquid releases indicates concentrations with a release duration of 60 minutes or longer in excess of two tirne the Technical Requirements Manual liquid effluent limits.
(CONTINUED ON NEXT PAGE)
EP-AD-000-200, Revision 19, Page 25 of 38
Tab 6 EP-PS-1 11-6 15 - RADIOLOGICAL EFFLUENT (continued)
ALERT EAL# 15.2 Any unplanned release of gaseous or liquid radioactivity to the environment that exceeds 200 times Technical Requirement Manual limits for 15 minutes or longer.
EAL# 15.2 (1 or 2 or 3)
- 1. Valid Noble Gas vent stack monitor reading(s) that exceeds a total site release rate of 2E+8 IiCi/min and that is sustained for 15 minutes or longer.
OR 2 Confirmed sample analyses for airborne releases indicates total site release rates at the site boundary for 15 minutes or longer resulting in dose rates of:
a) Noble gases >1.OE+5 mrem/year whole body, or b) Noble gases >6.OE+5 mrem/year skin, or c) 1-131, 1-133, H-3, and particulates with half-lives >8 days >3.OE+5 mremlyear to any organ (inhalation pathways only).
- 3. Confirmed sample analyses for liquid releases indicates concentrations in excess of 200 times the Technical Requirements Manual liquid effluent limfits for 15 minutes or longer.
(CONTINUED ON NEXT PAGE)
EP-AD-000-200, Revision 19, Page 26 of 38
Tab 6 EP-PS-1 11-6 15- RADIOLOGICAL EFFLUENT (continued)
SITE AREA EMERGENCY EAL# 15.3 Dose at the Emergency Plan boundary resulting from an actual or imminent release of gaseous radioactivity exceeds 100 mrem whole body TEDE or 500 mrem child thyroid CDE for the actual or projected duration of release.
EAL# 15.3(1 or2or3or4or5)
- 1. Valid Noble Gas vent stack monitor readings(s) that exceeds a total release rate 6.2E8 giCi/min for greater than 15 minutes and Dose Projections are not available.
Note: If the required dose projection cannot be completed within the 15 minute period, then the declaration must be made based on a valid sustained monitor reading(s).
- 2. Valid dose assessment using actual meteorology indicates projected doses greater than 100 mrem whole body TEDE or 500 mrem child thyroid CDE at or beyond the EPB.
- 3. A valid reading sustained for 15 minutes or longer on the RMS perimeter radiation monitoring system greater than 100 mRfhr.
- 4. Field survey results indicate Emergency Planning boundary dose rates exceeding 100 mR/hr expected to continue for more than one hour.
- 5. Analyses of field survey samples indicate child thyroid dose commitment at the Emergency Planning Boundary of 500 mrem for one hour of inhalation.
(CONTINUED ON NEXT PAGE)
EP-AD-000-200, Revision 19, Page 27 of 38
Tab 6 EP-PS-1 11 -6 15 - RADIOLOGICAL EFFLUENT (continued)
GENERAL EMERGENCY EAL# 15.4 Dose at the Emergency Planning Boundary resulting from an actual or imminent release of gaseous radioactivity exceeds 1000 mrem whole body TEDE or 5000 mrem child thyroid CDE for the actual or projected duration of the release using actual meteorology.
EAL# 15.4 (1 or 2 or 3 or 4 or 5)
- 1. Valid Noble Gas vent stack monitor readings(s) that exceed a total release rate of 6.2E9 j+/-Cimin for greater that 15 minutes and Dose Projections are not available.
Note: If the required dose projection cannot be completed within the 15 minute period, then the declaration must be made based on a valid sustained monitor reading(s).
- 2. Valid dose assessment using actual meteorology indicates projected doses greater than 1000 mrem whole body TEDE or 5000 mrem child thyroid CDE at or beyond the EPB.
- 3. A valid reading sustained for 15 minutes or longer on the RMS perimeter radiation monitoring system greater than 1000 mR/hr.
- 4. Field survey results indicate Emergency Planning Boundary dose rates exceeding 1000 mR/hr expected to continue for more than one hour.
- 5. Analyses of field survey samples indicate child thyroid dose commitment at the Emergency Planning Boundary of 5000 mrem for one hour of inhalation.
EP-AD-000-200, Revision 19, Page 28 of 38
Tab 6 EP-PS-1 11-6 16 - SECURITY EVENT UNUSUAL EVENT EAL# 16.1 Security threat or attempted entry or attempted sabotage as indicated by:
(A or B or C)
A. A report from Security of a security threat, attempted entry, or attempted sabotage of the owner controlled area adjacent to the site.
or B. Any attempted act of sabotage which is deemed legitimate in the judgment of the SHIFT SUPERVISOR/EMERGENCY DIRECTOR, and affects plant operation.
or C. A site specific credible security threat notification.
ALERT EAL# 16.2 Ongoing Security Compromise as indicated by:
(A or B)
A. A report from Security that a security compromise is at the site but no penetration of protected areas has occurred.
or B. Any act of sabotage which results in an actual or potential substantial degradation of the level of safety of the plant as judged by the SHIFT SUPERVISOR/EMERGENCY DIRECTOR.
SITE AREA EMERGENCY EAL# 16.3 An ongoing adversary event threatens imminent loss of physical control of plant as indicated by:
(A or B)
A. Report from Security that the security of the plant vital area is threatened by unauthorized (forcible) entry into the protected area.
or B. Any act of sabotage which results in actual or likely major failures of plant functions needed for protection of the public as judged by the SHIFT SUPERVISOR/EMERGENCY DIRECTOR.
(CONTINUED ON NEXT PAGE)
EP-AD-000-200, Revision 19, Page 29 of 38
Tab 6 EP-PS-1 11-6 16 - SECURITY EVENT (continued)
GENERAL EMERGENCY EAL# 16.4 Loss of physical control of facilities as indicated by:
(A or B)
A. Report from Security that a loss of physical control of plant vital areas has occurred.
or B. Any act of sabotage which results in imminent significant cladding failure or fuel melting with a potential for loss of containment integrity or the potential for release of significant amounts of radioactivity in a short time as judged by the SHIFT SUPERVISORtEMERGENCY DIRECTOR.
EP-AD-000-200, Revision 19, Page 30 of 38
Tab 6 EP-PS-1 11-6 17 - SPENT FUEL RELATED INCIDENT UNUSUAL EVENT EAL# 17.1 Unanticipated or unplanned concentrations of airborne activity exist in normally accessible areas, which is not due to planned maintenance activities, as indicated by:
Concentrations exceed 500 times the DAC values of I0CFR20 Appendix B, Table I values for a single isotope, or full multiple isotopes where CA C + CC ... C 500 DA CA DA C. DA Cc DA Cm ALERT EAL# 17.2 Unexpected in-plant high radiation levels or airborne contamination which indicates a severe fuel handling accident as indicated by:
Refuel floor area radiation monitor reading 1000 times normal annunciation on Panel I C601 (2C601) or indication on Panel 1C600 (2C600).
(CONTINUED ON NEXT PAGE)
EP-AD-000-200, Revision 19, Page 31 of 38
Tab 6 EP-PS-1 11-6 17 - SPENT FUEL RELATED INCIDENT (continued)
SITE AREA EMERGENCY EAL# 17.3.a Major damage to irradiated fuel with actual or clear potential for significant release of radioactive material to the environment as indicated by:
(A and B)
A. Dropping, bumping, or otherwise rough handling of a new OR irradiated fuel bundle with irradiated fuel in the pool.
and B. (1 or2)
- 1. Refueling floor area radiation monitor reading 1000 times normal annunciation on Panel C601 (2C601) or indication on Panel I C600 (2C600).
or
- 2. Reactor Building vent stack monitoring system high radiation annunciation or indication on Panel 0C630 or 0C677.
OR EAL# 17.3.b Damage to irradiated fuel due to uncontrolled decrease in the fuel pool level to below the level of the fuel as indicated by:
(A and B)
A. (1 or2)
- 1. Uncovering of irradiated fuel confirmation by verification of significant leakage from spent fuel pool.
or
- 2. Visual observation of water level below irradiated fuel in the pool.
and B. (1 or2)
- 1. Refueling floor area radiation monitor annunciation on Panel 1C651 (2C651) or indication on Panel 1C600 (2C600).
or
- 2. Reactor Building vent stack monitoring system high radiation annunciation or indication on Panel 0C630 or 0C677.
GENERAL EMERGENCY EAL# 17.4 None.
EP-AD-000-200, Revision 19, Page 32 of 38
Tab 6 EP-PS-1 11-6 18 - STEAM LINE BREAK UNUSUALEVENT EAL# 18.1 None.
ALERT EAL# 18.2 MSIV malfunction causing leakage as indicated by:
(A and B)
A. Valid MSIV closure signal or indication on Panel 1C601 (2C601).
and B. (1 or2)
- 1. Valid Main Steam Line flow indication on Panel 1C652 (2C652).
or
- 2. Valid Main Steam Line radiation indication on Panel 1C600 (2C600).
(CONTINUED ON NEXT PAGE)
EP-AD-000-200, Revision 19, Page 33 of 38
Tab 6 EP-PS-1 11-6 18- STEAM LINE BREAK (continued)
SITE AREA EMERGENCY EAL# 18.3 Steam line break occurs outside of containment without isolation as indicated by:
(A or B or C or D)
A. (1 and 2)
- 1. Failure of both MSIVs in the line with the leak to close as indicated by position indication on Panel I C601 (2C601).
and
- 2. (a orb)
- a. High MSL flow annunciation on Panel C601 (2C601) or indication on Panel 1C652 (2C652). _
or
- b. Other indication of main steam leakage outside containment.
or B. (1 and 2)
- 1. Failure of RCIC steam isolation valves HV-F008 and HV-F007 to close as indicated on Panel 1C601 (2C601).
and
- 2. (aorborcordoreorf)
- a. RCIC steamline pipe routing area high temperature annunciation on Panel 1C601 (2C601), or indication on Panel 1C614 (2C614).
or
- b. RCIC equipment area high temperature annunciation on Panel 1C601 (2C601) or indication on Panel 1C614 (2C614).
or
- c. RCIC steamline high flow annunciation on Panel 1C601 (2C601).
or
- d. RCIC steamline tunnel ventilation high delta temperature annunciation on Panel 1C601 (2C601).
or
or
- f. Other indication of steam leakage from the RCIC system.
(CONTINUED ON NEXT PAGE)
EP-AD-000-200, Revision 19, Page 34 of 38
Tab 6 EP-PS-1 11-6 18 - STEAM LINE BREAK (continued)
SITE AREA EMERGENCY (continued) or C. (1 and 2)
- 1. Failure of HPCI steam isolation valves HV-F002 and HV-F003 to close as indicated by position indicator on Panel C601 (2C601).
and
- 2. (aorborcordoreorf)
- a. HPCI steamline pipe routing area high temperature annunciation on Panel 1C601 (2C601), or indication on Panel 1C614 (2C614).
or
- b. HPCI equipment area high temperature annunciation on Panel 1C601 (2C601) or indication on Panel 1C614 (2C614).
or
- c. HPCI steamline high flow annunciation on Panel 1C601 (2C601).
or
- d. HPCI steamline tunnel ventilation high delta temperature annunciation on Panel 1C601 (2C601).
or
or
- f. Other indication of steam leakage from the HPCI system.
or D. Any other un-isolatable steam line breaks.
GENERAL EMERGENCY EAL# 18.4 None.
EP-AD-000-200, Revision 19, Page 35 of 38
Tab 6 EP-PS-1 11-6 19 - TOXIC/FLAMMABLE GASES UNUSUAL EVENT EAL# 19.1 Nearby or onsite release of potentially harmful quantifies of toxic or flammable material as indicated by:
Visual observation or notification received by the control room operator.
ALERT EAL# 19.2 Entry of toxic or flammable gases into the facility, with subsequent habitability problem as indicated by:
Visual observation, direct measurement, or notification received by the control room operator.
SITE AREA EMERGENCY EAL# 19.3 Toxic or flammable gases enter vital areas, restricting access and restricted access constitutes a safety problem, as determined by:
(A and B)
A. Shift Supervisor's evaluation.
and B. Visual observation, direct measurement, or notification received by control room operator.
GENERAL EMERGENCY EAl# 19.4 None.
EP-AD-000-200, Revision 19, Page 36 of 38
Tab 6 EP-PS-1 11 -6 20 - TECHNICAL SPECIFICATION SAFETY LIMIT UNUSUAL EVENT EAL# 20.1 Abnormal occurrences which result in operator complying with any of the Technical Specification SAFETY LIMIT ACTION statements indicated by:
(A or B or C or D)
A. Exceeding THERMAL POWER, low pressure or low flow safety limit 2.1.1.1.
or B. Exceeding THERMAL POWER, high pressure and high flow safety limit 2.1.1.2.
or C. Exceeding REACTOR VESSEL WATER LEVEL safety limit 2.1.1.3.
or D. Exceeding REACTOR COOLANT SYSTEM PRESSURE safey limit 2.1.2.
ALERT EAL# 20.2 None.
SITE AREA EMERGENCY EAL# 20.3 None.
GENERAL EMERGENCY EAL# 20.4 None.
EP-AD-000-200, Revision 19, Page 37 of 38
Tab 6 EP-PS-1 11-6 21 - DRY FUEL STORAGE UNUSUAL EVENT EAL# 21.1.a. Situations are occurring or have occurred during the transport of the irradiated spent fuel to the onsite storage facility, which jeopardize the integrity of the spent fuel or its container as indicated by:
(A or B)
A. Radiological readings exceed 2 R/hour at the external surface of any transfer cask or horizontal storage module.
or B. Radiological readings exceed 1 R/hour one foot away from the external surface of any transfer cask or horizontal storage module.
OR EAL# 21.1.b. Situations are occurring or have occurred at the irradiated spent fuel storage facility, which jeopardize the integrity of the dry cask storage system as indicated by:
(A or B)
A. Radiological readings exceed 2 R/hour at the external surface of any transfer cask or horizontal storage module.
or B. Radiological readings exceed 1 Rlhour one foot away from the external surface of any transfer cask or horizontal storage module.
ALERT EAL# 21.2 None.
SITE AREA EMERGENCY EAL# 21.3 None.
GENERAL EMERGENCY EAL# 21.4 None EP-AD-000-200, Revision 19, Page 38 of 38
Tab 8 EP-PS-1 11-8 Control #_
I EMERGENCY NOTIFICATION REPORT I El THIS IS A DRILL THIS IS NOT A DRILL
- 1. This is: at Susquehanna Steam Electric Station.
(Communicators Name)
My telephone number is: . The time is (Callback telephone number) (Time notification initiated)
- 2. EMERGENCY CLASSIFICATION:
o UNUSUAL EVENT O STE AREA EMERGENCY El ALERT al GENERAL EMERGENCY Ea The event has been terminated.
UNIT: a] ONE T IME: DATE:
El Two (lime classification/
termination declared)
- (Date classilcation/
temination declared)
El ONE &TWo El INITIAL DECLARATION THIS REPRESENTS A/AN:
El ESCALATION El No CHANGE
} IN CLASSIFICATION STATUS For initial declaration, static update, or escalation, provide current EAL number only.
- For status reports, significant events, or when directed by the
DESCRIPTION OF THE EVENT:
- For termination, write emergency has been terminated.
- 4. THERE IS: El No
[- AN AIRBORNE NON-ROUTINE RADIOLOGICAL RELEASE IN PROGRESS El A LIQUID
- 5. WHEN GENERAL EMERGENCY IS THE INITIAL EVENT, PROVIDE PROTECTIVE ACTION RECOMMENDATIONS BELOW: (Control Room Use only, TSC and EOF mark NfA.)
- 6. WIND DIRECTION IS FROM: WIND SPEED IS: mph.
(Data from 10 meter meteorological tower, available on PICSY.)
[- THIS IS A DRILL El THIS IS NOT A DRILL APPROVED: Time: Date:
(ED, RM, or EOFSS) (Time form approved) (Date form approved)
EP-AD-000-310, Revision 4, Page 1 of 1
Tab 11 EP-PS-1 11-11 CORE DAMAGE ESTIMATE I (Primary System Breach Inside Containment)
NOTE: It is important to quickly provide a status of the present situation and a prognosis on whether the situation is expected to degrade, improve, or remain the same, (i.e., within 5 to 10 minutes of a change in plant status).
1.0 INDICATORS USED 1.1 Containment Radiation Use Attachment 1, A, B, or C, as applicable, to determine the amount and type of fuel damage using containment radiation monitors. These figures were taken from the US NRC Response Technical Manual, RTM-96. Obtain the containment radiation levels from SPDS or the Control Room indicators.
NOTE (1): Correction for the pre-release backgroundcadiation levels may be required as listed below.
Gap or In-Vessel Melt - The background radiation monitor value is normally low ( 4 R/hr) relative to 1% gap or in-vessel melt release. Consequently, the monitor reading does not require correction for background level in determining the type and amount of fuel damage. If the background radiation monitor reading is > 4 R/hr, the monitor reading should be corrected for the background level in determining the type and amount of fuel damage.
Spiked or Normal Coolant - The radiation monitor value requires correction for the background level. Correct the monitor reading to account for the normal background level in determining the type and amount of fuel damage.
NOTE (2): Containment radiation will go up if there is fuel damage. The increase will depend on the type of fuel damage, and whether or not there was a LOCA, Drywell and/or Wetwell sprays were used, and the amount of blowdown from the Reactor Vessel to the Suppression Pool.
In the case of a LOCA, the fuel damage estimate depends strongly on whether or not containment sprays are being used.
Special care should be taken to confirm the operation of containment sprays.
EP-AD-000457, Revision 7, Page 1 of 10
Tab 11 EP-PS-111-11' 1.2 Containment Hydrogen Use Attachment 2, taken from the US NRC Response Technical Manual RTM-96, to determine the amount and type of fuel damage using Hydrogen Concentration. Obtain the containment Hydrogen levels from SPDS or the Control Room indicators.
NOTE: Containment Hydrogen will increase if there is a LOCA inside the containment and significant fuel damage.
1.3 Coolant Fission Product Concentration vs. Core Damage Coolant sampling will indicate the amount of fuel damage, but in most cases, will take too long for use in dose projections. If PASS sample data becomes available, the Nuclear Fuels Engineer is responsible for assuring a fuel damage calculation based on the measured fission product inventories is performed. The results of this analysis should be compared to previous calculations using other methods.
1.4 Plant Transient Precipitating Fuel Damage If the core experienced a loss of coolant accident and is not covered within 15 minutes, refer to Attachment 3 taken from the US NRC Response Technical Manual RTM-96. The amount of time the core was uncovered can be determined using SPDS. Using the attached figures will provide an estimate of potential fuel damage. Coolant samples must be taken to accurately assess fuel damage.
The type of transient experienced by the reactor leading to fuel damage can be an indicator of the amount and type of fission products released.
- If the core experienced an overpower/pressure transient, a gap release may have occurred.
- If the core experienced a mechanical failure, which could produce flow blockage, there may be localized fuel melt.
- If the core experienced a mechanical perturbation, such as a seismic event or a large steam line break causing a large delta pressure across the core, a gap release could result.
- If the Reactor failed to shut down (ATWS) with a subsequent loss of cooling, there may be fuel melt.
EP-AD-000-457, Revision 7, Page 2 of 10
Tab 11 EP-PS-111-11 Containment Radiation Monitor Response Direct Release Path to Drv well (Sprays Off)
I.E+07 I.E+06 I .E405 I .E+04 C I.E.03 I)
EC 1.E+02 C
0 C.
0 4-. I.E+01 C
U C
C I.E+00 I.E-01 0
I.E-02 I.E-03 I.E-04 I.E-05 In-Vessel Melt Gap Spiked Coolant Normal Coolant Core Damage and Time After Reactor Shutdown Note 1: This figure should be used only when there Is a primary system breach Inside containment and a direct release path to the Drywell.
Note 2: See Attachment 3 to determine If fuel melt occurred (core uncovered or fuel blockage).
ATTACHMENT A EP-AD-000-457, Revision 7, Page 3 of 10
Tab 11 EP-PS-111-11 Containment Radiation Monitor Response Direct Release Path to DPv well (Sprays On) 1.E'07 1.E+06 100%
I .E+05 50%
_-10% 100%
100%
1.E+04 5% _50%,
1%___10%*_ 5%_
1 -- %10%
i_
-= I.E+03 _ .--50% ^0 55% _
0 .%
1.E+01 i> 1.Esoo 10 i -%
0 0 ~~~ ~ ~ ~ ~ ~~~~~=-0
~~~ 1.E+OO
~ ~ ~ ~ ~ ~ 1 10 1 .E-01-%_ 50% -0 S .E0l 100%
1.1E-03 10% 0
=5% -
I.E10 24h th 24 h 10%4 I .E-04 1% %
1.BD~~~~~~~ATCMN 1.E.05 h 24h lh 24h lb 24h lb 24h In-Vessel Melt Gap Spiked Coolant Normal Coolant Core Damage and Time After Reactor Shutdown Note 1: This figure should be used only when there Is a primary system breach Inside containment and a direct release path to the Drywell.
Note 2: See Attachment 3 to determine If fuel melt occurred (core uncovered or fuel blockage).
ATTACHMENT IlB EP-AD-000-457, Revision 7, Page 4 of 10
Tab 11 EP-PS-1 11-1 1 Containment Radiation Monitor Response Direct Release to Wetwell and Not to Drvwell 1.E+06 1.E+05 1.E+04 1.E403 c1.E+02 V
E C
- 1.E01 C
0 U
2 i) 1.E+O0 V
- 1.E01
° 1.E-02 0
1.E-03 I.E-04 I.E-05 1.E-06 1h 24h 1h 24h 1h 24h 1h 24h In-Vessel Melt Gap Spiked Coolant Normal Coolant Core Damage and Time After Reactor Shutdown Note 1: This figure should be used only when there Is a primary system breach inside containment and a direct release path to the Wetwell without a primary release to the Drywell.
Note 2: See Attachment 3 to determine If fuel melt occurred (core uncovered or fuel blockage).
ATTACHMENT IC EP-AD-000-457, Revision 7, Page 5 of 10
Tab 11 EP-PS-1 11-1 1 CONTAINMENT HYDROGEN VS CORE DAMAGE
% LacJWae Raction & Cor Damage ib" 40 30 4.p paftl Ma Tbp:
Pi rodbe com 4 Sw~db 10 <.aaah~dlf
.* SPFB 0.'
0.1 1 10 HMO2 % InContainment
- U kI&fl Sw., NURW R72. p. 4.3; do=e Ms NUREG43U. VOL S.:
I Zarcce, NUREG-1L3; NUREGfCRA4O4; NUREGICK67. Tae 49. P.71 dy ,,=
AnACHMENT 2 EP-AD-000-457, Revision 7, Page 6 of 10
Tab 11 EP-PS-111-11 WATER INJECTION REQUIRED TO COOL CORE BY BOILING CAUTION:
These rates are those required to remove decay heat from a 3000 MW(t) plant by boiling. If there is a break requiring make up or injected water, more water than indicated will be required to both keep the core covered and cooled.
CAUTION:
If the core has been uncovered, the fuel temperature will have increased significantly.
[ Additional flow will be required to accommodate the heat transfer necessary to return to equilibrium fuel temperature. ]
NOTE:
These curves are based on a 3000 MW(t) plant operated at a constant power for an infinite period and then shutdown instantaneously. The decay heat power is based on ANS-5.1N18.6.
Assuming the injected water is at 800 F, these curves are within 5% for pressures between 14 psia to 2500 psia. These curves are within 20% for injected water temperatures up to 212 0F.
ATTACHMENT 3 (Page 1 of 4)
EP-AD-000-457, Revision 7, Page 7 of 10
Tab 11 EP-PS-1 11-1 1 WATER INJECTION REQUIRED TO COOL CORE BY BOILING While the top of the active cre is uncovered, assume that the fuel vill heat up at -21/sc. n increased core tperature will resut in fuel pi damage as shown below.
noMA _
Thes estimates are reasonable factor of Grdy p116 SONa)
- 2) i the core s Unovered within a few boors of shutdown (Including failure to d scra) . if thee Ls sufficient injection, core eatup may be stopped or sloved due POB IP Cg bm to tam cooling.
stem cooling say not .- Wf prevent core damage under accidenticnodW coudtions.h - ftJ~~~~~ IF -'
-' Af _d mam.~~wmim a~ 11 vid air ~
am9 o od m- OF
-_mm -
Il panso ki XuIN i: 0-f9oo, H34CM4S24, _MUZG0956 ATTACHMENT 3 (Page 2 of 4)
CAUTION: If the core Is severely damaged, It may not be In a coolable state even If covered again with water.
NOTE: If there is sufficient injection, core heatup may be stopped or slowed due to steam cooling. Steam cooling may not prevent core damage under accident conditions.
EP-AD-000-457, Revision 7, Page 8 of 10
Tab 11 EP-PS-1 11-1 1 WATER INJECTION REQUIRED TO COOL CORE BY BOILING WEQiOf I(gm) ReQUD To US LOS?
Xy DOZLZNG tE to DEC FM k 3000 X(t)
PUM (1/2-24 ROM AFZZR SHUTDO) cm sol C nenJac so as m ~~~~~~~~~a
\
30 Is o -_s Im o oO S
0.5 ¶ 4 5 WW10 (gpu) REUD TO R1 WER LOST BY BOZLG DE TO EaY E R A 3000 301(t)
PLAN ( to 20 DAYS AE SHUTDOW) a &*l se 6Cnecw dWZo 100
.4o0 . 40 a
o .3. .. . , ,. . C I1 a a A 5 6 7 *
- tO t0 30 Cm" At W Vpa_
ATTACHMENT 3 (Page 3 of 4)
EP-AD-000-457, Revision 7, Page 9 of 10
Tab 11 EP-PS-111-11 WATER INJECTION REQUIRED TO COOL CORE BY BOILING CWeerre redeu
)(F) (C) POSSible cut damage 0 >600 >315
- Nose 05 to 0.75 1800-2400 980-1300
- Locat fad melfg
- Dum= of dadfg with steam prondonm (qwdesc: Zr-lO macsia with ppid H6 gemmlion)
- Rapid fad ddLug fim (Vp
-uefitm d oe wC
- Possie alo m (up) of motc cag
- Possie oo le ce 1 to 3+ >4200 >2300
- Mehbomh of vess with possiBe Co i= failue and td of additional ess volade fission Zsxw: NUREG=424S. NUREOfCRh462. NUREGICR-429. UREGICR374. SUREG0900, NURE-96. NUREG-1 M. and NURBG-146S.
ATTACHMENT 3 (Page 4 of 4)
EP-AD-00-457, Revision 7, Page 10 of 10
Tab 12 EP-PS-111-12 CORE DAMAGE ESTIMATE II (Small or no primary system breach inside Containment)
This instruction provides a method of estimating the percentage of fuel that has failed using the Containment Post-Accident Radiation Monitor (CPARM) readings on panel 1C601 (2C601) during an accident. Since the Containment Post-Accident Radiation Monitor readings are readily available, this calculation provides a quick assessment of core damage. This estimate only applies if there is a small or no primary system breach within containment.
1.0 LIMITAIONS OF THE METHOD 1.1 This procedure will only determine qualitatively the amount of fuel damage. The method uses Containment Post-Accident Radiation Monitor Readings to calculate the percentage of failed fuel during an accident where the fission products are released from the fuel rod cladding. The methodology is based on assumptions with large uncertainties that can significantly affect the results.
1.2 To use this method, the accident scenario up to the time of the Containment Post-Accident Radiation Monitor Reading must be well understood to estimate the fuel temperatures required by this procedure.
1.3 In addition, a Containment Post-Accident Radiation Monitor Reading and the time the reading was obtained must be available.
2.0 RESPONSIBILITIES 2.1 The Nuclear Fuels Engineer, Lead Technical Support Engineer, or designee collects information and makes estimates and determinations described in this procedure.
3.0 INSTRUCTIONS 3.1 Determine if Cladding Failure, Fuel Overheat, or Fuel Melt has occurred:
3.1.1 Cladding Failure is expected if peak cladding temperature remains less that 22000 F, but the Containment Post-Accident Radiation Monitor readings have increased.
3.1.2 Fuel Overheat is expected if peak cladding temperature exceeds 2200 0F, but the maximum volume-averaged fuel pellet temperature remains less that 45000°F.
3.1.3 Fuel .Melt is expected if any volume-averaged fuel pellet temperature exceeds 45000 F.
EP-AD-000-270, Revision 6, Page 1 of 3
Tab 12 EP-PS-1 11-1 2 3.2 Since the fuel melt temperatures are dependent on the event progression, specific guidelines cannot be given to cover all scenarios. Some judgment will have to be made or specific temperature calculations will have to be performed during the event. However, the following provides guidelines for a few known scenarios.
3.2.1 If a main steamline high radiation trip causes the scram and the core remains covered, usually cladding failure can be assumed and is possibly due to debris fretting, short term DNB, or PCI. However, if channel flow blockage is suspected, overheat or melting may occur.
3.2.2 For loss-of-inventory-after-the-reactor-is-shutdown scenarios, use Attachment 3 to Tab 4 to estimate if Fuel Melt has occurred.
3.3 Determine the Time After Reactor Shutdown that a Containment Post-Accident Radiation Monitor Reading was obtained. -
3.4 Determine if the event has resulted in a primary system breach inside primary containment (increase in drywell pressure/temperature and inventory makeup to the vessel is required to maintain level in the vessel). If the total primary system water released to the drywell is equivalent to less than 9,000 gallons or no primary system breach has occurred inside primary containment, use Figure 1.
Otherwise, use Core Damage Estimate I (Tab 4).
Note: The 9,000 gallon value Is about 10% of the fluid volume of the reactor vessel and primary piping (main steam, reactor recirculation, and feedwater).
3.5 Determine Fraction of Fuel Failed (FFF) as follows:
CPARM Reading Expected 100% FuelFailureCPARM Reading EP-AD-000-270, Revision 6, Page 2 of 3
Tab 12 EP-PS-1 11 -12 GURE t CONTAINMENT HIGH RANGE RADION MONIOR READINGS THAT ARE PECD WITH 100% OF THE FUEL FAID FOR AN EVENT WITH NO PRIMARY SYSTEM BREACH INSIDE OONTANMENT i_
it .
. V--
a.........- -----------------.
. Cr----Ji
. * \ \ 8; *.FUEL T-'- MB cs: \ .* .*... . ..... ._ . :
.E.
ch a .-
aO 0 I
..... . V. .
.4....n..........
. .................. . .......... ,.................. :==
04
/ 1...............................
0
.. UFL OVERHEAT~
0 C,
oidI .............. .. a. ........ .................
- .......... - ..-.----- r.....-... .. ---------
0 la
....... ... . ....... C. ;._..5.- .
0 LU
-i oF~e*sso........... .... ...... .. <
CLADDING FAWRE III *4
- v 0 6 ' 15 20 25 Post-Shutdown Time HRS)
EP.AD-000-270, Revision 6, Page 3 of 3
Tab 13 EP-PS-1 11-1 3 FUEL DAMAGE WORKSHEET General The following information should be kept current at all times after facility activation. This information is used by Dose Calculators to perform dose projections for support of Protective Action Recommendations required within fifteen minutes of a General Emergency classification.
Engineering Support is required to provide an estimate of percent fuel damage to the TSC Dose Calculator and the EOF Dose Assessment Staffer in a timely manner, allowing sufficient time for a dose projection to be performed.
The following information is the best estimate possible within the time and using available data.
If no information is provided to the Dose Calculator, default values will be used to determine dose projections. This may result in more severe conditions prompting a non-conservative protective action recommendation.
1.0 ISOTOPIC DETERMINATION: (choose one)
UNKNOWN MIX (Containment Rad <5R hr)
NORMAL COOLANT LEAK (Containment Rad <5Rfhr)
LOCA No Fuel Damage (Iodine Spike, Containment Rad <1OR/hr)
LOCA CLAD FAILURE (Containment Rad 1.5E+02 - 5.OE+04 Rhr)
LOCA FUEL MELT (Containment Rad 8.OE+03 - 1.OE+06 R/hr)
FUEL HANDLING ACCIDENT EP-AD-000-454, Revision 3, Page 1 of 2
Tab 13 EP-PS-1 11-1 3 2.0 CORE CONDITION: (choose one)
Gap Release (Core uncovered for 15-30 minutes)
In Vessel Severe Damage (Core Uncovered >30 Minutes)
Vessel Melt Through EP-AD-000-454, Revision 3, Page 2 of 2