ML021190536
| ML021190536 | |
| Person / Time | |
|---|---|
| Site: | Prairie Island  |
| Issue date: | 04/17/2002 |
| From: | Nazar M Nuclear Management Co |
| To: | Document Control Desk, Office of Nuclear Security and Incident Response |
| References | |
| Download: ML021190536 (226) | |
Text
Committed to Nuclear Excellence Nuclear Management Company, LLC Prairie Island Nuclear Generating Plant 1717 Wakonade Dr. East
- Welch MN 55089 April 17, 2002 US Nuclear Regulatory Commission Attn: Document Control Desk Washington, DC 20555 PRAIRIE ISLAND NUCLEAR GENERATING PLANT Docket Nos. 50-282 License Nos. DPR-42 Docket Nos. 50-306 License Nos. DPR-60 Prairie Island Emergency Plan Implementing Procedures - F3 Emergency Response Plan Implementing Procedures Furnished with this letter are the Prairie Island Nuclear Generating Plant Emergency Plan Implementing Procedures F3. This revision includes the following procedures:
INDEXES: Emergency Plan Implementing Procedures TOC REVISIONS F3-17 Core Damage Assessment Rev 9 F3-3 Responsibilities During a Notification of Unusual Event Rev 18 F3-4 Responsibilities During an Alert, Site Area, or General Emergency Rev 28 F3-6 Activation & Operation of Technical Support Center Rev 16 F3-30 Transition to Recovery Rev 6 F3-2 Classifications of Emergencies Rev 29 DELETIONS INSTRUCTIONS:
Please post changes in your copy of the Prairie Island Nuclear Generating Plant Emergency Plan Implementing Procedures. Procedures, which have been superseded or deleted, should be destroyed.
Please sign and return the acknowledgment of this update to Bruce Loesch, Prairie Island Nuclear Generating Plant, 1717 Wakonade Drive East, Welch, MN 55089.
If you have any questions, please contact Mel Agen at 651-388-1121 Extension 4240.
Mano K. Nazar Site Vice Preside Prairie Island Nu ear Generating Plant c: USNRC - Steve Orth, Region 111 (2 copies)
NRC Resident Inspector (w/o attachment)
J Silberg (w/o attachment)
M Agen (w/o attachment)
Records Management (Doc Control Copy) (w/o attachment)
NL File (w/o attachment)
Mfst Num: 2002 - 0291 Date : 04/15/02 FROM Bruce Loesch/Mary Gadient Loc : Prairie Island TO UNDERWOOD, BETTY J Copy Num: 515 Holder : US NRC DOC CONTROL DESK SUBJECT : Revisions to CONTROLLED DOCUMENTS ******** * ********* ******
Procedure # Rev Title Revisions:
F3-17 9 CORE DAMAGE ASSESSMENT F3-3 18 RESPONSIBILITIES DURING A NOTIFICATION OF EVENT F3-4 28 RESPONSIBILITIES DURING AN ALERT, SITE ARE OR GENERAL EMERGENCY F3-6 16 ACTIVATION & OPERATION OF TECHNICAL SUPPOR F3-30 6 TRANSITION TO RECOVERY F3-2 29 CLASSIFICATIONS OF EMERGENCIES UPDATING INSTRUCTIONS Place this material in your Prairie Island Controlled Manual or File. Remove revised or cancelled material and recycle it. Sign and date this letter in the space provided below within ten working days and return to Bruce Loesch or Mary Gadient, Prairie Island Nuclear Plant, 1717 Wakonade Drive E.,
Welch, MN 55089.
Contact Bruce Loesch (ext 4664) or Mary Gadient (ext 4478) if you have any questions.
Received the material stated above and complied with the updating instructions Date
PRAIRIE ISLAND NUCLEAR
Title:
GENERATING PLANT Emergency Plan Implementing Procedures TOC Effective Date : 04/15/02 Approved By: (JbJtL4 SBPS Sup !/J*L--Il Document # Title Rev F3-1 ONSITE EMERGENCY ORGANIZATION 19 F3-2 CLASSIFICATIONS OF EMERGENCIES 29 F3-3 RESPONSIBILITIES DURING A NOTIFICATION OF UNUSUAL 18 EVENT F3-4 RESPONSIBILITIES DURING AN ALERT, SITE AREA, 28 OR GENERAL EMERGENCY F3-5 EMERGENCY NOTIFICATIONS 20 F3-5.1 SWITCHBOARD OPERATOR DUTIES 8 F3-5.2 RESPONSE TO FALSE SIREN ACTIVATION 9 F3-5.3 RESPONSE TO RAILROAD GRADE CROSSING BLOCKAGE 8 F3-6 ACTIVATION & OPERATION OF TECHNICAL SUPPORT CENTER 16 F3-7 ACTIVATION & OPERATION OF OPERATIONAL SUPPORT 15 CENTER (OSC)
F3-8 RECOMMENDATIONS FOR OFFSITE PROTECTIVE ACTIONS 19 F3-8.1 RECOMMENDATIONS FOR OFFSITE PROTECTIVE ACTIONS FOR 12 THE ON SHIFT EMERGENCY DIRECTOR /SHIFT MANAGER F3-9 EMERGENCY EVACUATION 16 F3-10 PERSONNEL ACCOUNTABILITY 18 F3-11 SEARCH & RESCUE 7 F3-12 EMERGENCY EXPOSURE CONTROL 14 F3-13 OFFSITE DOSE CALCULATIONS 14 F3-13.3 MANUAL DOSE CALCULATIONS 10 F3-13.4 MIDAS METEOROLOGICAL DATA DISPLAY 6 F3-13.5 ALTERNATE METEOROLOGICAL DATA 4 Page 1 of 3
PRAIRIE ISLAND NUCLEAR Title : Emergency Plan Implementing Procedures TOC GENERATING PLANT Effective Date : 04/15/02 Document # Title Rev F3-13.6 WEATHER FORECASTING INFORMATION 11 F3-14.1 ONSITE RADIOLOGICAL MONITORING F3-14.2 OPERATIONS EMERGENCY SURVEYS 9 F3-15 RESPONSIBILITIES OF THE RADIATION SURVEY TEAMS 22 DURING A RADIOACTIVE AIRBORNE RELEASE F3-16 RESPONSIBILITIES OF THE RADIATION SURVEY TEAMS 16 DURING A RADIOACTIVE LIQUID RELEASE F3-17 CORE DAMAGE ASSESSMENT 9 F3-18 THYROID IODINE BLOCKING AGENT (POTASSIUM IODIDE) 9 F3-19 PERSONNEL & EQUIPMENT MONITORING & DECONTAMINATION 6 F3-20 DETERMINATION OF RADIOACTIVE RELEASE 17 CONCENTRATIONS F3-20.1 DETERMINATION OF STEAM LINE DOSE RATES 9 F3-20.2 DETERMINATION OF SHIELD BUILDING VENT STACK 9 DOSE RATES F3-21 ESTABLISHMENT OF A SECONDARY ACCESS CONTROL POINT 10 F3-22 PRAIRIE ISLAND RADIATION PROTECTION GROUP RESPONSE 16 TO A MONTICELLO EMERGENCY F3-23 EMERGENCY SAMPLING 18 F3-23.1 EMERGENCY HOTCELL PROCEDURE 11 F3-24 RECORD KEEPING DURING AN EMERGENCY 7 F3-25 REENTRY 8 F3-26.1 OPERATION OF THE ERCS DISPLAY 7 F3-26.2 RADIATION MONITOR DATA ON ERCS 6 F3-26.3 ERDS - NRC DATA LINK 1 F3-29 EMERGENCY SECURITY PROCEDURES 18 F3-30 TRANSITION TO RECOVERY 6 Page 2 of 3
PRAIRIE ISLAND NUCLEAR Title : Emergency Plan Implementing Procedures TOC GENERATING PLANT Effective Date : 04/15/02 Document # Title Rev F3-31 RESPONSE TO SECURITY RELATED THREATS 5 F3-32 REVIEW OF EMERGENCY PREPAREDNESS DURING OR AFTER 2 NATURAL DISASTER EVENTS Page 3 of 3
PRAIRIE ISLAND NUCLEAR GENERATING PLANT EMERGENCY PLAN IMPLEMENTING PROCEDURES REFERENCE USE
"* Proceduresegments may be performed from memory.
"* Use the procedure to verify segments are complete.
"* Mark off steps within segment before continuing.
"* Procedureshould be availableat the work location.
O.C. REVIEW DATE: OWNER: EFFECTIVE DATE 0 " ý *_ - c, M. Werner 'i-/S --0,)
Page 1 of 12
PRAIRIE ISLAND NUCLEAR GENERATING PLANT EMERGENCY PLAN IMPLEMENTING PROCEDURES TABLE OF CONTENTS Section Title Page 1.0 PURPOSE ...................................................................................................... 3 2.0 APPLICABILITY ............................................................................................... 3 3.0 PRECAUTIONS ............................................................................................... 3 4.0 RESPONSIBILITIES ........................................................................................ 4 5.0 DISCUSSION ................................................................................................... 5 5.1 Definitions ............................................................................................. 5 5.2 Emergency Action Levels ..................................................................... 6 5.3 The Emergency Classification/Declaration/Implementation Process .... 6 5.4 Technical Specification Required Shutdown NUEs ............................... 7 5.5 Rapidly Escalating then De-escalating Events ...................................... 8 5.6 The Emergency Action Level Reference Manual Number ..................... 9 6.0 PREREQUISITES ............................................................................................ 9 7.0 PROCEDURE ................................................................................................. 10 LIST OF ATTACHMENTS Attachment 1 - Summary of Emergency Action Levels Page 2 of 12
PRAIRIE ISLAND NUCLEAR GENERATING PLANT EMERGENCY PLAN IMPLEMENTING PROCEDURES 1.0 PURPOSE The purpose of this procedure is to specify the Emergency Action Levels that indicate an emergency condition exists and to properly classify the emergency into one of the four graded levels of emergency classifications. This procedure partially satisfies the requirement of 1 OCFR50.47 concerning the existence of an emergency classification and action level scheme.
2.0 APPLICABILITY This instruction SHALL apply to all Shift Managers (SM), Shift Supervisors (SS), Control Room Operators (CRO), Emergency Directors (ED) and Emergency Manager (EM).
3.0 PRECAUTIONS 3.1 Attempt to verify the indications by checking secondary or coincident indicators.
3.2 An emergency classification should be made based on current plant conditions described in Attachment 1 of this procedure.
3.3 These emergency classifications do not apply to offsite transportation incidents that do NOT affect safe operation of the plant. Currently, the Radiation Protection group is responsible for offsite transportation incident assessment involving plant related shipments.
3.4 Rapidly Escalating Then De-escalating Events 3.4.1 In the case of an event that rapidly escalates then de-escalates or begins at a higher emergency class then rapidly de-escalates, the initial emergency classification should be based on current plant conditions.
3.4.2 During initial notifications to the NRC, the NRC should be informed of the current emergency classification and also the highest emergency classification reached during the course of the event. Emphasize the current emergency classification.
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I-i PRAIRIE ISLAND NUCLEAR GENERATING PLANT EMERGENCY PLAN IMPLEMENTING PROCEDURES NUMBER:
F3 CLASSIFICATIONS OF EMERGENCIES REV:
F3-2 29 3.4.3 It may be discovered that a condition existed which met the emergency plan criteria but no emergency was 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. The NRC 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 previously reported event was misclassified) within one hour of the discovery of the undeclared (or misclassified) event provides an acceptable alternative.
3.5 Continuously monitor the Control Room instrumentation, radiation monitors, or any other developments which would be indicative of further system degradation. Be prepared to escalate to a more severe emergency classification.
4.0 RESPONSIBILITIES 4.1 Duty Shift Manager has the responsibility to authorize the initial emergency classification.
4.2 Shift Supervisor of the unaffected unit has the responsibility to assist the Shift Manager as necessary including authorization of an emergency classification.
4.3 Shift Supervisor of the affected unit has the responsibility to direct activities related to the operation of the affected unit.
4.4 Emergency Director has the responsibility to authorize an emergency classification whenever an Alert, Site Area, or General Emergency is declared and the EOF is not activated.
4.5 If the EOF is activated and fully functional, the Emergency Manager has the responsibility to authorize an emergency classification.
4.6 Control Room Operators and affected unit Shift Supervisor have the responsibility to assist the Shift Manager or unaffected unit Shift Supervisor in the identification and verification of control board indications.
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PRAIRIE ISLAND NUCLEAR GENERATING PLANT EMERGENCY PLAN IMPLEMENTING PROCEDURES 5.0 DISCUSSION 5.1 Definitions 5.1.1 Notification of Unusual Event - events that are in progress 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.
5.1.2 Alert - events are in progress or have occurred which involve actual or potential substantial degradation of the level of safety of the plant. It is the lowest level of emergency classification when some necessity for emergency planning and offsite response is necessary.
Any releases expected are limited to small fractions of the EPA Protective Action Guideline exposure levels.
5.1.3 Site Area Emergency - events are in progress or have occurred which involve actual or likely major failure of plant functions needed for protection of the public.
Any releases are not expected to exceed the EPA Protective Action Guideline exposure levels except near the site boundary.
5.1.4 General Emergency - events are in progress or have occurred which involve actual or imminent substantial core degradation or melting with a potential for loss of containment integrity.
Releases during a General Emergency can be reasonably expected to exceed the EPA Protective Action Guideline exposure levels offsite for more than the immediate site area.
5.1.5 Emergency Action Levels (EAL) - specific instrument readings, surface or airborne contamination levels or radiation dose rates that designate a specific emergency class requiring emergency measures for that class.
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PRAIRIE ISLAND NUCLEAR GENERATING PLANT EMERGENCY PLAN IMPLEMENTING PROCEDURES 5.2 Emergency Action Levels Attached to this procedure is a Summary of Emergency Action Levels, Attachment 1. This summary identifies the four emergency classifications, the initiating condition(s), emergency action levels for each classification, and, where applicable, specific instruments and indications to be used to detect and classify an emergency.
The emergency action levels for each classification and the instrument readings and indications listed do not reflect a complete list of instrumentation that will show abnormal indications but does list those key parameters useful in classifying the event.
The Summary of Emergency Action Levels lists are based on the initiating conditions as required by Appendix 1 of NUREG-0654, accidents analyzed in the Prairie Island USAR, and the NRC Branch Position on Acceptable Deviations From NUREG-0654/ FEMA-REP-1, July 11, 1994.
5.3 The Emergency Classification/Declaration/Implementation Process There are three distinct phases to consider: Classification, Declaration and Implementation.
5.3.1 Classification
The act of assessing the EALs to determine the appropriate classification which the ongoing events are categorized. This may take a reasonable length of time (5 to 15 minutes for most situations) depending upon the complexity of the situation. This assessment period is consistent with the NRC Branch Position on Timeliness of Classification of Emergency Conditions, EPPOS No. 2.
5.3.2 Declaration
The act of formally declaring the classification based on the assessment of EALs. This is the point at which the classification time is set and the 10CFR50, App. E 15-minute offsite notification clock starts.
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PRAIRIE ISLAND NUCLEAR GENERATING PLANT EMERGENCY PLAN IMPLEMENTING PROCEDURES
5.3.3 Implementation
The act of making the notification and/or augmentation of the emergency organizations.
5.3.4 Ideally, the Emergency Notification Report Form (PINGP 577) should be filled out to near completion while the classification phase is being conducted. Once the declaration is made by the SM/ED/EM, the 15-minute offsite notification time starts. The SM/ED/EM should review the contents of the Emergency Notification Report Form (PINGP 577) to ensure its completeness, verify the correct declaration time arid then sign the form which gives permission to the Shift Emergency Communicator (or Offsite Communicator in EOF) to implement the E-Plan notifications.
5.3.5 Per 10CFR50.72 (a)(3) NRC notification is required immediately after the notification of the state and local agencies (which is completed within about 15 minutes) and not later than one hour after the emergency declaration.
5.4 Technical Specification Required Shutdown NUEs 5.4.1 In some cases, exceeding Technical Specification limits (e.g., RCS leakage, reactor coolant activity, etc.) is considered to be precursors to more serious events and warrant declaration of an NUE.
5.4.2 In other cases, exceeding Technical Specification limits for the period designated in the action statement (ITS. ACTIONS Table) is an analyzed condition of the plant and does not, by itself, represent an emergency.
These events are reportable in accordance with 10 CFR 50.72 as a non-emergencies.
However, if the plant is not brought to the required operating mode within the allowable Technical Specifications action statement time limit (ITS. ACTIONS Table COMPLETION Time), then a declaration of an Unusual Event should be declared.
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PRAIRIE ISLAND NUCLEAR GENERATING PLANT EMERGENCY PLAN IMPLEMENTING PROCEDURES 5.4.3 With regard to Emergency Plan classifications, Operations should handle a Technical Specification required shutdown in the following manner:
A. The conditions of the plant should come first. That is, if the condition warrants initiating power reduction immediately, do so. The E-Plan classification can appropriately follow.
B. Following the initiation of the reduction in power or temperature, the classification phase of the E-Plan is started. Review of the EALs should be done to assess for proper classification. Once the Shift Manager has determined the appropriate classification for the event, the Shift Manager should declare the classification and note the time of declaration (this begins the 15-minute offsite notification clock).
This classification phase should be done within a reasonable time frame (5 to 15 minutes for most instances) determined by the circumstances.
C. Once the declaration is made, the Shift Manager should review the contents of the Emergency Notification Report Form (PINGP 577) to ensure its completeness, verify the correct declaration time and then sign the form which gives permission to the Shift Emergency Communicator to implement the E-Plan notifications.
5.5 Rapidly Escalating then De-escalating Events In the case of an event that rapidly escalates then de-escalates or begins at a higher emergency class then rapidly de-escalates, the initial emergency classification should be based on current plant conditions.
It may be discovered that a condition existed which met the emergency plan criteria but no emergency was 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. The NRC 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 previously reported event was misclassified) within one hour of the discovery of the undeclared (or misclassified) event provides an acceptable alternative.
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PRAIRIE ISLAND NUCLEAR GENERATING PLANT EMERGENCY PLAN IMPLEMENTING PROCEDURES The NRC should be informed of the current emergency classification and the highest emergency classification reached during the course of the event during the initial NRC notification via the ENS phone. The Shift Manager should ensure that this notification be performed by an appropriate individual other than the SEC using PINGP Form 666, Event Notification Worksheet. To avoid possible confusion, other offsite authorities will be informed of the current classification during the initial notification and then given the full description of the highest emergency classification reached during the first update on the follow-up notification.
5.6 The Emergency Action Level Reference Manual Number Prairie Island Nuclear Generating Plant has prepared a written manual (EAL Reference Manual) to provide general information about Emergency Action Levels to offsite authorities who are involved in nuclear plant emergency planning. This manual provides a description with text and drawings of the various conditions that might cause the Prairie Island Nuclear Generating Plant to classify an event. By understanding what a particular condition or event means, emergency workers at the various offsite agencies should develop a clear idea of what is occurring at the plant during the emergency.
Each initiating condition in this procedure is followed by a cross reference number that corresponds to the appropriate classification condition in the EAL Reference Manual. When the Emergency Notification Report Form (PINGP 577) is completed, the initiating condition statement and the EAL Reference Manual cross reference number should be included on the form by using the label from "Emergency Initiating Conditions Labels - PINGP 1189" or writing this information as it appears in the shaded box on the appropriate page in the Summary of Emergency Action Levels.
6.0 PREREQUISITES An off-normal condition corresponding to one of the initiating events described in Attachment 1 of this procedure is occurring or has occurred.
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PRAIRIE ISLAND NUCLEAR GENERATING PLANT EMERGENCY PLAN IMPLEMENTING PROCEDURES 7.0 PROCEDURE 7.1 Any significant event that may be classified as an emergency condition SHALL be reported to the Shift Supervisor, Shift Manager and/or Emergency Director immediately.
NOTE: The events may be instrumentation readings or visual observations made by plant personnel.
7.2 Attempt to verify the initial indication by comparing the indication to redundant instrument channels or to other related parameters, visual observations, and field reports as applicable.
7.3 The Shift Manager, unaffected Shift Supervisor or Emergency Director SHALL assess the situation and determine the emergency classification, using the guidelines of Attachment 1.
7.4 In those cases when an event rapidly escalates, then de-escalates or begins at a higher classification, then rapidly de-escalates, the initial emergency classification should be based on current plant conditions.
7.4.1 Inform the NRC of the current emergency classification and the highest emergency classification reached during the course of the event during the initial NRC ENS notification.
7.4.2 It may be discovered that a condition existed which met the emergency plan criteria but no emergency was 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. The NRC 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 previously reported event was misclassified) within one hour of the discovery of the undeclared (or misclassified) event provides an acceptable alternative.
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PRAIRIE ISLAND NUCLEAR GENERATING PLANT EMERGENCY PLAN IMPLEMENTING PROCEDURES 7.5 The Shift Supervisor of the affected unit should take immediate actions, using applicable plant operating procedures to return the plant to normal (or Mode 5, Cold Shutdown, if determined to be necessary).
7.6 If the EOF is not activated, the Shift Manager or Emergency Director SHALL declare the appropriate emergency classification and perform actions as specified in the appropriate responsibility procedure applicable to emergency classification:
7.6.1 For a Notification of Unusual Event, proceed to F3-3.
7.6.2 For an Alert, Site or General Emergency, proceed to F3-4.
If the EOF is activated, contact the Emergency Manager for consultation on whether or not to change the emergency classification. The Emergency Director is responsible to formulate the new classification while the Emergency Manager is responsible to authorize the reclassification.
7.7 Continue to assess and watch for changing parameters or visual indication of further system degradation and be prepared to escalate to a more severe emergency classification as indicated by the Emergency Action Levels in Attachment 1.
7.8 As plant conditions stabilize during a Notification of an Unusual Event (NUE) or Alert, consider terminating the event classification.
7.8.1 For the NUE and Alert classifications, the event may be terminated once the following criteria are met:
A. The plant is in stable condition with at least one fission product barrier intact, and B. Radioactive gaseous and liquid effluent are being controlled within the following limits:
- 1. Gaseous effluent release rates (or resulting dose rates) are within plant limits as defined in Section 3.1 of H4, Offsite Dose Calculation Manual (ODCM), and Page 11 of 12
PRAIRIE ISLAND NUCLEAR GENERATING PLANT EMERGENCY PLAN IMPLEMENTING PROCEDURES
- 2. Liquid effluent release rates (or resulting concentrations) are within the plant limits as defined in "Old 10CFR20 Appendix B in Table II, Column 2 (April 1992)" located in H4, ODCM, Table 4.3 and C. The potential for future degradation of plant conditions is small.
7.8.2 Termination of an NUE classification may be performed by the Shift Manager. Review of 5AWI 1.13.0, Plant Event Investigation and Recovery, should be completed to determine if this procedure needs to be implemented and the Plant Manager informed of the review.
7.8.3 Termination of an Alert classification may be performed by the Emergency Director if the EOF is not activated. Once the EOF is fully functional, the Emergency Manager SHALL terminate the Alert classification when the conditions are met for termination.
7.8.4 Termination of an Alert classification includes the dismissal of the site Emergency Response Organization. Any necessary in-plant or on-site follow-up activities should be coordinated and managed by the normal plant site organization. In some cases, conditions may require the establishment of a Recovery Organization in which case the Emergency Director and Emergency Manager should make this determination based on the extent of damage or other considerations. In this case 5AWI 1.13.0 should be reviewed and implemented as determined appropriate.
7.9 As plant conditions stabilize during a Site Area or General Emergency, consider transition to the Recovery phase.
Ifthe Site Area Emergency event does not require significant repairs or analysis beyond the capabilities of the normal plant site organization and the conditions of 7.8.1. A, B, & C NOTE: are met, then the Site Area Emergency may be terminated without a transition to Recovery. However review and implementation of 5AWI 1.13.0 should be completed as appropriate.
Transition to Recovery should be directed by the Emergency Manager with coordinated recovery planning by the site Emergency Response Organization.
See F3-30, "Recovery", for instruction on transition to Recovery. Review and implementation of 5AWI 1.13.0 should be completed as appropriate.
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PRAIRIE ISLAND NUCLEAR GENERATING PLANT EMERGENCY PLAN IMPLEMENTING PROCEDURE NUMBER:
CLASSIFICATION OF EMERGENCIES F3-2 F3 ATTACHMENT 1 REV: 29
SUMMARY
OF EMERGENCY ACTION LEVELS Page 1 of 60
PRAIRIE ISLAND NUCLEAR GENERATING PLANT EMERGENCY PLAN IMPLEMENTING PROCEDURE NUMBER:
F3 CLASSIFICATION OF EMERGENCIES
~~ATTACHMENT 1RE: F3-2 2
INITIATING CONDITION INDEX No. Condition Description Page 1 Safety System Functions 3 2 Abnormal Primary Leak Rate 4 3 Deleted 8 4 Abnormal Primary/Secondary Leak 9 5 Core Fuel Damage 13 6 Loss of 2 of 3 Fission Product Barriers 15 7 Secondary Coolant Anomaly 21 8 Radiological Effluents 26 9 Major Electrical Failures 31 10 Control Room Evacuations 35 11 Fires 36 12 Plant Shutdown Functions 38 13 Fuel Handling Accidents 44 14 Deleted 46 15 Deleted 46 16 Security Threats 47 17 Hazards to Plant Operations 48 18 ISFSI (Independent Spent Fuel Storage Installation) Events 53 19 Natural Events 54 20 Other 59 Page 2 of 60
F3-2 CLASSIFICATION OF EMERGENCIES ATTACHMENT 1 REV. 29 Page 3 of 60 Condition 1 : Safety System Functions Failure of a safety or relief valve in a safety related system (EAL Ref Manual 1B) to close following reduction of applicable pressure.
Visual and/or audible indication at discharge of open steam generator Notification of safety or relief valve Unusual Event As determined by SS or SM ý-
('D Pressurizer safety > 185 *F & NOT valve outlet decreasin temperature, TI-436 Pressurizer safety > 185 *F & NOT valve outlet decreasing temperature, TI-437 Pressurizer relief > 185 'F & NOT valve outlet decreasing temperature, TI-438
-CN Acoustic Monitors indicate flow from L..
either the pressurizer PORV's or safeties PZR Pressure on at I< 2335 psig least 2/3 channelsI PRT Level increasing PRT Temperature increasing PRT Pressure increasing
F3-2 CLASSIFICATION OF EMERGENCIES ATTACHMENT 1 REV. 29 Page 4 of 60 Condition 2: Abnormal Primary Leak Primary system leak rate from unidentified or uncontrolled sources exceeding Technical (EAL Ref Manual 2A)
Specifications.
Primary system leak rate, other than controlled leakage sources, exceeds Technical Specifications 3.1 .C (IT.S. 3.4.14)
Requires a unit shutdown, as determined Notification of by the SM or SS Unusual Event Reactor power reduction or cooldown has been initiated Primary coolant leak rate greater than 50 gpm. (EAL Ref Manual 2B )
Above Cold Shutdown (IT.S. >200 *F)
(Modes 1, 2, 3, & 4)
SAlert
- Sum of charging line flow + seal injection flow >50 gpm more than A Flows > 50 pm the sum of letdown flow + RCP seal leakoff flow A rapid temperature decrease in the RCS results in the same indications SS, SM, or ED opinion
- Decreasing pressurizer level, as indicated by LI-426, LI-427, LI-428 R2 (Containment Area Rad Mon) > 50 mR/hr R7 (Containment Area Rad Mon) >50 mR/hr R-1 1 (Containment Vent Rad Mon) > 5x10 4 cpm
[Containment Position]
R-12 (Containment Vent Rad Mon) > 6x103 cpm
[Containment Position]
F3-2 CLASSIFICATION OF EMERGENCIES ATTACHMENT 1 REV. 29 Page 5 of 60 Condition 2: Abnormal Primary Leak LOCA with leak rate in excess of available pump capacity.
(Small LOCA with failure of SI or Large LOCA greater than (EAL Ref Manual 2C)
SI flow.)
Above Cold Shutdown (Modes 1, 2, 3, & 4)
Site Area SI Actuated annunciator illuminated Emergency "A" SG is not depressurizing in an uncontrolled manner "B" SG is not depressurizing in an uncontrolled manner High containment radiation levels High containment sump levels High containment humidity High containment pressure
F3-2 CLASSIFICATION OF EMERGENCIES ATTACHMENT 1 REV. 29 Page 6 of 60 Condition 2: Abnormal Primary Leak Small LOCA and initially successful ECCS.
Subsequent failure of containment heat removal systems over several hours could lead to core melt (EAL Ref Manual 2D) and likely failure of containment.
Above Cold Shutdown I (IT.S. >200 *F)
bi (Modes 1,2, 3, & 4)I General LOCA has occurred Emergency SS, SM or ED opinion that _
loss of ECCS is imminent Containment temp has become excessive, and is >300 0 F still rising Containment pressure is >23 psig high and increasing Operating containment Zero spray pumps Operating containment fan < One train coil unit trains
F3-2 CLASSIFICATION OF EMERGENCIES ATTACHMENT 1 REV. 29 Page 7 of 60 Condition 2: Abnormal Primary Leak Small and large LOCA's with failure of ECCS to perform leading to severe core degradation or melt In (EAL Ref Manual 2E) from minutes to hours. Ultimate failure of containment likely for melt sequences.
Above Cold Shutdown (Modes 1, 2, 3, & 4) General SS, SM or ED opinion that Emergency core cooling capability does not exist High Head SI failure Low Head SI failure High dose equivalent 1-131 as determined by sample analysis Valid high radiaiton R-48 (Containment High Range Area Rad Mon)
Valid high radiaiton R-49 (Containment High Range u=
Area Rad Mon) 0 ot~ Decreasing RCS pressure
-J,,u Low subcooling margin and decreasing "A" SG is not depressurizing in an uncontrolled manner "B" SG is not depressurizing in an uncontrolled manner Pressurizer Low Pressure Trip RCS pressure decreasing uncontrollably High containment radiation levels High containment sump levels High containment humidity High containment pressure
F3-2 CLASSIFICATION OF EMERGENCIES ATTACHMENT 1 REV. 29 Page 8 of 60 Condition 3 : Abnormal Coolant Temperature/Pressure DELETED Deleted based on NRC Branch Position On Acceptable Deviations From Appendix 1 to NUREG-0654/FEMA-REP-1, July 11, 1994.
F3-2 CLASSIFICATION OF EMERGENCIES ATTACHMENT 1 REV. 29 Page 9 of 60 Condition 4: Abnormal Primary / Secondary Leak Primary /Secondary leak rate exceeding Technical (EAL Ref Manual 4A)
Specifications.
SG primary to secondary > 150 GPD through any 1 SG Notification of leakage exceeds Tech Spec Unusual Event 3.1.C (IT.S. 3.4.14) limits Primary /Secondary leak rate greater than 50 gpm. (EAL Ref Manual 4B)
Above Cold Shutdown (Modes 1,2, 3, & 4)
Decreasing pressurizer Alert level LI-426, LI-427, LI-428 R1 5 (Air ejector)
Sum of charging line flow plus seal injection greater than 50 gpm more than the sum of Letdown Flow and RCP Seal Leakoff flow SS, SM, ED opinion
F3-2 CLASSIFICATION OF EMERGENCIES ATTACHMENT 1 REV. 29 Page 10 of 60 Condition 4: Abnormal Primary /Secondary Leak Failure of steam generator tube(s) resulting in ECCS (EAL Ref Manual 4C) actuation.
"SI Actuated " annunciator illuminated on _
the Aqua panel I Alert Safety Injection Flow indicated by FI-925 and/or FI-924 and pump discharge pressure corresponding to flow Lo Pzr pressure ; PI-429,PI-430, PI-431 < 1815 psig PZR level cannot be maintained on at >5%
least 2/3 channels LI-426, LI-427, LI-428I R-51 Steam Line Rad Monitor I > bkg R-52 Steam Line Rad Monitor > bk R-15 Air Ejector Rad Monitor 2x1° 4 cpm Unexpected Rise in either SG level
F3-2 CLASSIFICATION OF EMERGENCIES ATTACHMENT 1 REV. 29 Page 11 of 60 Condition 4: Abnormal Primary /Secondary Leak Failure of steam generator tube(s) resulting in ECCS (EAL Ref Manual 4D) actuation and high potential for loss of containment.
"SI Actuated "annunciator illuminated on the Aqua panel Site Area Safety Injection Flow indicated by FI-925 Emergency and/or FI-924 and pump discharge pressure corresponding to flow Lo Pzr pressure; PI-429,PI-430, PI-431 Cu PZR level cannot be maintained on at Cu least 2/3 channels LI-426, LI-427, LI-428 R-51 Steam Line Rad Monitor Cu 2.
R-52 Steam Line Rad Monitor R-15 Air Ejector Rad Monitor Unexpected Rise in either SG level Indications of a steam line break outside containment, upstream of the ruptured 0 Steam Generator MSIV 0 Cu Failure of ruptured Steam Generator MSIV 0 to close with indications of steam line
'4 Cu break downstream of MSIV 0
Cu 0 Ruptured SG safety or relief valve opens
- 0. and then fails to reseat indicated by visual and/or audible indication at discharges As determined by SS, SM, or ED
F3-2 CLASSIFICATION OF EMERGENCIES ATTACHMENT 1 REV. 29 Page 12 of 60 Condition 4 : Abnormal Primary /Secondary Leak Failure of steam generator tube(s) resulting in ECCS actuation and loss of offsite power. (EAL Ref Manual 4E)
"SI Actuated " annunciator illuminated on I_____
the Aqua panel SIeAe Safety Injection Flow indicated by FI-925 Emergency and/or FI-924 and pump discharge pressure corresponding to flow C.)
Lo Pzr pressure ; PI-429,PI-430, PI-431 PZR level cannot be maintained on at least 2/3 channels LI-426, LI-427, LI-428 (j5 R-51 Steam Line Rad Monitor R-52 Steam Line Rad Monitor 4
R-15 Air Ejector Rad Monitor I >2x1 0_cm Unexpected Rise in either SG level -
I.
4.16 KV Bus 11 [21], 4119304 [4172804] 350Volts
.0 4.16 KV Bus 12 [22], 4119305 [4172805] < 3 5 0 0 Volts
< 3500 Volts 4.16 KV Bus 13 [23], 4119306 [4172806]
< 3500 Volts 4.16 KV Bus 14 [24], 4119307 [4172807]
D1 [D5 ] Tach I900 rpm[1200 rpm D1 [D5] Gen Volts 4000 - 4400 volts -
Safeguard Bus 15 [25] source from DG IL closed, Breaker 15-2 [25-2]
D2 [D6 ] Tach 9 10 [1 rpm m D2 [D6 ] Gen Volts 4000- 4400 volts Safeguard Bus 16 [26] source from DG closed, Breaker 16-9 [26-16]
F3-2 CLASSIFICATION OF EMERGENCIES ATTACHMENT 1 REV. 29 Page 13 of 60 Condition 5: Core Fuel Damage Fuel Damage Indication (EAL Ref Manual 5A)
R-9 (Letdown line radiation monitor) in alarm Notification of Unusual Event R-9 > 1 R/hr R9 confirmed by portable instrumentation High coolant activity sample exceedinc Tech Specs 3.1 .D (IT.S. 3.4.17) for total specific activity in primary coolant High coolant activity sample exceeding Tech Specs 3.1 .D (IT.S. 3.4.17) limit on primary coolant iodine activity which requires a unit Severe Loss of fuel (EAL Ref Manual 5B) cladding Dose equivalent 1-131 Alert R-9 (Letdown line radiation monitor) in alarm R-9 R-9 confirmed by portable instrumentation
F3-2 CLASSIFICATION OF EMERGENCIES ATTACHMENT 1 REV. 29 Page 14 of 60 Condition 5 : Core Fuel Damage Degraded core with possible loss of coolable geometry (EAL Ref Manual 5C)
SS, SM, or ED opinion Site Area Emergency 4ý 1 Core Delta T increasing No core Delta T Dose Equivalent 1-131 >
I-I- Core exit thermocouples RVLIS full range
F3-2 CLASSIFICATION OF EMERGENCIES ATTACHMENT 1 REV. 29 Page 15 of 60 Condition 6 : Loss of 2 of 3 Fission Product Barriers Loss of 2 of 3 fission product barriers with a potential loss of 3rd barrier. (EAL Ref Manual 6)
GENERAL EMERGENCY
- 1. Three permutations exist for loss of 2 of 3 fission product barriers with a potential loss of 3rd barrier; A. Failure of cladding and primary coolant boundary with potential loss of containment.
B. Failure of cladding and containment with potential loss of primary coolant boundary.
NOTES:
C. Failure of containment and primary coolant boundary with potential loss of cladding.
These 3 permutations are represented in the following 5 cases, each with its own set of EAL's:
- 2. All cases are applicable to operations above Cold Shutdown (Modes 1, 2, 3, & 4).
F3-2 CLASSIFICATION OF EMERGENCIES ATTACHMENT 1 REV. 29 Page 16 of 60 Condition 6: Loss of 2 of 3 Fission Product Barriers Case 1: Loss of clad, loss of primary coolant boundary (LOCA), and high potential for loss of (EAL Re *f Manual 6) containment.
Valid high radiation R-48 (Containment High Range Area Rad Man)
Valid high radiation R-49 General
' (Containment High Range Area 7ý Emergency Rad Mon)
Dose equivalent 1-131 > 300 uCi/cc Pressurizer low pressure trip <1900 si S RCS pressure decreasing uncontrollably EE High containment pressure o High containment humidity
' High containment sump levels High containment radiation levels "o "A" SG is not depressurizing in an uncontrolled manner "B" SG is not depressurizing in an uncontrolled manner Containment approaching > 23 p 46psig design pressure and increasing Control Room containment isolation status lights for Train A and B indicate open or partially Z open for any containment penetration SS, SM°or ED opinion Operating containment LZero o spray pumps Operating containment fani < One train k coil unit trains
F3-2 CLASSIFICATION OF EMERGENCIES ATTACHMENT 1 REV. 29 Page 17 of 60 Condition 6: Loss of 2 of 3 Fission Product Barriers Case 2: Loss of clad, SG tube rupture and high potential (EAL Ref Manual 6) for loss of containment Indications of a steam line break outside containment, upstream of the ruptured SG MSIV General Emergency Failure of ruptured SG MSIV to close with indications of steam line break Ruptured SG stuck open safety or relief valve indicated by visual and/or audible steam relief As determined by SS, SM, or ED a
R-51 (Loop A Steam Line)
R-52 (Loop B Steam Line)
R-15 (Air Ejector)
Unexpected Rise in either SG level SI Actuated annunciator illuminated on Aqua panel SI flow indicated by FI-925 and/or FI-924 and pump discharge
.0 pressure corresponding to flow J
Lo PZR pressure PI-429, PI-430, PI-431 PZR level cannot be maintained on at least 2/3 channels LI-426, LI-427, LI-428 Dose equivalent 1-131
F3-2 CLASSIFICATION OF EMERGENCIES ATTACHMENT 1 REV. 29 Page 18 of 60 Condition 6: Loss of 2 of 3 Fission Product Barriers Case 3: Loss of clad, containment failure, and a high potential for loss of the RCS boundary. (EAL Ref Manual 6)
Dose equivalent 1-131 > 300 uCi/cc General Emergency b.
Cold leg temperature decrease in th > 100 *F last 60 minutesl Cu RCS cold leg temperature < 250 'F 00 Steam line break between
-J (
containment and the MSIV Fl o Steam line break downstream of MSIVI MSIV open-close indicator(s) shov open or partially open.
Control room containment isolationi status lights for Train A and B indicate[
open or partially open for any containment penetration SS, SM,or ED opinionL
F3-2 CLASSIFICATION OF EMERGENCIES ATTACHMENT 1 REV. 29 Page 19 of 60 Condition 6: Loss of 2 of 3 Fission Product Barriers Case 4: Loss of RCS boundary (LOCA), loss of (EAL Ref Manual 6) containment, and high potential for loss of cladding Control room containment isolation status lights for Train A and B indicate U) Z open or partially open for any 0*. containment penetration General SS, SM,or ED opinion Emergency RVLIS full range Core exit thermocouples Pressurizer low pressure trip RCS pressure decreasing uncontrollably L "A" SG is not depressurizing in an uncontrolled manner "B" SG is not depressurizing o
in an uncontrolled manner Co High containment radiation levels (3
'4 0 High containment sump levels
,(0 High containment pressure
F3-2 CLASSIFICATION OF EMERGENCIES ATTACHMENT 1 REV. 29 Page 20 of 60 Condition 6: Loss of 2 of 3 Fission Product Barriers Case 5 : Loss of RCS Boundary (SG Tube Rupture),
loss of containment, and high potential for loss of (EAL Ref Manual 6) cladding.
Steam line break between containment and the ruptured Steam Generator MSIV General Stuck open ruptured SG Emergency 0
PORV or Safety valve 0
U) Steam line break downstream of U) 0
-J ruptured Steam Generator MSIV MSIV open-close indicator shows open or partially open L_
.20 RVLIS full range 0 Core exit thermocouples CL U)
U) 0 R-51 (Loop A Steam Line)
R-52 (Loop B Steam Line) > bk -
cc R-15 Air Ejector Rad Monitor >2x104 cpm Unexpected Rise in either SG level Sl Actuated annunciator illuminated on Aqua panel Sl flow indicated by FI-925 and/or FI-924 and pump discharge pressure corresponding to flow Lo PZR pressure 1815 psig PI-429, PI-430, PI-431 I PZR level cannot be maintained on at least 2/3 channels >5%
LI-426, LI-427, LI-428
F3-2 CLASSIFICATION OF EMERGENCIES ATTACHMENT 1 REV. 29 Page 21 of 60 Condition 7: Secondary Coolant Anomaly Rapid depressurization of secondary (EAL Ref Manual 7A) side.
SS or SM opinion that a rapid reduction in secondary pressure Notification of occurred Unusual Event Steam Generator A/B pressure differential Increasing Containment Pressure Indications of a steam line break outside of containment Failed open Steam Dump, SG PORV, or SG Safety Valve Hi Steam Flow Lo-Lo Tavg Indication of steam line break downstream of MSIVs
F3-2 CLASSIFICATION OF EMERGENCIES ATTACHMENT 1 REV. 29 Page 22 of 60 Condition 7: Secondary Coolant Anomaly Steam line break inside containment with significant (greater than 10 gpm) (EAL Ref Manual 7B) primary to secondary leak rate.
High containment pressure h Alert R-15 Air Ejector Rad Monitor reading [
Steam Line Monitor reading above [
normal background on faulted SG qo)
Unexpected rise in faulted SG level
.2~
Pri/Sec Leak rate calculation Steam generator A/B pressure L differentialI 4C Valid Steamline low pressure safety injection signal 6
> .745x10 lb/hr Hi steam flow Lo-Lo Tavg <540 0F
F3-2 CLASSIFICATION OF EMERGENCIES ATTACHMENT 1 REV. 29 Page 23 of 60 Condition 7: Secondary Coolant Anomaly Unisolable steam line break outside containment with significant (greater (EAL Ref Manual 7C) than 10 gpm) primary to secondary leak rate.
Indications of unisolable steam line break outside containment Site Area Emergency R-15 Air Ejector Rad Monitor reading Steam Line Monitor reading above normal background on faulted SG Unexpected rise in faulted SG level Pri/Sec Leak rate calculation QQ,)
Steam generator A/B pressure differential Valid Steamline low pressure safety Mu injection signal 4C Hi steam flow Lo-Lo Tavg
F3-2 CLASSIFICATION OF EMERGENCIES ATTACHMENT 1 REV. 29 Page 24 of 60 Condition 7: Secondary Coolant Anomaly Steam Line break in containment with greater than 50 gpm primary to secondary leakage and indication of fuel (EAL Ref Manual 7D) damage.
If steam line break is outside containment with SG tube rupture Note: and fuel failure, this may be a General Emergency. See condition # 6, case 2 Primary to Secondary Leakage 1 >50 Site Area Valid high radiaiton R-48 > 2o( Emergency (Containment High Range Area Rad Mon)
Valid high radiaiton R-49 (Containment High Range Area Rad Mon)
Steam generator A/B pressure I >100 psig differential 4K 4pi High containment pressure [4 psig Valid steamline low pressure < 500. psip E safety injection signal Hi steam flow > .745x 106 lb/hr -
Lo-Lo Tavg
F3-2 CLASSIFICATION OF EMERGENCIES ATTACHMENT 1 REV. 29 Page 25 of 60 Condition 7 : Secondary Coolant Anomaly Transient initiated t ,y loss of feedwater and condensate sysstems (principal heat removal system) folIlowed by failure of emergency feedwat er system for (EAL Ref Manual 7E) extended period. Core meliing possibe in several hours. UItimate failure of containment likely if core melts.
Unable to establish* Condensate and Main Feedwater General Emergency Unable to establish* Aux Feedwater High Head Safety Injection not available No Condensate and Main Feedwater No Auxiliary Feedwater SI Actuation per FR-H.1 30 minutes elapsed since SI actuation per FR-H.1 Auxiliary Feedwater not established ECCS Injection not established Note:
"*"Unable to establish" criteria met if procedural attempt to establish condition has been made, but was unsuccessful or if an attempt cannot be made.
F3-2 CLASSIFICATION OF EMERGENCIES ATTACHMENT 1 REV. 29 Page 26 of 60 Condition 8 : Radiological Effluents Airborne Radiological effluent Technical Specifications exceeded (EAL Ref Manual 8A)
Unmonitored gaseous release to the atmosphere which is estimated or suspected to exceed Operations Manual H4, ODCM, Sect. 3.1 limits as determined by Rad Notification of Unusual Event Protection Confirmation of total gaseous activity by Rad i Chemist that a 300 second marinelli gas grab >lx 10-4 uCi/cc sample analysis from 1[2]R-22 (Shield Bldg Vent Rad Mon) 3 Valid 1R-22 (Shield Bldg Vent Rad Mon) >5x 10 m Valid 2R-22 (Shield Bldg Vent Rad Mon) >5x103 cpm Liquid Radiological effluent Technical Specification limits exceeded. (EAL Ref Manual 8B)
Unmonitored liquid release to river which exceeds the limits of Operations Manual H4, ODCM, Sect. 2.1 as determined by Notification of Rad Protection Unusual Event Liquid release exceeds the limits of Operations Manual H4,ODCM, Sect. 2.1 based on sample analysis R-21 (Circ Water Discharge Rad Mon)
>High Alarm setpoint R-18 (Waste Liquid Discharge Rad Mon) 0
>High Alarm setpoint R-18 (Waste Liquid Discharge Rad Mon) 0U isolation valve fails to close R-19 (SG Blowdown Rad Mon) >High Alarm setpoint (while blowdown is directed to river) tJJ R-19 (SG Blowdown Rad Mon) isolation valve fails to close
F3-2 CLASSIFICATION OF EMERGENCIES ATTACHMENT 1 REV. 29 Page 27 of 60 Condition 8: Radiological Effluents Liquid Radiological effluent greater than (EAL Ref Manual 8C) ten times Technical Specification limits.
Unmonitored liquid release to river which exceeds 10 times the limits of Operations Manual H4, ODCM, Sect. 2.1 as Alert determined by Rad Protection Liquid release exceeds 10 times the limits of Operations Manual H4, ODCM, Sect.
2.1 based on sample analysis R-21 (Circ Water Discharge Rad Mon) I
> 10 times High Alarm setpointI R-18 (Waste Liquid Discharge Rad Mon)
> 10 times High Alarm setpoint R-18 (Waste Liquid Discharge Rad Mon) isolation valve fails to close R-1 9 (SG Blowdown Rad Mon) > 10 times Uj High Alarm setpoint (while blowdown is directed to river)
R-1 9 (SG Blowdown Rad Mon) isolation valve fails to close Airborne Radiological effluents greater than ten times Technical Specification instantaneous limits (an instantaneous rate which, if continued for over two hours, would result in about 1 mrem TEDE (EAL Ref Manual 8D) at the site boundary under average met conditions. TEDE = Total Dose Equivalent).
Unmonitored gaseous release to the atmosphere which is estimated or suspected to exceed 10 times Operations Manual H4, ODCM, Sect. 3.1 limits as Alert determined by Rad Protection.
Confirmation by RPS that local measured count rate on 1[2]R-50 (Shield Bldg Vent > 3000 cpm Hi Range Rad Mon) gas chamber Valid 1R-22 (Shield Bldg Vent Rad Mon) >5x104 Cpm q)
Valid 2R-22 (Shield Bldg Vent Rad Mon) > 5xlO4 m Co Valid 1R-50 (Shield Bldg Vent Hi Range > 1.2 mR/hr 0 E* Rad Mon)
Valid 2R-50 (Shield Bldg Vent Hi Range > 1.2 mR/hr Rad Mon)I
F3-2 CLASSIFICATION OF EMERGENCIES ATTACHMENT 1 REV. 29 Page 28 of 60 Condition 8 : Radiological Effluents Airborne Effluent monitors detect levels corresponding to greater than:
- 1. 50 mremrhr TEDE for one-half hour, or
- 2. 250 mrem/hr Thyroid CDE for one-half hour, or
- 4. 2500 mrem/hr Thyroid CDE for two minutes at the site boundary for adverse meteorology.
TEDE = Total Effective Dose Equivalent.
CDE = Committed Dose Equivalent.
r.1 2_h
[/-- --
R a d ia t io n s u rv e y t e a m s m e a s u re >5_
gamma dose rates at site boundary > 500 mR/hr for 2 m Site Area S gsEmergency SRadiation survey teams measure field > 250 mrem/hr for 1A hr
- data at site boundary corresponding to > 2500 mrem/hr for 2 mn Thyroid CDE SValidlR.50 (shield bldg vent hirange) 70mhr Z C Valid 2R-50 (shield bldg vent hi range)
CDE offsite dose projections > 250 mrem/hr for 1/2 hr at the site boundary > 2500 mrem/hr for 2 min 50 mrem hr for 1/2 hr
( TEDE offsite dose projections at the site boundary >:500 mrem/hr for 2 mi
F3-2 CLASSIFICATION OF EMERGENCIES ATTACHMENT 1 REV. 29 Page 29 of 60 Condition 8: Radiological Effluents Effluent monitors detect levels corresponding to:
- 1. 1 rem/hr TEDE, or
- 2. 5 rem/hr Thyroid CDE (EAL Ref Manual 8F) at the site boundary under actual meteorological conditions.
TEDE = Total Effective Dose Equivalent.
CDE = Committed Dose Equivalent.
-ft Radiation survey teams measure gamma dose rates at or beyond the site boundary .General Emergency Radiation survey teams measure field data at the site boundary, corresponding to Thyroid CDE Lu 0 Cu Valid 1 R-50 (shield bldg vent hi range)
Cu Lu Cu Cu Valid 2R-50 (shield bldg vent hi range)
TEDE offsite dose projections at the site Cu.o boundary Thyroid CDE offsite dose projections at the 0.. site boundary
F3-2 CLASSIFICATION OF EMERGENCIES ATTACHMENT 1 REV. 29 Page 30 of 60 Condition 8 : Radiological Effluents Radiation levels or airborne contamination which indicate a severe degradation in the control of radioactive materials (e.g., increase of factor of (EAL Ref Manual 8G) 1000 in direct radiation readings within facility).
_v Alert Unexpected increase in plant area radiation factor of 1000 monitorsI Widespread unexpected iodine or > 1000 DAC particulate airborne concentrations as per 10CFR20 appendix B table 1, Column 3.
> 1,000,000 dpm/100cm 2 beta-gamma Widespread unexpected contamination in the Auxiliary Building > 50,000 dpm/1 00cm 2 alpha activit ..
F3-2 CLASSIFICATION OF EMERGENCIES ATTACHMENT 1 REV. 29 Page 31 of 60 Condition 9 : Major Electrical Failures (EAL Ref Manual 9A) 4.16 KV Bus 11 [21], 4119304 [4172804]
Notification of 4.16 KV Bus 12 [22], 4119305 [4172805] Unusual Event 4.16 KV Bus 13 [23], 4119306 [4172806]
4.16 KV Bus 14 [24], 4119307 [4172807]
L.
a)
C D1 [D5 ] Tach C.
a)
U) D1 [D5] Gen Volts 0
'9 0 Safeguard Bus 15 [25] source from DG U) closed, Breaker 15-2 [25-2]
U) 0
-J D2 [D6 ] Tach D2 [D6 ] Gen Volts Safeguard Bus 16 [26] source from DG closed, Breaker 16-9 [26-16]
(EAL Ref Manual 9B)
Loss of D1 & D2 [D5 & D6] diesel Notification of generators when they are required to Unusual Event be operable by Tech Specs
F3-2 CLASSIFICATION OF EMERGENCIES ATTACHMENT 1 REV. 29 Page 32 of 60 Condition 9 : Major Electrical Failures Loss of offsite power and loss of all onsite AC power (See Site Area (EAL Ref Manual 9C)
Emergency for extended loss).
4.16 KV Bus 11 [21], 4119304 [4172804] <3500 Volts 0
4-.
4.16 KV Bus 12 [22], 4119305 [4172805] 1<3500 Volts Alert 0* 0 4.16 KV Bus 13 [23], 4119306 [4172806] <3500 Volts 0
4.16 KV Bus 14 [24], 4119307 [4172807] 1<350o Volts
-J4 4.16 KV Bus 15 [25], 4191801 [4190401] <4000 Volts 0
4.16 KV Bus 16 [26], 4192401 [4192301] <4000 Volts Loss of offsite power and loss of onsite Loss of offsite power an._dd loss of onsite AC power for more than 15 minutes. (EAL Ref Manual 9D)
Blackout duration I > 15 minutes Site Area Emergency 4.16 KV Bus 11 [21], 4119304 [4172804]
0 4.16 KV Bus 12 [22], 4119305 [4172805]
44 (X 0
4.16 KV Bus 13 [23], 4119306 [4172806]
0
-J 4.16 KV Bus 14 [24], 4119307 [4172807]
'.4 I0
- 4. 4.16 KV Bus 15 [25], 4191801 [4190401]
4.16 KV Bus 16 [26], 4192401 [4192301]
F3-2 CLASSIFICATION OF EMERGENCIES ATTACHMENT 1 REV. 29 Page 33 of 60 Condition 9 : Major Electrical Failures Failure of offsite and oinsite power along with total loss of emer gency feedwater (EAL Ref Manual 9E) makeup capability for hours. This would leacIto eventual core melt and likely failure of containment.
Above Cold Shutdown (Modes 1, 2, 3, & 4) General Emergency Turbine Driven Aux Feedwater pump unavailable Motor Driven Aux Feedwater pump from other unit unavailable 4.16 KV Bus 11 [21], 4119304 [4172804]
0 L.
0 4.16 KV Bus 12 [22], 4119305 [4172805]
0 0 0.
4.16 KV Bus 13 [23], 4119306 [4172806]
0
-J 4.16 KV Bus 14 [24], 4119307 [4172807]
4.16 KV Bus 15 [25], 4191801 [4190401]
oJ~
4.16 KV Bus 16 [26], 4192401 [4192301]
F3-2 CLASSIFICATION OF EMERGENCIES ATTACHMENT 1 REV. 29 Page 34 of 60 Condition 9 : Major Electrical Failures Loss of all onsite DC power (see Site (EAL Ref Manual 9F)
Area Emergency for extended loss).
SS, SM opinion that DC power is not available for one unit Alert "SAFEGUARD LOGIC TRAIN A DCj FAILURE" 47018-0501 [47518-0501]
"SAFEGUARD LOGIC TRAIN B DCI FAILURE" 47018-0502 [47518-05021 Loss of all vital onsite DC power for more than 15 minutes (EAL Ref Manual 9G)
"SAFEGUARD LOGIC TRAIN A DC FAILURE" 47018-0501 [47518-0501] Site Area "SAFEGUARD LOGIC TRAIN B DC 41- Emergency FAILURE" 47018-0502 [47518-0502]f Loss of DC power has lasted more than > 15 min.
15 minutes
F3-2 CLASSIFICATION OF EMERGENCIES ATTACHMENT 1 REV. 29 Page 35 of 60 Condition 10 : Control Room Evacuations Evacuation of the Co ntrol Room anticipated or requirned with control of (EAL Ref Manual 10A) shutdown systems e stablished from Hot Shutdown Panels an d local stations.
SS,SM, or ED determines evacuation of Control Room is anticipated or required with
- Alert control of shutdown systems establishedI from Hot shutdown Panels and local stations Ifreason for evacuation is fire in Control Room or Relay Room, see initiating condition 11 C, t: Fire compromising the functions of safety systems" for possible reclassification.
Evacuation of the Control Room and control of shutdown systems not (EAL Ref Manual 10B) established from hot shutdown panel and local stations within 15 minutes.
Site Area Evacuation of Control Room conducted Emergency Control of shutdown systems not established from S>15 mai.
Hot Shutdown Panels and local stations within 15 minutes
F3-2 CLASSIFICATION OF EMERGENCIES ATTACHMENT 1 REV. 29 Page 36 of 60 Condition 11 : Fires Fire within the plant or ISFSI lasting (EAL Ref Manual 11A) more than 10 minutes.
FIRE: is combustion characterized by heat and light (flame). 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.
Physical observation of a fire within the Reactor Bldg., Auxiliary Bldg., Turbine Bldg.,
Service Bldg., Rad Waste Bldg., Plant Screen House, D1/D2 Room, D5/D6 Bldg., Notification of Cooling Tower Equip. House, Transformers, Unusual Event or on the surface of a loaded spent fuel cask in the ISFS1 Area Operation of an automatic fire suppression I system on the fire Application of manual fire extinguishing equipment by the Fire Brigade Fire continuing after at least 10 minutes of > 10 ml .
fire suppression Fire potentially affecting safety systems. (EAL Ref Manual 111B)
FIRE: is combustion characterized by heat and light (flame). 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.
of a fire that is observation Physical Physical observation of a fire that i'sI affecting one train of a safety systemI Alert SS, SM, or ED opinion
F3-2 CLASSIFICATION OF EMERGENCIES ATTACHMENT 1 REV. 29 Page 37 of 60 Condition 11 : Fires Fire compromising the functions of (EAL Ref Manual 11C) safety systems.
FIRE: is combustion characterized by heat and light (flame). 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.
Physical observation of a fire that is affecting both trains of a safety system for Site Area the same unit Emergency SS, SM, or ED opinion
F3-2 CLASSIFICATION OF EMERGENCIES ATTACHMENT 1 REV. 29 Page 38 of 60 Condition 12 : Plant Shutdown Functions Nonfunctional al*arms in the Ccrntrol (EAL Ref Manual 12B)
Room.
i SS or SM opinion that annunciators 1 -1 a are nonfunctional 1 Notification of Unusual Event Cu Loss of NSSS Annunciator System b C.) and loss of plant computer Loss of BOP Annunciator System and Cu loss of plant computer (ERCS)
Loss of NSSS Annunciator System and BOP Annunciator System (EAL Ref Manual 12C)
Above Cold Shutdownn (IT.S. >200 'F)
(Modes 1, 2, 3, & 4)
annunciators are non-functional Loss of plant computer (ERCS)
L.
b 0 0
CCC.)
Loss of NSSS Annunciator System K "1-i Loss of BOP Annunciator System -
Cu
F3-2 CLASSIFICATION OF EMERGENCIES ATTACHMENT 1 REV. 29 Page 39 of 60 Condition 12 : Plant Shutdown Functions Most or all alarms (annunciators) lost (EAL Ref Manual 12D) and plant transient initiated or in progress.
Above Cold Shutdown (IT.S. >200 F, (Modes 1, 2, 3, & 4) Site Area Emergency Loss of plant computer (ERCS)
SS, SM, or ED opinion that annunciators are non-functional SS, SM, or ED opinion that a transient I has occurred or is in progress I 0) 0 Loss of NSSS Annunciator System Loss of BOP Annunciator System 0)
F3-2 CLASSIFICATION OF EMERGENCIES ATTACHMENT 1 REV. 29 Page 40 of 60 Condition 12: Plant Shutdown Functions Momentary loss of core cooling needed (EAL Ref Manual 12E) for plant Cold Shutdown.
At Cold Shutdown (Mode 5)
Notification of Unusual Event Interruption of RHR core cooling Entry into procedure E4, Core Cooling Following Loss of RHR flow Inability to maintain plant in Cold Shutdown.
Mitigation actions are unable to prevent RCS temperature > 200 0 F I (EAL Ref Manual 12F)
SAlert exceeding 200 'F Loss of water level that has uncovered or will uncover the fuel in the reactor (EAL Ref Manual 12G) vessel while at Cold Shutdown.
At Cold Shutdown (Mode 5) (IT.S" <200 °F)
Site Area Emergency Total loss of RHR system cooling Mitigating actions unable to restore makeup flow greater than inventory loss
F3-2 CLASSIFICATION OF EMERGENCIES ATTACHMENT 1 REV. 29 Page 41 of 60 Condition 12: Plant Shutdown Functions Complete loss of any function needed for plant Hot Shutdown. (Also see (EAL Ref Manual 12H)
Condition #7 for possible General if feed and bleed is initiated).
At or above Hot Shutdown I >2350 F Site Area (IT.S. Hot Standby) (Modes 1, 2, & 3)1 Emergency Inadequate hi pressure injection capability (i.e., loss of 2/2 SI pumps > 15 min.
and loss of all charging pumps for more than 15 minutes)
SS, SM, or ED opinion that plant functions required to maintain Hot Shutdown (IT.S. Hot Standby) not available All steam generator safety valves inoperable Both steam generator PORVs inoperable All steam generator steam dump to atmosphere valves inoperable Steam generator steam dump to condenser valve inoperable Unable to establish* Condensate and Main Feedwater fl ow Unable to establish* Auxiliary Feedwater flow Wide range level in both < 25 %
SG'sI Note:
"*"Unable to establish" criteria met if procedural attempt to establish condition has been made, but was unsuccessful or if an attempt cannot be made.
F3-2 CLASSIFICATION OF EMERGENCIES ATTACHMENT 1 REV. 29 Page 42 of 60 Condition 12 : Plant Shutdown Functions ig a reactor/ (EAL Ref Manual 121)
Reactor or turbine trip due to turbine failure Notification of Unusual Event SS or SM opinion (EAL Ref Manual 12J)
As determined by visual inspection b, Alert SS, SM or ED opinion
F3-2 CLASSIFICATION OF EMERGENCIES ATTACHMENT 1 REV. 29 Page 43 of 60 Condition 12 : Plant Shutdown Functions Failure of the reactor protection system (EAL Ref Manual 12K) to initiate and complete a trip which brings the reactor subcritical.
Any valid reactor trip setpoint has I been exceeded I Alert Intermediate range detector I output not decayingI Transient requiring operation of shutdown systems with failure to trip (EAL Ref Manual 12L)
(continued power generation but no core damage immediately evident).
Failure to bring reactor subcritical with control rods inserted Site Area Emergency No indication of core damage SS, SM, or ED opinion that a transient is in progress Transient requiring operation of shutdown systems with failure to trip which results in core damage or (EAL Ref Manual 12M) additional failure of core cooling and makeup systems (which could lead to core melt).
Failure to bring reactor subcritical General with control rods inserted Emergency RCS pressure Rapidly increasing containment pressure and temperature 0
Dose equivalent 1-131
F3-2 CLASSIFICATION OF EMERGENCIES ATTACHMENT 1 REV. 29 Page 44 of 60 Condition 13 : Fuel Handling Accidents Fuel damage accident with release of (EAL Ref Manual 13A) radioactivity to containment.
During refueling operations Alert SS, SMor ED opinion Valid R2 (containment area)
Valid R7 (containment area)
Containment vent monitor in the Containment Position Valid R1 1 (containment vent)
Valid R12 (containment vent)
Fuel damage accident with release of radioactivity to the fuel handling (EAL Ref Manual 13B) building.
SS, SM, or ED opinion Alert Valid Spent Fuel Pool Area rad monitor R5 Valid R-22 (shield bldg vent)
Valid R-50 (shield bldg vent hi range)
F3-2 CLASSIFICATION OF EMERGENCIES ATTACHMENT 1 REV. 29 Page 45 of 60 Condition 13: Fuel Handling Accidents Major damage to spent fuel in containment or fuel handling building (EAL Ref Manual 13C)
(e.g., large object damages fuel or water loss below fuel level.
SS, SM, or ED opinion Site Area Emergency Valid R48 (containment area hi range) hi Alarm Valid R49 (containment area hi range) hi Alarm SS, SM, or ED opinion Valid R50 (shield bldg vent hi range)
Oil)
I . Valid R5 (SFP area)
F3-2 CLASSIFICATION OF EMERGENCIES ATTACHMENT 1 REV. 29 Page 46 of 60 Condition 14: Coolant Pump DELETED Deleted based on NRC Branch Position On Acceptable Deviation From Appendix 1 to NUREG-0654/FEMA-REP-1, July 11, 1994.
Condition 15 : Contaminated Injured Person DELETED Deleted based on NRC Branch Position On Acceptable Deviation From Appendix 1 to NUREG-0654/FEMA-REP-1, July 11, 1994.
F3-2 CLASSIFICATION OF EMERGENCIES ATTACHMENT 1 REV. 29 Page 47 of 60 Condition 16 : Security Threats Security threat or attempted entry or (EAL Ref Manual 16A) attempted sabotage Security determines the threat to be credible
- nd tfh* thr~t wnr id hut _ An Adirhiqr. imn*r~t1 on the safe operation or shutdown capability ol Notification of theUnusual Event Security discovers an unauthorized attempteq entry by force or stealth (secret) into the, plant's protected area The FBI confirms that an act of attempted sabotage did occur to vital plant equipment orJ security equipment Ongoing security compromise (EAL Ref Manual 16B)
Security Safeguards Contingency event that results in unauthorized personnel commandeering an area within the plant4 protected area, but not controlling shutdown capability or any vital areas Alert Bomb device discovered within plant[
protected area and outside of any vital area Imminent loss of physical control of the (EAL Ref Manual 16C) plant Physical attack on the plant involving imminent occupancy of the Control Room,I F, Site Area auxiliary shutdown panels or other vital areas1 Emergency Bomb device discovered within a vital arealI Loss of physical control of the plant (EAL Ref Manual 16D)
Physical attack on the plant has resulted inr General unauthorized personnel occupying the Contro[ Emergency Room or any other vital areas
F3-2 CLASSIFICATION OF EMERGENCIES ATTACHMENT 1 REV. 29 Page 48 of 60 Condition 17 : Hazards to Plant Operations Aircraft crash onsite or unusual aircraft activity over facility. (EAL Ref Manual 17A)
Reported to the SS or SM _ ____ Notification of Unusual Event C
Aircraft crash onsite within owner controlled fence Unusual threatening aircraft activity over 0 facility Aircraft crash in the protected area. (EAL Ref Manual 17B)
Visual observation by plant personnel and/
or plant security personnel of an aircraft o Alert crash in the protected area Aircraft crash within protected area and affecting vital structures by impact or (EAL Ref Manual 17C) fires with plant not in Cold Shutdown.
Visual observation by plant personnel and/
or plant security personnel of an aircraft crash within protected area and affecting (IT.S. >200 TF) Site Area vital structures by impact or fires with either Emergency unit above Cold Shutdown (Modes 1,2, 3, &4)
F3-2 CLASSIFICATION OF EMERGENCIES ATTACHMENT 1 REV. 29 Page 49 of 60 Condition 17 : Hazards to Plant Operations Near or onsite explosion (EAL Ref Manual 17D)
Visually observed evidence of an unplanned or unexplained explosion within the owner Notification of controlled fence but not affecting plant safe Unusual Event operation Reported to the SS or SM Known explosion damage to facility affecting plant operation (EAL Ref Manual 17E)
Visually observed evidence of an unplanned explosion directly affecting plant safe operation Alert Reported to SS or SM
F3-2 CLASSIFICATION OF EMERGENCIES ATTACHMENT 1 REV. 29 Page 50 of 60 Condition 17 : Hazards to Plant Operations Missile impacts from whatever source on facility (EAL Ref Manual 17F)
Visually observed evidence by plant personnel and/or plant security personnel of missile I j Alert impacts on facility from whatever source Severe damage to safe shutdown equipment from missiles or explosion (EAL Ref Manual 17G) with plant not in Cold Shutdown Above cold shutdown I (IT.S. >200 °F)
(Modes 1,2, 3, & 4)1 Site Area Visually observed evidence by plant Emergency personnel and/or plant security personnel of missile impacts and/or explosion on facility from whatever source Reactor trip system is inoperable 0 Inadequate high pressure injection b
flow (no safety injection or charging pumps are available) 4
'U All steam generator safety valves 0
inoperableI Both steam generator PORVs CU inoperable 4 All steam generator steam dump to 0
atmosphere valves inoperable CU Steam generator steam dump to L condenser valve inoperable 0 Unable to establish* any Condensate L CU and Main Feedwater flow Unable to establish* any Auxiliary Feedwater flow Note:
"* Unable to establish" criteria met if procedural attempt to establish condition has been made, but was unsuccessful or if an attempt cannot be made.
F3-2 CLASSIFICATION OF EMERGENCIES ATTACHMENT 1 REV. 29 Page 51 of 60 Condition 17: Hazards to Plant Operations Near or onsite toxic or flammable gas release (EAL Ref Manual 17H)
Widespread toxic or flammable gaseot hazard being experienced or projectedonsil Notification of (out side of plant) leading to evacuation Unusual Event sheltering of personnel outside the pla Receipt of recommendation by Local, Coun or State Officials to evacuate personnel fro site based on an offsite hazardous flammable gaseous release eve Entry into the plant environs of toxic or flammable gases (EAL Ref Manual 171)
Explosive gas concentrations being
> explosive limits ert measured within the plant at a distance o greater than 10 feet from the source
> 50 ppm hydrazine Toxic gaseous concentrations being > 300 ppm ammonia measured within a larae area of the plant t> 50 pm hydrochloric acid I
at the breathing zone
> 1400 ppm morpholine
> 30 ppm ethanolamine (ETA)
> IDLH for any toxic gas (see D14.4 AOP 1 Note: IDLH =Immediately Dangerous to Life or Health.
Reference:
NIOSH Pocket Guide to Chemical Hazards, June 1994.
F3-2 CLASSIFICATION OF EMERGENCIES ATTACHMENT 1 REV. 29 Page 52 of 60 Condition 17 : Hazards to Plant Operations Entry of toxic or flammable gases into vital areas with plant not in Cold (EAL Ref Manual 17J)
Shutdown.
Above cold shutdown (IT.S.>200 *F)
(Modes 1, 2, 3, & 4) "I- Site Area Emergency Explosive gas concentrations being measured in a maior portion of a vital area from an unisolable source such that lim its further access to the vital area is being prevented at a time when it is needed Toxic gaseous concentrations being measured in a maior portion of a vital area at the breathing zone such that further access to the vital area is being restricted at a time when it is needed Note: IDLH = Immediately Dangerous to Life or Health. IDLH
Reference:
NIOSH Pocket Guide to Chemical Hazards, June 1994.
F3-2 CLASSIFICATION OF EMERGENCIES ATTACHMENT 1 REV. 29 Page 53 of 60 Condition 18 :ISFSI Events ISFSI cask tip over or drop resulting in cask seal leakage (EAL Ref Manual 18A)
Loaded cask tipped over or droppe Notification of e cUnusual Event Overpressure monitoring tank o si pressure decreasing to 0 psi Loss of ISFSI cask/fuel containment barrier (EAL Ref Manual 18B)
While loaded cask is outside the plant_
protected area Alert 42 O
A radiation field near ISFSI cask increase by a factor of 1000 above the 1 ambient radiation field (background > 1,000 Xambient tad. field plus the normal radiation field from the cask(s))
"ED 12.
> 1,000,000 dpm/100 cmý beta-gamma Widespread contamination outside the ISFSI protected >50,000 dpm/100 k a area cmralpha "ao
F3-2 CLASSIFICATION OF EMERGENCIES ATTACHMENT 1 REV. 29 Page 54 of 60 Condition 19 : Natural Events Any confirmed earthquake (EAL Ref Manual 19A)
Confirmed by one or more off-site sources Notification of Unusual Event Annun "Seismic Event" received on the seismograph As experienced by plant personnel Earthquake greater than Operational Basis Earthquake (EAL Ref Manual 19B)
Annun "Operational Basis Earthquake" received on the Alert seismograph As experienced by plant personnel H Earthquake greater than Design Basis Earthquake with plant not in Cold Shutdown (EAL Ref Manual 19C)
Above cold shutdown (IT.S. >200 °F)
(Modes 1, 2, 3, &4)I Site Area Emergency Annun "Design Basis Earthquake" received on the seismograph As experienced by plant personnel
F3-2 CLASSIFICATION OF EMERGENCIES ATTACHMENT 1 REV. 29 Page 55 of 60 Condition 19 : Natural Events Any tornado on site (EAL Ref Manual 19D)
Reported to the SS or SM Notification
" Unusual of Event Visual observation by plant personnel and/or plant security Any tornado striking the facility (EAL Ref Manual 19E)
Reported to the SS or SM, or ED 0 Alert Visual observation by plant personnel and/or plant securityI
F3-2 CLASSIFICATION OF EMERGENCIES ATTACHMENT 1 REV. 29 Page 56 of 60 Condition 19 : Natural Events 50 Year Flood (EAL Ref Manual 19F)
River Intake Elevation > 686 ft Notification of Unusual Event Flood levels approaching design levels (EAL Ref Manual 19G)
River Intake Elevation (USAR 2.4.3.5) - requires both units to > 692 ft Alert be shut down to Mode 2 or lower Flood levels exceeding design levels with plant not in cold shutdown (EAL Ref Manual 19H)
Above Cold Shutdown (IT.S. >200 TF)
(Modes 1,2, 3, & 4)[ Site Area River Intake Elevation - Power operation design level (highest level
>698 ft FLz-ý Emergency transformers will function)
F3-2 CLASSIFICATION OF EMERGENCIES ATTACHMENT 1 REV. 29 Page 57 of 60 Condition 19 : Natural Events Low water levels being experienced or projected beyond usual levels. (EAL Ref Manual 191)
River intake elevation (11/21 Cooling I < 672.5 ft Notification of Water Pump - Low Water Level Trip)I Unusual Event Low water levels being experienced or projected to be near design levels. (EAL Ref Manual 19J)
River intake elevation <669.5 ft Alert With plant not in Cold Shutdown, low water levels being experienced or projected to be less than design levels, or failure of vital equipment with low (EAL Ref Manual 19K) water level.
Above cold shutdown Site Area (Modes 1, 2, 3, &4) Emergency River intake elevation corresponding to loss of Lock &
Dam # 3 Major vital equipment failure River Intake Elevation
F3-2 CLASSIFICATION OF EMERGENCIES ATTACHMENT 1 REV. 29 Page 58 of 60 Condition 19 : Natural Events Sustained winds being experienced or projected near design levels. (EAL Ref Manual 19L)
Sustained wind speed indicated I> 90 mph by met towerF SAlert Sustained winds being in excess of design levels being experienced or projected with plant not in Cold (EAL Ref Manual 19M)
Shutdown.
Above cold shutdown I (IT.S. >200 °F)
(Modes 1, 2, 3, &4)I Site Area Emergency Sustained wind speed indicated I > 0o0 mph by met towerI Any major internal or external events (e.g., fires, earthquake, substantially (EAL Ref Manual 19N) beyond design levels) which could or has caused massive damage to plant systems resulting or potential for resulting in large releases to the offsite environment in excess of the EPA Protective Action Guides.
As determined by the SS, SM, or ED General Emergency
F3-2 CLASSIFICATION OF EMERGENCIES ATTACHMENT 1 REV. 29 Page 59 of 60 Condition 20: Other Conditions that warrant increased awareness on the part of plant operation (EAL Ref Manual 20A) staff or state and/or local offsite authorities.
SM and SS concurrence that plant Notification of conditions warrant increased awareness Unusual Event A plant event had occurred which met a classification, but rapidly de-escalated to current plant conditions which do not meet any classification Inability to reach required shutdown within Technical Specification Limits. (EAL Ref Manual 20B)
Technical Specification LCO action statement (IT.S. ACTION TABLE) not met Notification of requires plant shutdown or cooldown Unusual Event Reactor power reduction or cooldown requirements have been exceeded Conditions that involve other than normal controlled shutdown. (EAL Ref Manual 20C)
Reactor power reduction or cooldown has Notification of been initiated Unusual Event Emergency Safeguard Function equipment did NOT perform its function if required Cooldown rate exceeded Tech Spec Limit SS or SM opinion that lack of control of shutdown or cooldown warrants offsite agency notification
F3-2 CLASSIFICATION OF EMERGENCIES ATTACHMENT 1 REV. 29 Page 60 of 60 Condition 20 : Other Conditions that warrant activation of Technical support Center and nearsite (EAL Ref Manual 20D)
Emergency Operation Facility SS, SM, or ED opinion I ----------------- h.
Alert Other plant conditions that warrant activation of emergency operation centers and monitoring teams or a (EAL Ref Manual 20E) precautionary notification to the public near the site SS, SM, or ED opinion Site Area Emergency Other plant conditions exist, from whatever source, that make release of large amounts of radioactivity in a short (EAL Ref Manual 20F) time period possible, e.g., any core melt situation SS, SM, or ED opinion I General Emergency
PRAIRIE ISLAND NUCLEAR GENERATING PLANT PLANT SAFETY PROCEDURE NUMBER:
RESPONSIBILITIES DURING A F3-3 NOTIFICATION OF UNUSUAL EVENT SREV: 18
"* Proceduresegments may be performed from memory.
"* Use the procedure to verify segments are complete.
"* Mark off steps within segment before continuing.
"* Procedureshould be available at the work location.
O.C. REVIEW DATE: OWNER: EFFECTIVE DATE
) k- M.Werner qI- 159*5 -c; Page 1 of 8
PRAIRIE ISLAND NUCLEAR GENERATING PLANT PLANT SAFETY PROCEDURE NUMBER:
RESPONSIBILITIES DURING A F3 NOTIFICATION OF UNUSUAL EVENT REV:
F3-3 18 1.0 PURPOSE The purpose of this instruction is to delineate the responsibilities of various emergency organization personnel and onsite organizations required to respond to a Notification of Unusual Event.
2.0 APPLICABILITY This instruction SHALL apply to all plant personnel.
3.0 PRECAUTIONS 3.1 All personnel should stay clear of any areas as announced over the public address system.
3.2 All personnel should refrain from using the public address system or telephone system during any emergency situation.
4.0 RESPONSIBILITIES 4.1 Overall Responsibility - Shift Manager 4.2 Assistance, Control Room - Shift Supervisors Control Room Operators TSC - Shift Emergency Communicator Plant - Shift Radiation Protection Specialist Page 2 of 8
PRAIRIE ISLAND NUCLEAR GENERATING PLANT PLANT SAFETY PROCEDURE NUMBER:
RESPONSIBILITIES DURING A F3-3 F3 'NOTIFICATION OF UNUSUAL EVENT REV:
REV:
18 1
5.0 DISCUSSION 5.1 Notification of Unusual Event 5.1.1 Definition Unusual Events are events that are in progress or have occurred which indicate a potential degradation of the level of safety of the plant.
No release of radioactive material requiring offsite response or monitoring is expected unless further degradation of safety systems occurs.
5.1.2 Purpose of the Notification of Unusual Event Class The purpose of the Notification of Unusual Event Emergency classification is to:
A. Have the operating staff come to a state of readiness from the standpoint of emergency response in the event the handling of the initial condition needs to be escalated to a more severe action level class; B. Provide for systematic handling of Unusual Event information.
5.1.3 Plant Actions & Responsibilities:
A. Promptly inform State and/or local offsite authorities of nature of unusual condition as soon as discovered.
B. Assess and respond to Unusual Event.
C. Augment on-shift resources (if needed). This may involve full activation of the Emergence Response Organization depending on plant management discretion.
D. Activate the Technical Support Center, if needed. Working in the TSC may enhance the plant technical support assessment and response.
E. Close out with verbal summary to offsite authorities.
OR F. Escalate to a more severe class.
5.1.4 State and/or local offsite authority actions:
A. Provide fire or security assistance if required.
B. Standby until verbal closeout.
OR C. Escalate to a more severe class.
Page 3 of 8
PRAIRIE ISLAND NUCLEAR GENERATING PLANT PLANT SAFETY PROCEDURE NUMBER:
F3 RESPONSIBILITIES DURING A NOTIFICATION OF UNUSUAL E\ /ENT REV:
F3-3 18 6.0 PREREQUISITES An Unusual Event has been or will be declared.
7.0 PROCEDURE 7.1 Shift Supervisors 7.1.1 Shift Supervisor of affected unit:
A. Proceed to the Control Room (if not already there).
The affected unit Shift Supervisor SHALL remain in the Control Room at all times during accident conditions until properly relieved.
B. If needed, ensure that applicable Emergency Operating Procedures (EOPs) are implemented to respond to the Notification of Unusual Event conditions with the objective of returning the plant to a normal, safe condition (Mode 5, Cold Shutdown, if necessary).
N E The Shift Manager and SEC will be summoned to the Control NT:Room per the EOPs.
C. Direct activities of the Control Room Operators.
7.1.2 Shift Supervisor of unaffected unit:
A. Proceed to the Control Room (if not already there).
B. Direct operations on the unaffected unit.
C. Assist the Shift Manager, as necessary.
Page 4 of 8
PRAIRIE ISLAND NUCLEAR GENERATING PLANT PLANT SAFETY PROCEDURE NUMBER:
RESPONSIBILITIES DURING A F3-3 F NOTIFICATION SREV: OF UNUSUAL EVENT 18 7.2 Shift Manager The initial E-Plan response to the event should be made by the Shift Manager. However, in order to adequately maintain NOTE: ~oversight of the operational aspects of the event, it may be necessary for the Shift Manager to delegate specific E-Plan duties to the unaffected unit SS during the initial response while still maintaining overall authority and control.
7.2.1 Report to the Control Room immediately upon notification.
7.2.2 Assess the emergency condition, event evaluation, and safety related aspects of the plant.
It is recommended that the Shift Manager stands at the Reactor Operator's desk to hear key communications, use NOTE:. an ERCS terminal for monitoring Critical Safety Function Status Trees (CSFSTs), and solicit or answer questions of the SS.
7.2.3 Use PINGP 1125, "Shift Manager/Shift Supervisor Emergency Director Checklist," for specific guidance on emergency plan NUE duties ("Example Only" copy of PINGP 1125 is attached to F3-4).
7.3 Control Room Personnel 7.3.1 Assist the Shift Supervisor as requested.
7.3.2 Utilize applicable Operations Manual procedures to respond to the Notification of Unusual Event emergency classification, as appropriate, with the object of returning the plant to a normal safe status (Mode 5, Cold Shutdown, if necessary).
7.3.3 Announce the location and nature of the emergency over the public address system, if applicable. Direct all non-essential personnel to remain clear of the affected area (if applicable) and to refrain from using the public address system.
7.3.4 Continuously monitor the Control Room instrumentation, radiation monitors, or any other developments which would be indicative of further system degradation. Inform the Shift Supervisor immediately of any changes in the plant status.
Page 5 of 8
PRAIRIE ISLAND NUCLEAR GENERATING PLANT PLANT SAFETY PROCEDURE NUMBER:
RESPONSIBILITIES DURING A F3-3
-- 3- NOTIFICATION OF UNUSUAL EVENT 7.4 Emergency Director 7.4.1 The Emergency Director position is not normally activated during the Notification of Unusual Event.
7.4.2 IF the Shift Manager decides to activate the Emergency Response Organization, THEN the ED position should be initially staffed by the Shift Manager and eventually turned over to an oncoming Emergency Director designee.
7.4.3 IF the TSC is activated, THEN an Emergency Director should be designated as the person in command and control in the TSC.
7.4.4 Those TSC functions and activities necessary for the specific event response need be implemented. Facility checklists may be used as guides. Examples of TSC functions are shown below:
A. Plant Status Board Keeper B. ERCS Operation C. ED Log Keeper D. Emergency Work Status Board E. Emergency Communicators 7.5 Shift Emergency Communicator (SEC) 7.5.1 Report to the Control Room immediately upon notification.
7.5.2 Complete the "Emergency Notification Report Form," (PINGP 577), F3-5 and have it reviewed and approved by the Shift Manager.
NOTE: State and local authorities SHALL be notified within 15 minutes of the decisions of the emergency classification.
7.5.3 Complete the required notification of state and local authorities, and site personnel in accordance with PINGP 579 of F3-5, "Emergency Notifications."
Page 6 of 8
PRAIRIE ISLAND NUCLEAR GENERATING PLANT PLANT SAFETY PROCEDURE NUMBER:
RESPONSIBILITIES DURING A F3-3 F3 NOTIFICATION OF UNUSUAL EVENT REV: 18 7.5.4 Assist as necessary the Shift Manager in the coordination of other required communication contacts (e.g., medical support, fire support, etc.). See the "Mo & PI Emergency Preparedness Telephone Directory" for local support services phone numbers.
7.5.5 Notify appropriate plant staff support personnel to augment the onsite staff or activate the ERO as deemed necessary by the Shift Manager.
7.5.6 Maintain communications throughout the emergency classification as required by F3-5, "Emergency Notifications."
conditions NOTE: Be prepared to notify applicable authorities if escalate to a more severe emergency class.
7.5.7 When the Notification of Unusual Event class has been terminated, close-out the emergency classification, by notifying the state, local, and site personnel in accordance with F3-5, "Emergency Notifications."
.7.6 Other Operations Personnel 7.6.1 Assistant Plant Equipment Operators and Plant Attendants should continue with assigned duties or report to the Control Room if requested.
7.6.2 Relief Shift Operators onsite, should continue with assigned operations unless their assistance is requested.
7.7 Radiation Protection Group 7.7.1 The Shift Radiation Protection Specialist SHALL provide assistance (e.g., sampling, chemistry, radio-chemistry, surveys, etc.) as requested by the Shift Supervisor.
7.7.2 An REC designee should report to the TSC and provide event updates to the Wisconsin Health Dept. representative using PINGP 1246, WI/MN Health Department Initial Call Back Information.
7.7.3 Other Radiation Protection Group personnel should continue with normal duties unless directed to do otherwise.
Page 7 of 8
PRAIRIE ISLAND NUCLEAR GENERATING PLANT PLANT SAFETY PROCEDURE NUMBER:
RESPONSIBILITIES DURING A F3-3 F NOTIFICATION OF UNUSUAL EVENT REV: 18 7.8 Plant Security Force 7.8.1 The designated Security Officer or Security Staff member SHALL perform the responsibilities of the Shift Emergency Communicator (SEC).
7.8.2 All other plant security force should continue with normal duties unless directed to do otherwise.
7.8.3 During off-normal work hours and when requested by the Shift Emergency Communicator (SEC), a designated member of the Security Force should control the telephone switchboard in the TSC.
7.9 All Other Personnel On Site 7.9.1 Continue with normal duties unless directed to do otherwise.
7.9.2 Stay clear of areas as announced over the public address system.
7.10 Business Support Group During normal work hours and when requested by the SEC, a designated person from the plant switchboard operator group should report to the TSC to control the telephone switchboard.
Page 8 of 8
PRAIRIE ISLAND NUCLEAR GENERATING PLANT EMERGENCY PLAN IMPLEMENTING PROCEDURES F 3 RESPONSIBILITIES DURING AN ALERT, SITE AREA, OR GENERAL EMERGENCY NUMBER:
F3-4 REV: 28 REFERENCE USE
"* Proceduresegments may be performed from memory.
"* Use the procedure to verify segments are complete.
"* Mark off steps within segment before continuing.
"* Procedureshould be available at the work location.
O.C. REVIEW DATE: OWNER: EFFECTIVE DATE 0"c-1e -5 :M. Werner q--i5-O0-.)
Page 1 of 74
PRAIRIE ISLAND NUCLEAR GENERATING PLANT EMERGENCY PLAN IMPLEMENTING PROCEDURES NUMBER:
RESPONSIBILITIES DURING AN ALERT, F3-4 F3 JSITE AREA, OR GENERAL EMERGENCY REV: 28 TABLE OF CONTENTS Section Title Page 1L0 PURPOSE .................................................................................................... 4 2.0 APPLICABILITY ............................................................................................... 4 3.0 PRECAUTIONS ............................................................................................... 4 4,0 RESPONSIBILITIES ........................................................................................ 5 5 .0 D IS C US S IO N ................................................................................................. . .6 6.0 PREREQUISITES .......................................................................................... 12 7 .0 P R O C E D UR E ............................................................................................... . . 12 7.1 Shift Supervisor of Affected Unit .......................................................... 12 7.2 Shift Supervisor of Unaffected Unit ...................................................... 13 7 .3 S hift Ma nage r .................................................................................... . . 13 7.4 Operations Group ................................................................................ 14 7.5 Shift Emergency Communicator .......................................................... 17 7.6 Technical Support Coordinator ............................................................ 18 7.7 Operational Support Coordinator ........................................................ 18 7.8 Assembly Point Coordinator ................................................................ 18 7.9 Radiological Emergency Coordinator .............................. 18 7.10 Radiation Protection Group ................................................................. 19 7.11 Security Group .................................................................................... 20 7.12 Logistics Group .................................................................................... 22 7.13 Instrument & Control Group ................................................................. 23 7.14 Maintenance Group ............................................................................. 24 7.15 Engineering Group .............................................................................. 24 7.16 Severe Accident Management Group ................................................. 26 7.17 Contract, Temporary Personnel and Visitors ......................... 26 Page 2 of 74
PRAIRIE ISLAND NUCLEAR GENERATING PLANT EMERGENCY PLAN IMPLEMENTING PROCEDURES NUMBER:
F3RESPONSIBILITIES DURING AN ALERT, V F3-4 SITE AREA, OR GENERAL EMERGENCY REV: 28 TABLE OF CONTENTS [CONTINUED]
Section Title Page LIST OF FIGURES Figure 1 PINGP 1125 ......................................................................................... 27 Figure 2 PINGP 571 .................................................................................. .............. 46 Figure 3 PINGP 1095 ......................................................................................... 52 Figure 4 PINGP 572 ........................................................................................... 53 Figure 5 PINGP 573 ........................................................................................... 59 Figure 6 PINGP 574 ........................................................................................... 64 Figure 7 PINGP 911 ........................................................................................... 70 Figure 8 PINGP 1245 ......................................................................................... 72 Figure 9 PINGP 1237 ......................................................................................... 73 Figure 10 PINGP 1188 ......................................................................................... 74 Page 3 of 74
PRAIRIE ISLAND NUCLEAR GENERATING PLANT EMERGENCY PLAN IMPLEMENTING PROCEDURES NUMBER:
RESPONSIBILITIES DURING AN ALERT, F3-4 F 3p SITE AREA, OR GENERAL EMERGENCY REV: 28]
1.0 PURPOSE The purpose of this instruction is to delineate the responsibilities of various emergency organization personnel and onsite organizations required to respond to an Alert, a Site Area Emergency, or a General Emergency.
2.0 APPLICABILITY This instruction SHALL apply to all plant personnel.
3.0 PRECAUTIONS 3.1 All personnel should stay clear of any areas as announced over the public address system.
3.2 All personnel should refrain from using the public address system or telephone system during an emergency.
3.3 When the evacuation alarm is heard, evacuate your work area while listening to specific instructions on the plant's public address system. If you cannot hear or understand the instructions, continue to leave the immediate area until you learn of the evacuation instructions.
3.4 Anyone working in a contaminated area when the evacuation alarm sounds should remove as much protective clothing as time permits, especially gloves, booties or rubbers. If wearing a double suit, removal of outside clothing would only be necessary. Proceed to the designated assembly area. If unable to remove all protective clothing, inform personnel in charge at the assembly area of your condition.
NOTE: When the evacuation alarm sounds during a DRILL, remove ALL protective clothing prior to evacuating.
3.5 When exiting the Protective Area via the Guardhouse, proceed through the portal monitor quickly and step through without stopping. All I.D. cards (badges) should be collected and checked out by the Security Force, so an early printout of all personnel within the Protected Area can be obtained.
Page 4 of 74
PRAIRIE ISLAND NUCLEAR GENERATING PLANT EMERGENCY PLAN IMPLEMENTING PROCEDURES F3 3RESPONSIBILITIES SITE AREA, OR GENERAL EMERGENCY DURING AN ALERT, NUMBER:
F3-4 REV: 28 3.6 Everyone should remain at assembly area for monitoring and accountability checks until released by the Emergency Director or directed for reassignment for duty within the plant. Follow instructions from the Assembly Point Coordinator. When departing the site property, obey all instructions from traffic control personnel.
3.7 All entries into the Auxiliary Building should be controlled through the OSC.
Aux Building entries made for purposes of equipment operation, search and rescue, damage control, etc., should be accompanied by a Radiation Protection Specialist, or other qualified individual.
3.8 In the case of a credible security event, personnel may be asked to take cover for immediate personal protection or TSC and OSC staff may be directed to report to alternate locations different from the TSC and OSC (e.g., Plant Manager's Conference Room or EOF). See F3-31 for more security event guidance.
4.0 RESPONSIBILITIES 4.1 Overall Onsite Responsibility - Emergency Director 4.2 In Charge, Control Room - Shift Manager Technical Support Center - TSC Coordinator Operational Support Center - OSC Coordinator Assembly Point - Assembly Point Coordinator 4.3 Assistance, Control Room - Shift Supervisors
- Control Room Operators TSC - Operations Committee
- Shift Emergency Communicator
- Radiological Emergency Coordinator
- Engineering support as needed (i.e., systems experts)
OSC - Extra Operators
- Rad Survey Teams
- Maintenance Supervisors
- Chief Station Electrician and Alternates
- Additional Support as needed Page 5 of 74
PRAIRIE ISLAND NUCLEAR GENERATING PLANT EMERGENCY PLAN IMPLEMENTING PROCEDURES NUMBER:
- RESPONSIBILITIES DURING AN ALERT, F3-4 SITE AREA, OR GENERAL EMERGENCY 5.0 DISCUSSION A graded scale of response is provided for the different classes of emergencies, each requiring a specific response by emergency organization personnel for the protection of the public health and safety.
5.1 Alert 5.1.1 Definition The Alert Conditions are events which are in progress or have occurred which involve actual or potential substantial degradation of the level of safety of the plant.
Some releases of radioactive material to offsite areas are probable. Hence, there is some necessity for emergency planning and response by offsite agencies. Any radioactive release will be limited to a small fraction of the EPA Protective Action Guideline exposure levels.
5.1.2 Purpose of Alert Class The purpose of the Alert Emergency classification is to (1) assure that emergency personnel are readily available to respond if the situation becomes more serious or to perform confirmatory radiation monitoring, if required; (2) provide offsite authorities current status information.
5.1.3 Plant Actions and Responsibilities A. Promptly inform State and/or local authorities of Alert status and reason for Alert as soon as discovered.
B. Augment resources by activating onsite Technical Support Center, onsite Operational Support Center and Near-Site Emergency Operations Facility (EOF).
C. Assess and respond to the Alert condition.
D. Dispatch onsite or offsite survey teams and associated communications (if needed).
E. Provide periodic plant status updates to offsite authorities.
Page 6 of 74
PRAIRIE ISLAND NUCLEAR GENERATING PLANT EMERGENCY PLAN IMPLEMENTING PROCEDURES FRESPONSIBILITIES F DURING AN ALERT, NUMBER:
F3-4 SITE AREA, OR GENERAL EMERGENCY F. Provide periodic meteorological assessments to offsite authorities, and if any releases are occurring, dose estimates for actual releases.
G. Close out by verbal summary to offsite authorities and assess need for recovery.
or H. Escalate to a more severe class.
5.1.4 State and/or Local Offsite Authority Actions A. Provide fire or security assistance, if required.
B. Augment resources by activating Emergency Operating Centers and Emergency Alert System to standby status.
C. Alert to standby status key emergency personnel including monitoring teams and associated communications.
D. Provide confirmatory offsite radiation monitoring and ingestion pathway dose projections if actual releases substantially exceed technical specification limits.
E. Maintain alert status until verbal close-out.
or F. Escalate to a more severe class.
5.2 Site Area Emergency 5.2.1 Definition The Site Area Emergency describes events which are in progress or have occurred which involve actual or likely major failures of plant functions needed for protection of the public.
Significant offsite releases are likely to occur or are occurring, but where a core melt situation is not expected although severe fuel damage may have occurred.
Any radioactive releases are not expected to exceed the EPA Protective Action Guideline exposure levels except near the site boundary.
Page 7 of 74
PRAIRIE ISLAND NUCLEAR GENERATING PLANT EMERGENCY PLAN IMPLEMENTING PROCEDURES NUMBER:
RESPONSIBILITIES DURING AN ALERT, F3-4 F3 SREV: SITE AREA, OR GENERAL EMERGENCY 28 "
5.2.2 Purpose of Site Area Emergency Class The purpose of the Site Area Emergency classification is to:
A. Assure that response centers are manned.
B. Assure that monitoring teams are dispatched (if needed).
C. Assure that personnel required for evacuation of Near-Site areas are at duty stations if the situation becomes more serious.
D. Provide current information for and consultation with offsite authorities.
E. Provide updates for the public through offsite authorities.
5.2.3 Plant Actions and Responsibilities A. Promptly inform State and/or local offsite authorities of Site Area Emergency status and reason for emergency as soon as discovered.
B. Augment resources by activating onsite Technical Support Center, onsite Operational Support Center and the Near-Site Emergency Operations Facility (EOF).
C. Assess and respond to the Site Area Emergency.
D. If radiological or environmental conditions permit, evacuate onsite, nonessential personnel.
E. Dispatch onsite and offsite survey teams and associated communications (if needed).
F. Provide a dedicated individual for plant status updates to offsite authorities.
G. Make senior technical and management staff onsite available for consultation with NRC and State on a periodic basis.
H. Provide meteorological and dose estimates to offsite authorities for actual releases via a dedicated individual.
- 1. Provide release and dose projections based on available plant condition information and foreseeable contingencies.
Page 8 of 74
PRAIRIE ISLAND NUCLEAR GENERATING PLANT EMERGENCY PLAN IMPLEMENTING PROCEDURES F
3 RESPONSIBILITIES DURING AN ALERT, SITE AREA, OR GENERAL EMERGENCY NUMBER:
REV:
F3-4 28 J. Close out or terminate emergency class by contacting offsite authorities and assess need for recovery.
or K. Escalate to General Emergency class.
5.2.4 State and/or Local Offsite Authority Actions A. Provide any assistance requested.
B. If protective actions are desirable, activate the Public Alert &
Notification System.
C. Provide public within at least 10 miles, periodic updates on emergency status.
D. Augment resources by activating Emergency Operating Centers.
E. Dispatch key emergency personnel including monitoring teams and associated communications.
F. Alert to standby status other emergency personnel (e.g., those needed for evacuation) and dispatch personnel to Near-Site duty stations.
G. Provide offsite monitoring results to licensee and others and jointly assess them.
H. Continuously assess information from licensee and offsite monitoring with regard to changes to protective actions already initiated for public and mobilizing evacuation resources.
- 1. Recommend placing milk animals within 2 miles on stored feed and assess need to extend distance.
J. Provide press briefings, perhaps with licensee.
K. Maintain Site Area Emergency status until close-out.
or L. Escalate to General Emergency class.
Page 9 of 74
PRAIRIE ISLAND NUCLEAR GENERATING PLANT EMERGENCY PLAN IMPLEMENTING PROCEDURES NUMBER:
RESPONSIBILITIES DURING AN ALERT, F3-4 F3 SITE AREA, OR GENERAL EMERGENCY REV: 28 5.3 General Emergency 5.3.1 Definition The General Emergency describes events in progress or which have occurred which involve actual or imminent substantial core degradation or melting with the potential for loss of containment.
Radioactive releases can be reasonably expected to exceed the EPA Protective Action Guideline exposure levels offsite for more than the immediate site area. Hence, protective actions may have to be taken for protection of the general public.
5.3.2 Purpose of General Emergency Class The purpose of the General Emergency classification is to:
A. Initiate predetermined protective actions for the public.
B. Provide continuous assessment of information from licensee and offsite measurements.
C. Initiate additional measures as indicated by actual or potential releases.
D. Provide current information for the public and consultation with offsite authorities.
E. Provide updates for the public through offsite authorities.
5.3.3 Plant Actions and Responsibilities A. Promptly inform state and local offsite authorities of General Emergency status and reason for emergency as soon as discovered.
B. Augment resources by activating onsite Technical Support Center, onsite Operational Support Center and Near-Site Emergency Operations Facility (EOF).
C. Assess and respond to General Emergency.
Page 10 of 74
PRAIRIE ISLAND NUCLEAR GENERATING PLANT EMERGENCY PLAN IMPLEMENTING PROCEDURES NUMBER:
RESPONSIBILITIES DURING AN ALERT, F3-4 SITE AREA, OR GENERAL EMERGENCY R D. If radiological or environmental conditions permit, evacuate onsite, nonessential personnel.
E. Dispatch onsite and offsite survey teams and associated communications.
F. Provide a dedicated individual for plant status updates to offsite authorities.
G. Make senior technical and management staff onsite available for consultation with NRC and State on a periodic basis.
H. Provide meteorological and dose estimates to offsite authorities for actual releases via a dedicated individual.
- 1. Provide release and dose projections based on available plant condition information and foreseeable contingencies.
J. Implement the Severe Accident Management process as necessary.
K. Close out or terminate emergency class by briefing offsite authorities and transition to recovery.
5.3.4 State and/or Local Offsite Authority Actions A. Provide any assistance requested.
B. Activate immediate public notification of emergency status and provide public periodic updates.
C. Recommend evacuation for 2 mile radius and 5 miles downwind and assess need to extend distances.
D. Augment resources by activating Near-Site EOC and any other primary response centers.
E. Dispatch key emergency personnel including monitoring teams and associated communications.
F. Dispatch other emergency personnel to duty stations within 5 mile radius and alert all others to standby status.
Page 11 of 74
PRAIRIE ISLAND NUCLEAR GENERATING PLANT EMERGENCY PLAN IMPLEMENTING PROCEDURES S~NUMBER:
RESPONSIBILITIES DURING AN ALERT, F3 SITE AREA, OR GENERAL EMERGENCY REV:
F3-4 28 G. Provide offsite monitoring results to licensee and others and jointly assess these.
H. Continuously assess information from licensee and offsite monitoring with regard to changes to protective actions already initiated for public and mobilizing evacuation resources.
- 1. Recommend placing milk animals within 10 miles on stored feed and assess need to extend distance.
J. Provide press briefings, perhaps with licensee.
K. Maintain General Emergency status until close-out or termination of emergency class.
6.0 PREREQUISITES An Alert, Site Area, or General Emergency has been or will be declared.
7.0 PROCEDURE 7.1 Shift Supervisor of Affected Unit 7.1.1 Proceed to the Control Room (if not already there).
The Shift Supervisor of the affected unit SHALL remain in the NOTE: Control Room at all times during accident conditions until properly relieved.
7.1.2 Implement the appropriate Emergency Operating Procedures (EOPs) and Severe Accident Management Guidelines (SAMGs), as needed, and respond to the emergency condition with the objective of returning the plant to a normal safe condition (Mode 5, Cold Shutdown, if necessary).
NOTE: The Shift Manager and SEC will be summoned to the Control Room per the EO13s.
7.1.3 Direct activities of the Control Room Operators.
7.1.4 Coordinate, with the Emergency Director, all plant operations which may impact on radioactive releases.
Page 12 of 74
PRAIRIE ISLAND NUCLEAR GENERATING PLANT EMERGENCY PLAN IMPLEMENTING PROCEDURES SSITE 3RESPONSIBILITIES DURING AN ALERT, AREA, OR GENERAL EMERGENCY NUMBER:
F3-4 SREV: 28 7.2 Shift Supervisor of Unaffected Unit 7.2.1 Proceed to the Control Room (if not already there).
7.2.2 Direct operations on the unaffected unit.
7.2.3 Assist the Shift Manager, as necessary.
7.3 Shift Manager In the case of a credible security event, if you are unable to safely go to the Control Room, consider directing the emergency response from another safe location (e.g., SM Office, CAS or Plant NOTE: Manager's Conference Room). An informational copy of the SM/SS Emergency Director Checklist (PINGP 1125) is located in the back of F3-4, located in the SM Office, Plant Manager's Conference Room, EOF and New Admin Reference Library.
The initial E-Plan response to the event should be made by the Shift Manager. However, in order to adequately maintain oversight NOTE: of the operational aspects of the event, it may be necessary for the Shift Manager to delegate specific E-Plan duties to the unaffected unit SS during the initial response while still maintaining the ED position.
7.3.1 Report to the Control Room immediately upon notification.
7.3.2 Assess the emergency condition, event evaluation, and safety aspects of the plant.
NOTE Operator's desk to hear key communications, use an ERCS terminal for monitoring CSFSTs (Critical Safety Function Status Trees), and solicit or answer questions of the SS.
7.3.3 Temporarily assume the position of Emergency Director until relieved by the oncoming Designated Emergency Director using PINGP 1125 (Figure 1),
Shift Manager/Shift Supervisor Emergency Director Checklist, for specific guidance on emergency plan duties.
7.3.4 If the Severe Accident Management process is initiated and the TSC is functioning, the Shift Manager should report to the TSC and become a member of the Severe Accident Management Team.
Page 13 of 74
PRAIRIE ISLAND NUCLEAR GENERATING PLANT EMERGENCY PLAN IMPLEMENTING PROCEDURES RESPONSIBILITIES DURING AN ALERT, F3sSITE AREA, OR GENERAL EMERGENCY SREV:
F3-4 28j/
7.4 Designated Emergency Director (Plant Manager or ED Designee)
In the case of a credible security event, if you are unable to safely go to the Control Room or TSC, consider directing the emergency response from another safe location (e.g., Plant NOTE: Manager's Conference Room or EOF). An information copy of the Emergency Director Checklist (PINGP 571) is located
- in the back of F3-4, located in the Plant Manager's Conference Room, EOF and New Admin Reference Library.
See F3-31 for more security event guidance.
7.4.1 The Emergency Director (Plant Manager or designee) should report to Control Room and assume the role of Emergency Director from the Shift Manager/Shift Supervisor.
7.4.2 Use PINGP 571 (Figure 2), TSC Emergency Director Checklist.
7.5 Operations Group 7.5.1 General Superintendent Plant Operations or Designee A. Report to the Technical Support Center to perform the role of the Operations Group Leader.
B. Assist in the activation of onsite emergency centers and organization by ensuring adequate Operations staffing in Control Room and OSC.
C. Ensure the OSC Operations Advisor is staffed by an Operations Support Pool personnel.
D. Assess the operational aspects of the emergency.
E. Periodically review the status and implementation of the EOPs and/or Abnormal Operations procedures with the TSC Staff.
F. If the Severe Accident Management process is initiated, staff the Severe Accident Evaluation Team Leader position.
Page 14 of 74
PRAIRIE ISLAND NUCLEAR GENERATING PLANT EMERGENCY PLAN IMPLEMENTING PROCEDURES A ULR NUMBER:
F FRESPONSIBILITIES DURING AN ALERT, F3-4 SITE AREA, OR GENERAL EMERGENCY 7.5.2 Control Room Personnel A. Assist the Shift Supervisor as requested.
B. Utilize applicable Operations Manual procedures to respond to the Emergency Condition as appropriate, with the objective of returning the plant to a normal safe status (Mode 5, Cold Shutdown, if necessary).
C. Announce the location and nature of the Emergency over the public address system. See PINGP 1125 (Figure 1) for example of announcement.
D. When an evacuation is declared, sound the evacuation alarm and direct all nonessential personnel to evacuate to the designated assembly point. Direct all personnel to remain clear of the affected area (if applicable). See F3-9, Emergency Evacuation or PINGP 1125 (Figure 1) for example of announcement.
E. Continuously monitor the Control Room instrumentation, radiation monitors, or any other developments which could be indicative of further system degradation. Inform the Shift Supervisor immediately of any changes in plant status.
F. Implement Severe Accident Management strategies as directed by the TSC.
7.5.3 Auxiliary Building and BOP Operators A. Aux and BOP Operators should report to their duty station when the emergency is declared.
B. Aux and BOP Operators will continue to take direction from the Control Room during the emergency.
C. Aux Operators SHALL ensure that they have appropriate dosimetry and a dose rate indicating device for all further required operations in the Auxiliary Building.
D. Aux & BOP Operators should give consideration to terminating all nonessential plant operations (e.g., shutdown resin sluicing).
Page 15 of 74
PRAIRIE ISLAND NUCLEAR GENERATING PLANT EMERGENCY PLAN IMPLEMENTING PROCEDURES RESPONSIBILITIES DURING AN ALERT, F34 F S ITEAREA, OR GENERAL EMERGENCY REV: 281 E. When a Plant Evacuation is declared, the Aux & BOP Operators should periodically call the OSC (for exposure control purposes) and inform the OSC of outplant operational activities.
F. If the Auxiliary Building general rad levels exceed 100 mR/hr or upon recommendations from the Emergency Director or Radiation Protection Group, the Aux Operators SHALL evacuate to the OSC.
For exposure control purposes, all further Auxiliary Building entries SHALL be controlled through the OSC.
G. Perform the necessary onsite and in-plant radiation surveys as requested by the Shift Supervisor.
Page 16 of 74
PRAIRIE ISLAND NUCLEAR GENERATING PLANT EMERGENCY PLAN IMPLEMENTING PROCEDURES NUMBER:
3RESPONSIBILITIES DURING AN ALERT, F3-4 F SITE AREA, OR GENERAL EMERGENCY REV: 28 7.5.4 Relief Shift, Training Operators, and Operations Support Pool A. Relief Shift and Training Operators should proceed to the Operational Support Center for further instructions to support Operations.
B. An available Shift Supervisor should staff the OSC Coordinator position.
C. Operations Support Pool should provide assistance to the OSC Coordinator in an advisory capacity by staffing the OSC Operations Advisor position. PINGP 1095 (Figure 3), OSC Operations Advisor Checklist, should be used as a guide.
D. Operations Support Pool should assist the OSC in tracking work groups, outplant operational activities, and updating the emergency work status board.
7.6 Shift Emergency Communicator 7.6.1 Report to the Control Room immediately upon notification unless directed otherwise by Shift Manager/Emergency Director.
7.6.2 Complete the Notification Report Form (PINGP 577), and have it reviewed and approved by the Emergency Director.
NOTE: State and local authorities SHALL be notified within 15 minutes of the declaration of the emergency classification.
7.6.3 Complete the required notification of state and local authorities, and site personnel in accordance with F3-5, Emergency Notifications.
7.6.4 Notify applicable offsite authorities if conditions escalate to a more severe emergency class in accordance with F3-5, Emergency Notifications.
NOTE: Ifthe EOF has been activated, notifications of offsite agencies for an escalation or termination will be completed by EOF personnel.
7.6.5 When the emergency classification has been terminated, close-out the emergency classification, by notifying the state, local, and site personnel in accordance with F3-5, Emergency Notification.
Page 17 of 74
PRAIRIE ISLAND NUCLEAR GENERATING PLANT EMERGENCY PLAN IMPLEMENTING PROCEDURES RESPONSIBILITIES DURING AN ALERT, F3-4 SSITE AREA, OR GENERAL EMERGENCY REV: 28 7.7 Technical Support Coordinator 7.7.1 The Technical Support Center Coordinator SHALL be responsible for the general activation, operation and coordination of activities in the Technical Support Center (TSC).
7.7.2 The TSC Coordinator should report to the TSC and assume the role as TSC Coordinator. Use PINGP 573 (Figure 5), Technical Support Center Coordinator Checklist.
7.8 Operational Support Coordinator 7.8.1 The Operational Support Center Coordinator SHALL be responsible for the general activation, operation, and coordination of activities in the Operational Support Center (OSC).
7.8.2 The OSC Coordinator should report to the OSC and assume the role as OSC Coordinator. Use PINGP 574 (Figure 6), Operational Support Center Coordinator Checklist.
7.9 Assembly Point Coordinator 7.9.1 The Assembly Point Coordinator SHALL be responsible for the general operation of the assembly area.
7.9.2 The Assembly Point Coordinator should report to the Assembly Point and assume the role of Assembly Point Coordinator. Use PINGP 911 (Figure 7), Assembly Point Coordinator Checklist and F3-9, Emergency Evacuation.
7.10 Radiological Emergency Coordinator 7.10.1 The Radiological Emergency Coordinator (REC) SHALL be responsible for accident assessment, onsite and offsite.
7.10.2 The REC should report to the Technical Support Center and assume responsibility for the Radiological Emergency Coordinator position. Use Radiological Emergency Coordinator Checklist, PINGP 572 (Figure 4).
Page 18 of 74
PRAIRIE ISLAND NUCLEAR GENERATING PLANT EMERGENCY PLAN IMPLEMENTING PROCEDURES NUMBER:
RESPONSIBILITIES DURING AN ALERT, F3-4 F SITE AREA,SREV: OR GENERAL EMERGENCY 28 7.11 Radiation Protection Group 7.11.1 The Shift Radiation Protection Specialist SHALL provide assistance (e.g.,
sampling, chemistry, radio-chemistry, surveys, or dose assessment) as requested by the Emergency Director.
7.11.2 The Radiation Survey Teams should be dispatched to initiate offsite surveys as directed per F3-15 and/or F3-16.
7.11.3 All other radiation survey group members should report to the plant site for further instructions. The radiation survey group members reporting to the plant site should:
A. Proceed to the Operational Support Center and wait for further instructions, unless otherwise directed by the Emergency Director or Radiological Emergency Coordinator.
B. Supervise any checks for personnel contamination and direct decontamination at the assembly point.
C. Provide radiation protection coverage for:
- 1. Damage control and repair teams
- 2. First aid
- 3. Search and Rescue Teams
- 4. Reentry Teams D. Perform emergency sampling (air and liquid), chemistry, radio-chemistry, surveys, etc., as directed by the Emergency Director or the Radiological Emergency Coordinator.
7.11.4 One individual should assume OSC Radiation Protection Coordinator.
Use PINGP 1245 (Figure 8), OSC Radiation Protection Coordinator Checklist.
Page 19 of 74
PRAIRIE ISLAND NUCLEAR GENERATING PLANT EMERGENCY PLAN IMPLEMENTING PROCEDURES ESPONSIBILITIES DURING AN ALERT, F3-4 F SITE AREA, OR GENERAL EMERGENCY REV:
7.12 Security Group 7.12.1 The Superintendent Security, or designee, should report to the Technical Support Center (TSC). The Superintendent Security should:
A. Ensure FFD breath analysis testing is set up and performed as required for off-hours emergency call-in.
B. Consider suspending selected plant access controls to expedite the staffing of the emergency centers during an emergency. See SIP 5.2, Security Response to Site Emergencies.
C. Assume responsibility for personnel accountability following a plant evacuation and continuing accountability throughout the duration of the emergency. See Procedure F3-1 0.
D. Coordinate security control throughout the duration of the emergency situation. See Procedure F3-29.
E. Evaluate aspects concerning attempted acts of sabotage.
7.12.2 The Security Force A. The Security First or Second Lieutenant or designee SHALL perform the responsibilities of the Shift Emergency Communicator (SEC).
B. All other plant security force should continue with normal duties unless otherwise notified.
C. During off-normal work hours and when requested by the Shift Emergency Communicator (SEC), a designated Security Force Member will control the telephone switchboard in the TSC. See Procedure and F3-5.1.
D. When the evacuation alarm sounds, all Security Officers, with the exception of Roving Patrol and the SAS Operator, evacuate to the Guardhouse for further instructions.
Page 20 of 74
PRAIRIE ISLAND NUCLEAR GENERATING PLANT EMERGENCY PLAN IMPLEMENTING PROCEDURES 3RESPONSIBILITIES DURING AN ALERT, NUMBER:
F3-4 F SITE AREA, OR GENERAL EMERGENCY REV: 28 NOTE: The SAS Operator should evacuate when directed by the Emergency Director.
E. Assist with the evacuation of personnel to the designated assembly point in accordance with F3-29, "Emergency Security Procedures" and F3-9, Emergency Evacuation.
- 1. It should be necessary for personnel to exit quickly thru the portal monitor and turnstile. Collect all I.D.'s and process badges so an Employee Onsite List of personnel inside Protected Area can be obtained.
NOTE: 2. To speed evacuation from the Protected Area, it may be beneficial to open the vehicle gates and allow personnel to exit there.
- 3. The Security Force SHALL ensure that all personnel onsite, within the Protected Area, have heard the evacuation alarm.
F. Perform a check of all areas immediately surrounding the Protected Area so that all personnel are notified of the evacuation in progress.
NOTE:
NT:The owner Controlled Area will be checked when directed by the Emergency Director.
G. Control access to Protected Area per instructions from the Emergency Director.
H. Be prepared to obtain a printout for an accountability check in accordance with F3-1 0, Personnel Accountability.
- 1. Assist the Radiation Protection Group in establishing a secondary access control point when directed by the Emergency Director.
J. Station a Security Force Member, with dosimetry, at the plant entrance, if conditions permit, to control access to the plant site.
Page 21 of 74
PRAIRIE ISLAND NUCLEAR GENERATING PLANT EMERGENCY PLAN IMPLEMENTING PROCEDURES I I NUMBER:
F3 RESPONSIBILITIES DURING AN ALERT, SITE AREA, OR GENERAL EMERGENCY REV:
F3-4 28 7.13 Logistics Group The Logistics Support Group includes the Business Support Group (Administrative Services and Document Control), the Plant Services Group and Materials Management (Warehouse).
7.13.1 The Logistics Group Leader designee should report to the TSC and utilize PINGP 1188 (Figure 10), TSC Logistics Checklist.
7.13.2 The Business Support Group, and Materials Management Group should:
A. Continue with normal duties unless directed otherwise.
- 1. During off hours emergency activation, designated Warehouse personnel should report to the Operations Support Center (OSC) to provide support in retrieving emergency parts.
NOTE:
- 2. During normal hours emergency activation, designated Warehouse personnel should continue with normal duties until a Plant Evacuation occurs, at which time they should report to the OSC.
B. Immediately vacate any emergency operating center (Control Room, OSC, or TSC) when an emergency is declared.
C. Remain clear of any areas, as announced over the public address system.
D. When requested by the Shift Emergency Communicator (SEC), the office staff should transfer control of the telephone switchboard to the TSC.
The switchboard operator should report to the TSC to control the NOTE: switchboard from the TSC until relieved by an alternate communicator. See F3-5.1.
E. When the evacuation alarm sounds, proceed to the designated assembly point.
F. Follow instructions from the Assembly Point Coordinator for either reentry into the plant or departure from the site property.
Page 22 of 74
PRAIRIE ISLAND NUCLEAR GENERATING PLANT EMERGENCY PLAN IMPLEMENTING PROCEDURES 3RESPONSIBILITIES DURING AN ALERT, NUMBER:
F3-4 F SITE AREA, OR GENERAL EMERGENCY REV: 28 7.13.3 The Nuclear Plant Service Attendants should:
A. Report to the Operational Support Center to receive specific instructions from the OSC Coordinator.
B. Provide Offsite Survey Team Drivers and/or Sample Couriers for Offsite Radiation Survey Teams.
C. Provide assistance for monitoring and decontamination at the assembly point as requested.
D. Provide general support of emergency response and recovery actions, as requested.
7.14 Instrument & Control Group 7.14.1 The I&C Supervisor and Coordinators should report to the Operational Support Center (OSC) to provide support for:
A. Repair and corrective actions for instrument and control systems, and; B. Search and rescue efforts.
7.14.2 The I&C Specialists should:
A. Continue with normal duties unless directed otherwise.
B. Immediately vacate any emergency operating center (Control Room, OSC or TSC) when an emergency is declared unless directed otherwise.
C. Remain clear of any areas, as announced over the public address system.
D. When the evacuation alarm sounds, proceed to the designated assembly point unless directed to staff the OSC.
E. Follow instructions from the Assembly Point Coordinator for either reentry into the plant or departure from the site property.
Page 23 of 74
PRAIRIE ISLAND NUCLEAR GENERATING PLANT EMERGENCY PLAN IMPLEMENTING PROCEDURES F3-4 RESPONSIBILITIES DURING AN ALERT, FSITE AREA, OR GENERAL EMERGENCY REV: 28 7.15 Maintenance Group The Maintenance Group consists of all Maintenance personnel and plant Electricians.
7.15.1 The Maintenance Supervisors (Mechanical and Electrical), and designated Lead Electricians, Lead Machinists, and Lead Riggers should report to the OSC to provide support for:
A. Repair and corrective actions for mechanical and electrical systems, and; B. Search and rescue efforts.
7.15.2 All other Maintenance personnel and Electricians should:
A. Continue with normal duties unless directed otherwise.
B. Immediately vacate any emergency operating center (Control Room, OSC, or TSC) when an emergency is declared unless directed otherwise.
C. Remain clear of any areas, as announced over the public address system.
D. When the evacuation alarm sounds, proceed to the designated assembly point unless directed to staff the OSC.
E. Follow instructions from the Assembly Point Coordinator for either reentry into the plant or departure from the site property.
7.16 Engineering Group 7.16.1 All Superintendents, lead Engineers, and system experts (as requested by their supervisor) should report to the Technical Support Center (TSC). The Engineering Group should:
A. Provide technical & engineering support for plant systems.
B. Provide technical & engineering support for operating radioactive waste systems.
C. Provide technical & engineering support on core parameter analysis.
D. Provide plant parameter trending and analysis utilizing the Emergency Response Computer System (ERCS).
Page 24 of 74
PRAIRIE ISLAND NUCLEAR GENERATING PLANT EMERGENCY PLAN IMPLEMENTING PROCEDURES 3RESPONSIBILITIES DURING AN ALERT, NUMBER:
F3-4 F SITE AREA, OR GENERAL EMERGENCY REV: 28 E. Identify adverse trends and attempt to predict significant events that could adversely affect the plant or accident mitigation efforts.
F. Perform critical evaluations of "cause and effects" on failing equipment.
G. Update TSC staff of systems' status and key equipment problems or availability.
H. Provide technical support for emergency repairs and corrective action on electrical and mechanical systems.
I. Evaluate alternate systems, components or methods that may be used to restore needed capabilities or accomplish accident mitigation.
J. Develop and propose alternate electrical or fluid flow paths that would restore key functions that were lost.
K. If the Severe Accident Management process is initiated, monitor the Severe Challenge Status Trees (SCSTs) with the aid of the ERCS.
7.16.2 All other engineers (unless specifically requested to remain in the TSC) should:
A. Continue with normal duties unless directed otherwise.
B. Immediately vacate any emergency operating center (Control Room, OSC, or TSC) when an emergency is declared unless directed otherwise.
C. Remain clear of any areas, as announced over the public address system.
D. When the evacuation alarm sounds, proceed to the designated assembly point unless directed to staff the TSC.
E. Follow instructions from the Assembly Point Coordinator for either reentry into the plant or departure from the site property.
Page 25 of 74
PRAIRIE ISLAND NUCLEAR GENERATING PLANT EMERGENCY PLAN IMPLEMENTING PROCEDURES F3-4 RESPONSIBILITIES DURING AN ALERT, SSITE AREA, SREV:
OR GENERAL EMERGENCY EV-4 28 7.17 Severe Accident Management Group The Decision Maker and Evaluation Team of the Severe Accident Management (SAM) Group should report to the TSC.
7.17.1 The SAM Group Decision Maker SHALL:
A. Consult with the SAM Evaluation Team.
B. Authorize implementation of appropriate Severe Accident Management strategies as they are developed and evaluated.
7.17.2 The SAM Evaluation Team SHALL:
A. Evaluate the Diagnostics Flow Charts and Severe Challenge Status Trees.
B. Implement the use of the Severe Accident Guidelines and Severe Challenge Guidelines.
C. Recommend for authorization the implementation of Severe Accident Strategies to the SAM Group Decision Maker (Emergency Director).
D. See PINGP 1237 (Figure 9), SAM Evaluation Team Leader Checklist, for guidance.
7.17.3 The SAM Implementors who are site emergency response individuals SHALL implement the strategies as directed by the SAM Decision Maker.
7.18 Contract, Temporary Personnel and Visitors 7.18.1 Continue with normal duties unless directed otherwise.
7.18.2 Immediately vacate any emergency operating center (Control Room, OSC, or TSC) when an emergency is declared.
7.18.3 Remain clear of any areas, as announced over the public address system.
7.18.4 When the evacuation alarm sounds, proceed to the designated assembly point.
7.18.5 Follow instructions from the Assembly Point Coordinator for either reentry into the plant or departure from the site property.
Page 26 of 74
PRAIRIE ISLAND NUCLEAR GENERATING PLANT EMERGENCY PLAN IMPLEMENTING PROCEDURES NUMBER:
RESPONSIBILITIES DURING AN ALERT, F3-4 SITE AREA, OR GENERAL EMERGENCY L EXAMPLE ONLY USE CURRENT REVISION Figure 1 PINGP 1125 PINGP 1125, Rev. 15 Page 1 of 19 Document Type: 7.42L Retention: Life of Plant CONTROL ROOM SHIFT MANAGERJSHIFT SUPERVISOR EMERGENCY DIRECTOR CHECKLIST The initial E-Plan response to the event should be made by the Shift Manager.
However, in order to adequately maintain oversight of the operational aspects of the event, it may be necessary for the Shift Manager to delegate some E-Plan duties to the unaffected unit SS during the initial response while still maintaining the ED position.
The ED dutiesma be turned over to an oncoming ED designee, if necessary.
A cellular telephone exists in the TSC Emergency Locker. Use if necessary. It will i .' + , a work at TSC window and outdoors.
I For a (an): NUE, go to Section I, page 2.
7 ALERT, go to Section II, page 6.
SITE AREA EMERGENCY, go to Section III, page 11.
GENERAL EMERGENCY, go to Section IV, page 15.
J:\TEMPLATE1 125 Shift Manager-Shift Supervisor Emergency Director.dot Page 27 of 74
PRAIRIE ISLAND NUCLEAR GENERATING PLANT EMERGENCY PLAN IMPLEMENTING PROCEDURES NUMBER:
RESPONSIBILITIES DURING AN ALERT, F SITE SREV: AREA, OR GENERAL EMERGENCY F3-4 281 USE CURRENT REVISION EXAMPLE ONLY I Figure 1 PINGP 1125 PiNGP 1125, Rev. 15 Page 2 of 19
- 1. NOTIFICATION OF UNUSUAL EVENT (NUE) ACTIONS (F3-3)
Time NUE Declared INITIAL TIME In order to adequately maintain oversight of the operational aspects of t'*,/;r** lunaffected the event, Itunit maySS.
be necessary to delegate some E-Plan duties to the (1) IFsecurity threat, THEN go to F3-31 and complete actions for security related threats.
(2) A Local Evacuation of a specific area of the plant may be necessary because of hazards. A Local Evacuation may proceed as follows: local
- a. DETERMINE assembly points using the table below as general guidance.
LOCAL EVACUATION ASSEMBLY AREA FOR FROM THESE AREAS ACCOUNTABILITY Containment/Spent Fuel Pool 735' Basketball Court Aux Bldg Access Control HP Office Old Admin/Turb Bldg New Admin Lunchroom (non-outage)
New Admin Receiving Warehouse Turb Bldg (outage) or North Warehouse NPD/SBO/Office Trailers Contractor Fab Shop Warehouse #1/Const Fab Shop Guardhouse Contractor Trailers
- b. SOUND the EVACUATION ALARM.
- c. ANNOUNCE the following over the plant page.
"ATTENTION ALL PLANT PERSONNEL THERE IS A (hazard) OCCURRING IN (specify area) . ALL PERSONNEL SHOULD EVACUATE FROM THE (specify area) AND GO TO (assembly area) FOR ACCOUNTABILITY. STAY CLEAR OF (specify area) UNTIL FURTHER NOTIFIED."
- d. REPEAT the announcement.
- e. DIRECT security (4318) to conduct Personnel Accountability using F3-1 0 as guidance.
SNUE'- NUENEUE l*U& NE ""NUE N(EtUr i'r 'lUN.... S " N .U Page 28 of 74
PRAIRIE ISLAND NUCLEAR GENERATING PLANT EMERGENCY PLAN IMPLEMENTING PROCEDURES NUMBER:
RESPONSIBILITIES DURING AN ALERT, F3-4 F SITE AREA, OR GENERAL EMERGENCY ,
I EXAMPLE ONLY USE CURRENT REVISION 1 Figure 1 PINGP 1125 PINGP 1125, Rev. 15 Page 3 of 19 I. NUE ACTIONS (CONTINUED)
INITIAL TIME In the case of a credible security event, If the TSC or Turbine Floor are unsafe areas, It may be safer for the SEC to perform emergency notifications from another safe location (e.g., Control Room SAS or CAS). SEC notification forms (PINGP 577,579 & 580) are available In Control Room, SAS, TSC, CAS, and New Admin Reference Library.
(3) Ensure the SEC (pager 4427) has been summoned to the Control Room and starts the completion of the notification report form (PINGP 577).
State and local authorities SHALL be notified within 15 minutes of the declaration of the emergency class.
"÷ It Is recommended that you position yourself at the Reactor Operator's S'G1 desk to hear key communications, use an ERCS terminal for monitoring CSFSTs, and solicit or answer questions of the SS.
(4) Review and approve the notification report form PINGP 577.
In the case where the SEC or Duty Chemist Is NOT available, STEP (5)
S " - *'-'*:provides, faxing PINGP 577 to Offsite Agencies and has SAS complete
- '1 the notification.
(5) IF SEC is NOT available, THEN:
" Place PINGP 577 facedown on the fax.
"* Dial 5076 (or backup 4805).
" Press "START" button.
"* Call SAS (ext. 4323) and direct them to complete 15 minute notification using PINGP 579.
SGo to STEP (7).
(6) Direct the SEC to complete the notifications of state, local and site personnel in accordance with F3-5 and PINGP 579.
- .. NUEý NUE~ ýNU NUE rNENENUNENE
";NEKNEN Page 29 of 74
PRAIRIE ISLAND NUCLEAR GENERATING PLANT EMERGENCY PLAN IMPLEMENTING PROCEDURES F3-4 RESPONSIBILITIES DURING AN ALERT, F SITE AREA, OR GENERAL EMERGENCY REV: 281 EXAMPLE ONLY Figure 1 PINGP 1125 USE CURRENT REVISION PINGP 1125, Rev. 15 Page 4 of 19 NUE ACTIONS (CONTINUED)
INITIAL TIME (7) Announce NUE over PA System:
- '*TE... I During drills, the announcement should begin and end with
- . I"THIS IS A DRILL".
"ATTENTION ALL PLANT PERSONNEL:
A NOTIFICATION OF UNUSUAL EVENT HAS BEEN DECLARED BASED ON (Give brief description of event)
Repeat the announcement.
(8) IFstaff augmentation is desired because the event is technically challenging or THEN the following items should be completed: complex,
- a. Augment staff by directing the SEC to activate ERO pagers (for key responders) and activate the ERO automated telephone callout system (for all ERO) per the Emergency Response Organization Activation for Non-Security Events PINGP 1384.
- b. Augment staff by conducting the following plant page over the PA System:
,o Z During drills, the announcement should begin and end with I' - ,_ ' "THIS IS A DRILL."
"ATTENTION ALL PLANT PERSONNEL:
ALL MEMBERS OF THE EMERGENCY RESPONSE ORGANIZATION REPORT TO YOUR EMERGENCY DUTY STATIONS OR EMERGENCY CENTER. ALL OTHER PERSONNEL STANDBY FOR FURTHER INSTRUCTIONS."
Repeat the announcement.
- c. Complete a turnover to the designated TSC Emergency Director (ED) WHEN the oncoming ED is ready to assume the ED responsibilities.
S "' Per regulations, notification of the NRC is required to be conducted at least Immediately after the 15 minute state and local notifications and
", not later than 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> after emergency declaration. This notification should NOT be done by the SEC.
(9) Ensure the NRC is notified of the NUE. PINGP 666 SHALL be complete (to the extent possible) prior to notifying the NRC via the ENS red phone.
~N UE: INE- ONUEUE NEý-MUE`7 UEjOE UE NL0 E1E-7 5,v N EMMPU Page 30 of 74
PRAIRIE ISLAND NUCLEAR GENERATING PLANT EMERGENCY PLAN IMPLEMENTING PROCEDURES 3RESPONSIBILITIES DURING AN ALERT, NUMBER:
F3-4 F3 SITE AREA, OR GENERAL EMERGENCY REV: 28 EXAMPLE ONLY Figure 1 PINGP 1125 USE CURRENT REVISION PINGP 1125, Rev. 15 Page 5 of 19
- ~NUE*NUE IU\NU~ f4U\F*. NUEE4U _U E -N U*UE* 4I :-iNI*It-EJ
- I. NUE ACTIONS (CONTINUED)
INITIAL TIME (10) Direct the SEC to request offsite support (i.e., fire dept., police, etc.), ifnecessary, and report when the request is completed.
S_ . (11) IF the emergency condition escalates, THEN proceed to the next appropriate emergency class checklist.
(12) IF NUE is terminated, THEN:
Time NUE Terminated
- a. Ensure an appropriate PA announcement is made.
- b. Direct SEC to complete PINGP 577 for NUE termination and conduct appropriate notifications to state/local agency.
- c. Ensure the NRC is notified of the NUE termination.
- d. Review 5AWI 1.13.0, Plant Event Investigation and Recovery, and implement as appropriate.
- e. Sign and date this PINGP form on last page.
~N NU 7~.UE. E7N E NE2NQU NU UE NUE -NUE$NLE, N E -NUMR'ý Page 31 of 74
PRAIRIE ISLAND NUCLEAR GENERATING PLANT EMERGENCY PLAN IMPLEMENTING PROCEDURES PRAIRIE ISLAND NUCLEAR GENERATING I PLANT EMERGENCY PLAN IMPLEMENTING PROCEDURES I
NUMBER:
F3 RESPONSIBILITIES DURING AN ALERT, SITE AREA, OR GENERAL EMERGENCY REV:
F34 28 I
Figure 1 PINGP 1125 EXAMPLE ONLY USE CURRENT REVISION PINGP 1125, Rev. 15 I
I Page 6 of 19 II. ALERT ACTIONS (13-4)
Time ALERT Declared INITIAL TIME
. ~'
"" In order to adequately maintain oversight of the operational aspects of Sthe event, It may be necessary to delegate some E-Plan duties to the
[,.,. ~unaffected unit SS.
(1) IF security threat, go to F3-31 and complete actions for security related threats.
-THEN (2) A Local Evacuation of a specific area of the plant may be necessary because of local hazards. A Local Evacuation may proceed as follows:
- a. DETERMINE assembly points using the table below as general guidance.
LOCAL EVACUATION ASSEMBLY AREA FOR FROM THESE AREAS ACCOUNTABILITY Containment/Spent Fuel Pool 735' Basketball Court Aux Bldg Access Control HP Office Old Admin/Turb Bldg New Admin Lunchroom (non-outage)
New Admin Receiving Warehouse Turb Bldg (outage) or North Warehouse NPD/SBO/Office Trailers Contractor Fab Shop Warehouse #1/Const Fab Shop Guardhouse Contractor Trailers
- b. SOUND the EVACUATION ALARM.
- c. ANNOUNCE the following over the plant page.
"ATTENTION ALL PLANT PERSONNEL THERE IS A (hazard) OCCURRING IN (specify area) . ALL PERSONNEL SHOULD EVACUATE FROM THE (specify area) AND GO TO (assembly area) FOR ACCOUNTABILITY.
STAY CLEAR OF (specify area) UNTIL FURTHER NOTIFIED."
- d. REPEAT the announcement.
- e. DIRECT security (4318) to conduct Personnel Accountability using F3-10 as guidance.
-ALERV AE]ER AAE4R " ER *. ,
Page 32 of 74
PRAIRIE ISLAND NUCLEAR GENERATING PLANT EMERGENCY PLAN IMPLEMENTING PROCEDURES NUMBER:
RESPONSIBILITIES DURING AN ALERT, F3 SITE AREA, OR GENERAL EMERGENCY VF3-4 REV: 28 Figure 1 PINGP 1125 EXAMPLE ONLY USE CURRENT REVISION PINGP 1125, Rev. 15 Page 7 of 19 Il. ALERT ACTIONS (CONTINUED)
INITIAL TIME (3) Assume the role of Emergency Director (F3-4).
In the case of credible security event, if the TSC or Turbine Floor are unsafe areas, it may be safer for the SEC to perform emergency notifications from another safe location (e.g., Control Room SAS or CAS). SEC notification forms (PINGP 577, 579 & 580) are available in Control Room SAS, TSC, CAS, and New Admin Reference Library.
i (4) Ensure the SEC (pager 4427) has been summoned to the Control Room and starts the completion of the notification report form (PINGP 577).
State and local authorities SHALL be notified within 15 minutes of the
..w 1declaration of the ALERT.
It is recommended that you position yourself at the Reactor Operator's desk to hear key communications, use an ERCS terminal for i- monitoring CSFSTs, and solicit or answer questions of the SS.
(5) Review and approve the notification report form PINGP 577.
-. , ' In the case where the SEC or Duty Chemist Is NOT available, STEP (6) 71;.1 provides, faxing PINGP 577 to OffsIte Agencies and has SAS complete the notification.
(6) IF SEC is NOT available, THEN:
"* Place PINGP 577 facedown on the fax.
"*Dial 5076 (or backup 4805).
"*Press "START" button.
"*Call SAS (ext. 4323) and direct them to complete 15 minute notification using PINGP 580.
"* Go to STEP (8).
(7) Direct the SEC to complete the notifications of state and local agencies and activate the site Emergency Response Organization in accordance with F3-5 and PINGP 580.
-ALERTZJALERTZ..-. ALRt~rAL, Rr-Page 33 of 74
PRAIRIE ISLAND NUCLEAR GENERATING PLANT EMERGENCY PLAN IMPLEMENTING PROCEDURES NUMBER:
RESPONSIBILITIES DURING AN ALERT, F3-4 F3 SITE AREA, OR GENERAL EMERGENCY REV:
RE:2 28 I EXAMPLE ONLY USE CURRENT REVISION Figure 1 PINGP 1125 PINGP 1125, Rev. 15 Page 8 of 19 S...mT* ...
- 11. ALERT ACTIONS (CONTINUED)
INITIAL TIME (8) Announce the emergency class over PA System:
-NCrt - During drills, the announcement should begin and end with
.F "THIS IS A DRILL".
"ATTENTION ALL PLANT PERSONNEL:
AN ALERT HAS BEEN DECLARED BASED ON (Give brief description of event)
ALL MEMBERS OF THE EMERGENCY RESPONSE ORGANIZATION REPORT TO YOUR EMERGENCY DUTY STATIONS OR EMERGENCY CENTER. ALL OTHER PERSONNEL STANDBY FOR FURTHER INSTRUCTIONS."
Repeat the announcement.
Per regulations notification of the NRC Is required to be conducteda rE least immediately after the 15 minute state and local notifications and
, j not later than 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> after emergency declaration.
This notification should NOT be done by the SEC.
(9) Ensure the NRC is notified of the ALERT. PINGP 666 SHALL be completed (to the extent possible) prior to notifying the NRC via the ENS red phone.
ALERTY"'- WArRT, LE " --
Page 34 of 74
PRAIRIE ISLAND NUCLEAR GENERATING PLANT EMERGENCY PLAN IMPLEMENTING PROCEDURES F3 3RESPONSIBILITIES SITE AREA, OR GENERAL EMERGENCY DURING AN ALERT, NUMBER:
F3-4 SREV: 28 EXAMPLE ONLY Figure 1 PINGP 1125 USE CURRENT REVISION PINGP 1125, Rev. 15 Page 9 of 19 II. ALERT ACTIONS (CONTINUED)
INITIAL TIME 41 A Plant Evacuation is NOT required at an ALERT. However, a Plant
- OTE*Ž,* *Evacuation may be advantageous for personnel accountability Sreasons.
(10) Determine need to initiate a Plant Evacuation (F3-9). IF a Plant Evacuation is necessary, THEN:
- a. Designate assembly point (F3-9, North Warehouse or Receiving Warehouse).
"* May use North Warehouse IF wind is from:
2360 to 3600 or 00 to 1230
"* May use Receiving Warehouse IF wind is from:
1230 to 3600 or 00 to 340
- b. Ensure radiological monitoring of assembly area and guardhouse.
- c. Sound the evacuation siren (5 seconds).
- d. Announce the following on the PA system:
During drills, the announcement should begin and end with
"-THIS IS A DRILL".
"ATTENTION ALL PLANT PERSONNEL. A PLANT EVACUATION HAS BEEN DECLARED. ALL EMERGENCY ORGANIZATION PERSONNEL REPORT TO AND REMAIN AT YOUR EMERGENCY DUTY STATIONS. ALL OTHER PERSONNEL SHALL EVACUATE TO THE "
(specify assembly point)
Repeat the announcement.
(11) Monitor Aux Building habitability and direct evacuation of Aux Operators to OSC IF:
- a. General area rad levels >100 mrlhr; or
- b. Based on Rad Protection Group recommendations.
-L~t. AETALLR "ALIERT'ý-AL-RFt- ýALRA -PS'ŽAER~LET- LET Page 35 of 74
PRAIRIE ISLAND NUCLEAR GENERATING PLANT EMERGENCY PLAN IMPLEMENTING PROCEDURES NUMBER:
R ESPONSIBILITIES DURING AN ALERT, F3 SITE AREA, OR GENERAL EMERGENCY SREV:
F73-4 28 EXAMPLE ONLY I Figure 1 PINGP 1125 USE CURRENT REVISION I I Page PINGPlOof 1125, 19Rev. 15 lk 01, fAL-"WERTE-FA~ft II. ALERT ACTIONS (CONTINUED)
INITIAL TIME (12) Ensure:
- a. Onsite sampling is considered IF a fission product boundary has failed (F3-23).
- b. Dose calculations are performed by TSC IF there is a known radioactive release.
(13) Authorize emergency overexposure (F3-12), if necessary, UNTIL the oncoming designated TSC Emergency Director has taken over this responsibility. This authority may NOT be delegated.
(14) Escalate the emergency class, as appropriate (F3-2).
(15) Complete a turnover to the designated TSC Emergency Director (ED) WHEN the oncoming ED is ready to assume the ED responsibilities.
(16) Sign and date this PINGP form on the last page.
L~RTrAERT, ALE~r4.-AtlkrtýALEIRT '-AXET Page 36 of 74
PRAIRIE ISLAND NUCLEAR GENERATING PLANT EMERGENCY PLAN IMPLEMENTING PROCEDURES NUMBER:
RESPONSIBILITIES DURING AN ALERT, F3-4 SITE AREA, OR GENERAL EMERGENCY Figure 1 PINGP 1125 EXAMPLE ONLY USE CURRENT REVISION PINGP 1125, Rev. 15 1 1 Page 11 of 19 III. SITE AREA EMERGENCY ACTIONS (F3-4)
Time SITE AREA Declared INITIAL TIME In order to adequately maintain oversight of the operational aspects of the event, It may be necessary to delegate some E-Plan duties to the i,!r. unaffected unit SS.
(1) Assume the role of Emergency Director (F3-4).
In the case of a credible security event, if the TSC or Turbine Floor are I unsafe areas, it may be safer for the SEC to perform emergency notifications from another safe location (e.g., Control Room SAS or CAS). SEC notification forms (PINGP 577, 579 & 580) are available In
(2) Ensure the SEC (pager 4427) has been summoned and starts the completion of the notification report form (PINGP 577).
I N* 6-4 State and local authorities SHALL be notified within 15 minutes of the declaration of an emergency.
1; c It Is recommended that you position yourself at the Reactor Operator's desk to hear key communications, use an ERCS terminal for monitoring CSFSTs, and solicit or answer questions of the SS.
(3) Review and approve the notification report form PINGP 577.
In the case where the SEC or Duty Chemist is NOT available, STEP (4) provides, faxing PINGP 577 to Offsite Agencies and has SAS complete the notification.
(4) IF SEC is NOT available, THEN:
"* Place PINGP 577 facedown on the fax.
"* Dial 5076 (or backup 4805).
"* Press "START* button.
"* Call SAS (ext. 4323) and direct them to complete 15 minute notification using PINGP 580.
"* Go to STEP (6).
(5) Direct the SEC to complete the notifications of state and local agencies and, IF not already performed, activate the site Emergency Response Organization in accordance with F3-5 and PINGP 580.
ZIWWAREtA STEA E"TZR srr1 RE SE0ARASfEAR Page 37 of 74
PRAIRIE ISLAND NUCLEAR GENERATING PLANT EMERGENCY PLAN IMPLEMENTING PROCEDURES RESPONSIBILITIES DURING AN ALERT, F3 SITE AREA, OR GENERAL EMERGENCY R:
F3-4 28/
SREV:
E EXAMPLEM E ONLY NFigure 1 PINGP 1125 USE CURRENT REVISION PINGP 1125, Rev. 15 Page 12 of 19 III. SITE AREA EMERGENCY ACTIONS (CONTINUED)
INITIAL TIME (6) Announce the emergency class over PA System:
= During drills, the announcement should begin and end with "THIS IS A DRILL".
"ATTENTION ALL PLANT PERSONNEL:
A SITE AREA EMERGENCY HAS BEEN DECLARED BASED ON (Give brief description of event)
ALL MEMBERS OF THE EMERGENCY RESPONSE ORGANIZATION REPORT TO YOUR EMERGENCY DUTY STATIONS OR EMERGENCY CENTER. ALL OTHER PERSONNEL STANDBY FOR FURTHER INSTRUCTIONS."
Repeat the announcement.
,Per regulations notification of the NRCis required to be conducted at least Immediately after the 15 minute state and local notifications and not later than 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> after emergency declaration. This notification should NOT be done by the SEC.
(7) Ensure the NRC is notified of the emergency. PINGP 666 SHALL be completed (to the extent possible) prior to notifying the NRC via the ENS red phone.
~SI
. SITE
. AR
........ .. ........ ........ ....... - . ..... RR- . S AR EA ;
2TE Page 38 of 74
PRAIRIE ISLAND NUCLEAR GENERATING PLANT EMERGENCY PLAN IMPLEMENTING PROCEDURES NUMBER:
RESPONSIBILITIES DURING AN ALERT, F3-4 F3 SITE AREA, OR GENERAL EMERGENCY SREV: 28 EXAMPLE ONLY Figure 1 PINGP 1125 USE CURRENT REVISION PINGP 1125, Rev. 15 1 1 Page 13 of 19 Ill. SITE AREA EMERGENCY ACTIONS (CONTINUED)
INITIAL TIME
[T TA Plant Evacuation Is required at a Site Area Emergency unless hazardous conditions prohibit it. (F3-9).
A Site Evacuation is required when the normal onsite assembly areas (North Warehouse or Receiving Warehouse) are not habitable.
(8) IF initiating a Plant Evacuation, THEN perform:
- a. Designate assembly point as North Warehouse or Receiving Warehouse (F3-9).
", May use North Warehouse IF wind is from:
2360 to 3600 or 00 to 1230
"* May use Receiving Warehouse IF wind is from:
1230 to 3600 or 00 to 340
- b. Ensure radiological monitoring of assembly area and guardhouse.
- c. Sound the evacuation siren (5 seconds).
- d. Announce the following on the PA system:
During drills, the announcement should begin and end with "THIS IS A DRILL".
"ATTENTION ALL PLANT PERSONNEL. A PLANT EVACUATION HAS BEEN DECLARED. ALL EMERGENCY ORGANIZATION PERSONNEL REPORT TO AND REMAIN AT YOUR EMERGENCY DUTY STATIONS. ALL OTHER PERSONNEL SHALL EVACUATE TO THE (specify assembly point)
Repeat the announcement.
,0 W-;AREAk IT1 SA Sh SIThEARE"
ýEAR AR *A' tAlk S]T R SA Page 39 of 74
PRAIRIE ISLAND NUCLEAR GENERATING PLANT EMERGENCY PLAN IMPLEMENTING PROCEDURES ESPONSIBILITIES DURING AN ALERT, F3-4 F SREV: SITE I
AREA, OR GENERAL EMERGENCY
~ NUMBER:
R 28l Figure 1 PINGP 1125 EXAMPLE ONLY USE CURRENT REVISION PINGP 1125, Rev. 15 Page 14 of 19 Il. SITE AREA EMERGENCY ACTIONS (CONTINUED)
INITIAL TIME (9) IFinitiating a Site Evacuation, THEN perform:
- a. Designate assembly point as the training center unless advised otherwise by REC or Emergency Manager (F3-9).
_... ._ b. Ensure radiological monitoring of assembly area.
____c. Sound the evacuation siren (5 seconds).
- d. Announce the following on the PA system:
- *
- During drills, the announcement should begin and end with l*;*,,*,4; "THIS IS A DRILL".
"ATrENTION ALL PLANT PERSONNEL. PERSONNEL WITHOUT EMERGENCY ASSIGNMENTS SHALL EVACUATE THE PLANT SITE IMMEDIATELY. GET YOUR CAR KEYS AND EVACUATE TO THE PARKING LOT. USE YOUR CAR OR NSP VEHICLE AND PROCEED TO (specific assembly point)
ALL EMERGENCY PERSONNEL SHOULD REMAIN AT YOUR EMERGENCY OPERATING CENTERS."
Repeat the announcement.
(10) Monitor Aux Building habitability and direct evacuation of Aux Operators to OSC IF:
- a. General area rad levels >100 mr/hr; or
- b. Based on Rad Protection Group recommendations.
(11) Ensure:
- a. Onsite sampling is considered IF a fission product boundary has failed (F3-23).
- b. Dose calculations are performed by Shift Chemist or TSC IF R-50 is in valid alarm and such action will NOT prevent the completion of any other critical actions needed to mitigate the event.
(12) IFthe MIDAS 4-day Integrated Dose projection results exceed 1000 mRem TEDE or 5000 mRem Thyroid CDE (F3-8.1), THEN re-evaluate the emergency classification and reclassify to a General Emergency ifappropriate. (TEDE = Total Effective Dose Equivalent, CDE = Committed Dose Equivalent)
(13) Authorize emergency overexposure (F3-12), ifnecessary, UNTIL the oncoming designated TSC Emergency Director has taken over this responsibility. This authority may NOT be delegated.
(14) Escalate the emergency class, as appropriate (F3-2).
(15) Complete a turnover to the designated TSC Emergency Director (ED) WHEN the oncoming ED is ready to assume the ED responsibilities.
(16) Sign and date this PINGP form on the last page.
.. F AE
.... ...... E.A.h fS1AsS
,*s!EA Page 40 of 74
PRAIRIE ISLAND NUCLEAR GENERATING PLANT EMERGENCY PLAN IMPLEMENTING PROCEDURES NUMBER:
RESPONSIBILITIES DURING AN ALERT, F3-4 F SITE AREA, OR GENERAL EMERGENCY EXAMPLE ONLY Figure 1 PINGP 1125 USE CURRENT REVISION PINGP 1125, Rev. 15 Page 15 of 19 IV. GENERAL EMERGENCY ACTIONS (F3-4)
Time GENERAL Declared INITIAL TIME KItC
-4. .4 In order to adequately maintain oversight of the operational aspects of the event, it may be necessary to delegate some E-Plan duties to the
[I unaffected unit SS.
(1) Assume the role of Emergency Director (F3-4).
In the case of a security event, ifthe TSC or Turbine Floor are unsafe it may be safer for the SEC to perform emergency notifications from another safe location (e.g., Control Room SAS or CAS). SEC
- notification forms (PINGP 577, 579 &580) are available in Control r- 1
(2) Ensure the SEC (pager 4427) has been summoned and starts the completion of the notification report form (PINGP 577).
. State and local authorities SHALL be notified within 15 minutes of the declaration of an emergency.
it Is recommended that you position yourself at the Reactor Operator's
~r~Y~Idesk 1 to hear key communications, use an ERCS terminal for Ii1 monitoring CSFSTs, and solicit or answer questions of the SS.
(3) Recommend evacuation for the general public on PINGP 577 as designated in Figure 1.
F3-8.1.
" IFwind > 5 mph, THEN evacuate all sectors out to 2 miles AND the 5 downwind sectors out to 5 miles; AND advise remainder of plume EPZ to monitor radio/TV broadcasts for further emergency information.
"*IFwind < 5 mph, THEN evacuate all sectors out to 5 miles AND advise remainder of plume EPZ to monitor radio/TV broadcasts for further emergency information.
(4) Review and approve the notification report form PINGP 577.
-ii-l In the case where the SEC or Duty Chemist Is NOT available, STEP(5) provides, faxing PINGP 577 to Offsite Agencies and has SAS complete the notification.
(5) IF SEC is NOT available, THEN:
"* Place PING P 577 facedown on the fax.
"* Dial 5076 (or backup 4805).
"* Press "START" button.
"* Call SAS (ext. 4323) and direct them to complete 15 minute notification using PINGP 580.
"o Go to STEP (7).
(6) Direct the SEC to complete the notifications of state and local agencies and, IFnot already performed, activate the site Emergency Response Organization inaccordance with F3-5 and PINGP 580 909FNEALK 'GNERAtL6,,G.EN4EWALYG 4 mA MZEO Fý Page 41 of 74
PRAIRIE ISLAND NUCLEAR GENERATING PLANT EMERGENCY PLAN IMPLEMENTING PROCEDURES PR II IS A D N C E R G N R TNIL N M RG N Y P A M L M N I GP O E U E NUMBER:
RESPONSIBILITIES DURING AN ALERT, F3 SITE AREA, OR GENERAL EMERGENCY REV:
F3-4 281 I EXAMPLE ONLY USE CURRENT REVISION I Figure 1 PINGP 1125 PINGP 1125, Rev. 15 Page 16 of 19 IV. GENERAL EMERGENCY ACTIONS (CONTINUED)
INITIAL TIME (7) Announce the emergency class over PA System:
- " "TH~sDuring drills, the announcement should begin and end with
~ "THIS IS A DRILL".
"ATTENTION ALL PLANT PERSONNEL:
A GENERAL EMERGENCY HAS BEEN DECLARED BASED ON (Give brief description of event)
ALL MEMBERS OF THE EMERGENCY RESPONSE ORGANIZATION REPORT TO YOUR EMERGENCY DUTY STATIONS OR EMERGENCY CENTER. ALL OTHER PERSONNEL STANDBY FOR FURTHER INSTRUCTIONS."
Repeat the announcement.
4Per regulations notification of the NRC Is required to be conducted at least Immediately after the 15 minute state and local notifications and not later than 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> after emergency declaration. This notification should NOT be done by the SEC.
(8) Ensure the NRC is notified of the emergency. PINGP 666 SHALL be completed (to the extent possible) prior to notifying the NRC via the ENS red phone.
5GENERtW GENEWBAC'I ENXCýA, E G0ERM'-0E EA .
Page 42 of 74
PRAIRIE ISLAND NUCLEAR GENERATING PLANT EMERGENCY PLAN IMPLEMENTING PROCEDURES NUMBER:
RESPONSIBILITIES DURING AN ALERT, F3-4 F SITE AREA, OR GENERAL EMERGENCY I EXAMPLE ONLY Figure 1 PINGP 1125 USE CURRENT REVISION PINGP 1125, Rev. 15 1 1 Page 17 of 19
, G!.*.
+E F A.
- ... *ii...... * ,,*,G. ......
E~ :- * .** ........
- .I**N,
..+.. ....... ...... .... *-.
- IV. GENERAL EMERGENCY ACTIONS (CONTINUED)
INITIAL TIME A Plant Evacuation Is required at a General Emergency unless hazardous conditions prohibit it (13-9) or It has already been completed.
A Site Evacuation is required when the normal onsite assembly areas
"" -*"* (North Warehouse or Receiving Warehouse) are not habitable.
(9) if initiating a Plant Evacuation, THEN perform:
- a. Designate assembly point as North Warehouse or Receiving Warehouse (F3-9).
"* May use North Warehouse IF wind is from:
2368 to 360° or 00 to 1230
"* May use Receiving Warehouse IF wind is from:
1230 to 3600 or 0Qto 340
- b. Ensure radiological monitoring of assembly area and guardhouse.
- c. Sound the evacuation siren (5 seconds).
- d. Announce the following on the PA system:
During drills, the announcement should begin and end with "THIS IS A DRILL".
"ATTENTION ALL PLANT PERSONNEL A PLANT EVACUATION HAS BEEN DECLARED. ALL EMERGENCY ORGANIZATION PERSONNEL REPORT TO AND REMAIN AT YOUR EMERGENCY DUTY STATIONS. ALL OTHER PERSONNEL SHALL EVACUATE TO THE (specific assembly point)
Repeat the announcement.
ý GENERAWGENEMALGEN ERAL*GENERA GNERAE , E , N Page 43 of 74
PRAIRIE ISLAND NUCLEAR GENERATING PLANT EMERGENCY PLAN IMPLEMENTING PROCEDURES NUMBER:
RESPONSIBILITIES DURING AN ALERT, F SITE AREA, OR GENERAL EMERGENCY REV:
F3-4 28 EXAMPLE ONLY Figure 1 PINGP 1125 USE CURRENT REVISION PINGP 1-125, Rev. 15 Page 18 of 19 S.... ... ....... L*. .E.,* GENE*AGEr* AL IV. GENERAL EMERGENCY ACTIONS (CONTINUED)
INITIAL TIME (10) LF initiating a Site Evacuation, THEN perform:
- a. Designate assembly point as the training center unless advised otherwise by REC or Emergency Manager (F3-9).
_
- __ b. Ensure radiological monitoring of assembly area.
- c. Sound the evacuation siren (5 seconds).
- d. Announce the following on the PA system:
, , During drills, the
'r"* *=* ;/" "THIS IS A DRILL".announcement should begin and end with
.-. ;* -N1.:/
"ATTENTION ALL PLANT PERSONNEL. PERSONNEL WITHOUT EMERGENCY ASSIGNMENTS SHALL EVACUATE THE PLANT SITE IMMEDIATELY. GET YOUR CAR KEYS AND EVACUATE TO THE PARKING LOT. USE YOUR CAR OR NSP VEHICLE AND PROCEED TO (specific assembly point)
ALL EMERGENCY PERSONNEL SHOULD REMAIN AT YOUR EMERGENCY OPERATING CENTERS."
Repeat the announcement.
(11) Monitor Aux Building habitability and direct evacuation of Aux Operators to OSC IF:
- a. General area rad levels >100 mr/hr; or
- b. Based on Rad Protection Group recommendations.
(12) Ensure:
- a. Onsite sampling is considered IF a fission product boundary has failed (F3-23).
- b. Dose calculations are performed by Shift Chemist or TSC IF R-50 is in valid alarm and such action will NOT prevent the completion of any other critical actions needed to mitigate the event.
(13) Authorize offsite protective action recommendations changes (F3-8.1), if necessary, UNTIL the oncoming designated TSC Emergency Director has taken over this responsibility. This authority may NOT be delegated.
(14) Authorize emergency overexposure (F3-12), it necessary, UNTIL the oncoming designated TSC Emergency Director has taken over this responsibility. This authority may NOT be delegated.
(15) Complete a turnover to the designated TSC Emergency Director WHEN the oncoming ED is ready to assume the ED responsibilities.
(16) Sign and date this PINGP form on last page.
j,,"-,-EN9A GE~4r=FAL ER ~ E EIL R ENERIAEý-IIE -.0A.L-A Page 44 of 74
PRAIRIE ISLAND NUCLEAR GENERATING PLANT EMERGENCY PLAN IMPLEMENTING PROCEDURES NUMBER:
RESPONSIBILITIES DURING AN ALERT, F3-4 SITE AREA, OR GENERAL EMERGENCY Figure 1 PINGP 1125 EXAMPLE ONLY USE CURRENT REVISION I PINGP 1125, Rev. 15 Page 19 of 19 Sign AFTER NUE termination or turnover to oncoming Emergency Director.
Shift Manager/Shift Supervisor E.D. Date/Time Designated TSC Emergency Director Date/Time Page 45 of 74
PRAIRIE ISLAND NUCLEAR GENERATING PLANT EMERGENCY PLAN IMPLEMENTING PROCEDURES PR II D N C E R ISA G N R TN L N M RG N Y P A M L M N I GP O E U E RESPONSIBILITIES DURING AN ALERT, F3 S...REV:SITE AREA, OR GENERAL EMERGENCY-:
F3-4 28 Figure 2 PINGP 571 EXAMPLE ONLY PINGP 571, Rev 25 USE CURRENT REVISION Page 1 of 6 TSC EMERGENCY DIRECTOR CHECKLIST Retention: Life Document Type: 7.361
- 1. Determine Plant Status.
-1.1. Report to Control Room and become familiar with the sequence of events.
__ 1.2. Get copy of completed PINGP 1125, SM/ED Check List, and review the activities that have been completed.
__ 1.3. Note the items that have not been completed and ensure they are eventually addressed.
-2. IFthis is a security event, THEN ensure the guidelines given in F3-31 are considered.
- 3. IFthe event is an NUE which is technically challenging or complex AND you have decided to activate the ERO, THEN
-3.1. Direct the SEC to activate ERO pagers (for key responders) and activate the ERO automated telephone callout system (for all ERO) per the Emergency Response Organization Activation for Non Security Events PINGP 1384.
- 3.2. Direct the conducting of the following plant page over the PA System:
,, During drills the announcement should begin and end with "THIS IS A DRILL."
"ATTENTION ALL PLANT PERSONNEL:
ALL MEMBERS OF THE EMERGENCY RESPONSE ORGANIZATION REPORT TO YOUR EMERGENCY DUTY STATIONS OR EMERGENCY CENTER. ALL OTHER PERSONNEL STANDBY FOR FURTHER INSTRUCTIONS."
Repeat the announcement.
-4. AFTER you and the SM/ED have agreed to transfer Emergency Director responsibilities, THEN announce you are assuming Emergency Director responsibilities to the Shift Manager, Unit 1 Supervisor and Unit 2 Supervisor.
COMMENTS:
SM/ED Time TSC Emergency Director Time JATEMPLATE\0571 TSC Emergency Director Checklist dot Page 46 of 74
PRAIRIE ISLAND NUCLEAR GENERATING PLANT EMERGENCY PLAN IMPLEMENTING PROCEDURES FRESPONSIBILITIES DURING AN ALERT, NUMBER:
F3-4 SSITE AREA, OR GENERAL EMERGENCY F3-4 28 SREV:
Figure 2 - PINGP 571 EXAMPLE ONLY USE CURRENT REVISION PINGP 571, Rev 25 Page 2 of 6 TSC EMERGENCY DIRECTOR CHECKLIST
- 1. Sign-in on the TSC Personnel Status Board. (The TSC Coordinator will inform you when the TSC is operational.)
- 2. Update TSC personnel of plant status and declare TSC operational.
__ 2.1. Before you start your expectations announcement:
"* Review copy of Emergency Notification Report Form (PINGP 577) (TSC Coord. should provide.)
"* Put on ED hat.
- Test the microphone.
- Stand.
- Direct TSC Bell to be rung.
- 2.2. Make the following announcement (example of items to discuss):
"Attention all TSC workers. (Wait for all to be quiet and repeat announcement ifnecessary.)
I am (given name) and I am the Emergency Director.
I want all to listen and follow my expectations so we can have clear, concise integrated communications.
The TSC has been determined to be operational.
My expectations for TSC operation are:
- All communications should follow the plant's communications standard using 3-way communication and the phonetic alphabet.
, All personnel SHALL stop your conversations and listen to plant page announcements.
- All personnel SHALL stop your conversations and listen to facility updates.
- If someone is assigned a job or task, acknowledge the task you are going to do (use 3-Way Communication).
- Communicate only that information that should be heard by all.
- Do not repeat information from previous updates unless it is necessary.
- Let us attempt to keep facility updates to less than 8 minutes.
Synchronize your watches with the TSC wall clock now.
All personnel are to check in with the OSC before working in the field.
The present situation is:
Does anyone have anything to add?"
"Our current priorities are:
All personnel who are NOT assigned a specific TSC task please move to the TSC Lower Level or I&C Conference Room. Let you supervisor know where you are.
The next TSC Update will occur at hours. (TSC Coordinator will notify OSC &
EOF of next Facility Update times.)
This is the end of the TSC update."
Page 47 of 74
PRAIRIE ISLAND NUCLEAR GENERATING PLANT EMERGENCY PLAN IMPLEMENTING PROCEDURES RESPONSIBILITIES DURING AN ALERT, NUMBER:
F - SITE AREA, OR GENERAL EMERGENCY REV: 28 PIur 571Re251 R ENEXAM Pagere 32-PNGP o 6 T7C USE CURRENT PLE O REVISION NLYS Page 3 of 6 TSC EMERGENCY DIRECTOR CHECKLIST
-3. Direct the Event Evaluator to assess plant conditions for possible reclassifications.
-A Plant Evacuation may be Initiated during an Alert and SHALL be Initiated during ite Area or General Emergency unes radlological or environmental condItions prohibitdIt. aConsider early release of non-essentl-' personnel at the Alert.
- 4. Determine need or advantages to conduct early release or dismissal of nonessential personnel (F3-9).
-5. IF the event was initiated by a worker's performance, THEN consider Fitness for Duty "for cause testing".
.. 6. Establish communications with EOF to assess their readiness to assume offsite responsibilities (offsite communications, PAR formulation, offsite field team control and dose assessment).
__6.1. Establish a transfer time with EM.
-6.2. Transfer offsite responsibilities to EOF and announce the transfer in TSC.
-7. Determine when long-term staffing requirements should be addressed. (TSC Coordinator will establish 24-hour coverage when directed.)
- 8. Update OSC Coordinator (x4424) and Emergency Manager (x5242) of plant priorities after Facility Updates.
Emergency Director major responsibilities:
I1. Overall Command & Control of plant emergency response.
2.r AntNcip ant emergency reclassifications.
37.
__~ ~ ~~~3 SDfectyfthpu IthswsaNUcasetictiongDiec andwork
. priorties onallatigjo ar emergency clssoiatonkTE e.nsure thesEmer gencyioiicto eposForm beingP7)i cpatdofmpetgednby stheRCandbrugtsoo foryor fo5yeviw reiw u a ndapr dapoveov C ,al. C nan etfsi n r d c o n al s 10 IF lant6i termintin the ematerge nncy,a~jcnsde thrernes otasiint.eoer SeF.
- 11. IF plant is escalating 4__ the emergency classification, THEN proceed to the SITE AREA or GENERAL emergency section of this checklist.
,9. IF this.was an classificat
.. a n lt an Alerte cy classificatio THEN Page 48 of 74
PRAIRIE ISLAND NUCLEAR GENERATING PLANT EMERGENCY PLAN IMPLEMENTING PROCEDURES NUMBER:
RESPONSIBILITIES DURING AN ALERT, F3-4 F SITE SREV:AREA, OR GENERAL EMERGENCY 28 Piur NFigure 2-571Re251 PINGP 2-PNGP 571 CURRENT ONLY USE EXAMPLE REVISION Page 4 of 6 TSC EMERGENCY DIRECTOR CHECKLIST 1__
. IF EOF has not taken offsite responsibilities, THEN ensure the Notification Report Form (PINGP 577) is completed by the REC and brought to you for your review and approval, OTHERWISE immediately direct EM to escalate to a Site Area Emergency.
A Plant Evacuation SHALL be initiated during a Site Area or General Emergency UNLESS 00 radiological or environmental conditions prohibit it. IF and Plant Evacuation cannot be conducted, THEN a Site Evacuation should be conducted. (F3-9)
__ 2. Direct Plant Evacuation activities (F3-9):
__2.1. Direct REC to proclaim the appropriate Assembly Point.
__ 2.2. Direct Operations Group Leader to ensure Control Room Plant Evacuation siren is activated and plant page is conducted using F3-9 and designated Assembly Point.
__ 2.3. Direct Security Group Leader to:
"* Ensure Plant Evacuation and Accountability activities are conducted by security.
"* Complete accountability within 30 minutes after evacuation plant page (F3-10).
"* Ensure personnel outside the plant protected areas have evacuated to the Assembly Point.
__ 3. Update OSC Coordinator (x4424) and Emergency Manager (x5242) of plant priorities after Facility Updates.
Emergency Director major responsibilities:
- 1. Overall Command & Control of plant emergency response.
- 2. Anticipant emergency reclassifications.
- 3. Direct setting of work priorities on all major emergency work.
- 4. Ensure Offslte Agencies are being updated of emergency status.
- 5. Core, RCS, and Containment fission product boundaries.
- 6. Safety of plant personnel and workers.
S 7. Safety of the public.
-4. IFplant is escalating to a General Emergency, THEN proceed to the GENERAL emergency section of this checklist
- 5. iF plant is terminating the emergency, THEN advise EM and proceed to the recovery procedure, F3-30.
Page 49 of 74
PRAIRIE ISLAND NUCLEAR GENERATING PLANT EMERGENCY PLAN IMPLEMENTING PROCEDUJRESR PRAIRIE ISLAND NUCLEAR GENERATING PLANT EMERGENCY PLAN IMPLEMENTING PROCEDUiFs RESPONSIBILITIES DURING AN ALERT, F3 SITE AREA, OR GENERAL EMERGENCY REV:
F3-4 28 Figure 2 - PINGP 571 EXAMPLE ONLY PINGP 571, Rev 25 USE CURRENT REVISION F 5 Page 5 of 6 TSC EMERGENCY DIRECTOR CHECKLIST
- 1. Initial PARs for a General Emergency is evacuate 2 mile radius and five miles downwind In the affected sectors.
- 2. The authority to approve PARs remains with ED until the EM assumes SMEthe ED. This authority may not be delegated. this authority from
-1. IF EOF has not taken offsite responsibilities, THEN ensure the Notification Report completed by the REC and brought to you for your review and approval, OTHERWISE Form (PINGP 577) is immediately direct EM to escalate to a General Emergency and issue PARs.
A Plant Evacuation SHALL be Initiated during a Site Area or General Emergency UNLESS radIological or environmental conditions prohibit It. IFand Plant Evacuation cannot be
-conductd, THEN a Site Evacuation should be conducted. (F3-9)
-2. IFnot already conducted, THEN direct Plant Evacuation activities (F3-9):
-2.1. Direct REC to proclaim the appropriate Assembly Point.
- 2.2. Direct Operations Group Leader to ensure Control Room Plant Evacuation siren is activated and plant page is conducted using F3-9 and designated Assembly Point.
-2.3. Direct Security Group Leader to:
- Ensure Plant Evacuation and Accountability activities are conducted by security.
- Complete accountability within 30 minutes after evacuation plant page (F3-10).
- Ensure personnel outside the plant protected areas have evacuated to the Assembly Point.
-3. Update OSC Coordinator (x4424) and Emergency Manager (x5242) of plant priorities after Facility Updates.
Emergency Director major responsibilities:
- 1. Overall Command &Control of plant emergency response.
- 2. Anticipant emergency reclassifications.
- 3. Direct setting of work priorities on all major emergency work.
- 4. Ensure Offslte Agencies are being updated of emergency status.
- 5. Core, RCS, and Containment fission product boundaries.
- 6. Safety of plant personnel and workers.
- 7. Safety of the public.
-4. Implement TSC SAM process, as necessary.
-4.1. Direct the Operation Group Leader to ensure the Accident Evaluation Team Leader is staffed.
-4.2. Direct TSC Coordinator to ensure NRC is notified of transition to SAM.
-4.3. Address 10CFR 50.54(x) &(y) and 50.72 requirements, as necessary.
- 5. Terminate the emergency class, when appropriate. Advise EM, as appropriate.
(F3-2: classifications or F3-30: termination and recovery).
Page 50 of 74
PRAIRIE ISLAND NUCLEAR GENERATING PLANT EMERGENCY PLAN IMPLEMENTING PROCEDURES F3 3TRESPONSIBILITIES DURING AN ALERT, NUMBER:
F3-4 SITE AREA, OR GENERAL EMERGENCYRE: 2 Figure 2 - PINGP 571 EXAMPLE ONLY USE CURRENT REVISION PINGP 571, Rev 25 Page 6 of 6 TSC EMERGENCY DIRECTOR CHECKLIST Na-M M=ýWMM=='M 771 COMMENTS:
Emergency Director Daterrime
__ 1. IF turnover of ED responsibilities is required, THEN the on-coming ED should perform a thorough review of the sequence of events and conditions before turnover. They should use PINGP 571 as a guide.
-2.1. Announce to TSC that the off-going shift has been relieved.
__ 2.2. Notify the EM that the off-going TSC shift has been relieved.
-2.3. Notify the OSC Coordinator that the off-going TSC shift has been relieved.
Relief Emergency Director Date/Time
.t,RM - T Page 51 of 74
PRAIRIE ISLAND NUCLEAR GENERATING PLANT EMERGENCY PLAN IMPLEMENTING PROCEDURES S~NUMBER:
RESPONSIBILITIES DURING AN ALERT, F3-4 F3 SITES =REV: AREA, OR GENERAL EMERGENCY REV: 2 281 Figure 3 PINGP 1095 EXAMPLE ONLY PINGP 1095, Rev. 2 Page I of I USE CURRENT REVISION Document Type: 7.36T Retention: Lifetime OSC OPERATIONS ADVISOR CHECKLIST The OSC Operations Advisor position should be staffed by an available Operations person from the Operations Support Pool.
INOTE: Directien and control of outplant duty operators remains I with the Control Room.
INITIAL NOE The Emergency Work Status Board is used for tracking NOTE. emergency work repair groups and operational work activities. See the posted job aid for guidance.
- 1. Designate an individual (preferably from the Operations Support Pool) to update the OSC Emergency Work Status Board. (PINGP 1108)
- 2. Provide assistance to the OSC Coordinator in an advisory capacity.
- 3. Assist the OSC Coordinator, as appropriate, on the operational aspects of requested emergency work.
- 4. Assist, as necessary, the facilitation of directed emergency work requests.
- 5. If available, consider sending an Operations Support Pool individual to the EOF to assist the EOF Technical Support Supervisor in an advisory capacity.
- 6. Direct Operations Business Support Specialist (if available) to bring list of additional operators to the OSC Coordinator, to forward to Control Room.
OSC Ops Advisor:
Date:
JATEMPLATEN1095 OSC Operations Advisor Checklist.dot Page 52 of 74
PRAIRIE ISLAND NUCLEAR GENERATING PLANT EMERGENCY PLAN IMPLEMENTING PROCEDURES F3F3 S RESPONSIBILITIES DURING AN ALERT, SITE AREA, OR GENERAL EMERGENCY NUMBER:
F3-4 REV: 28 CURRPEN REVSIO Figure 4 PINGP 572 USEXAMPLE ONLY PINGP 572, Rev. 32 Page 1 of 6 RADIOLOGICAL EMERGENCY COORDINATOR CHECKLIST Retention: Life of Plant Document Type: 7.36K I_ 1. IF this4 isiHE an NUE asinaohrpro ocmlt 24 N/NHat et nta alBc
-_ 1. IFthis is an NUE, THEN assign another person to complete PINGP 1246 (WI/MN Health Dept. Initial Call Back Info.) in preparation for WI/MN call back.
__ 2. Determine Plant Status and if releases are occurring.
-3. Determine Radiation Protection Support Staff for: (See front of EP Phone Directory for PI ERO Roster.)
"* REC Assistant:
"* MIDAS Operator inTSC (PINGP 1312):
"* Field Tm Communicator (PINGP 1156):
"* Rad Status Board Keeper__
"* Phone Communicator:
"* OSC Rad Prot Coord:
"* EOF Count Room RPS:
"* Assembly Pt. Coord (should prepare assembly pt.) (PINGP 911):
"* MIDAS Operator at EOF (PINGP 1312):
"* RPSS or RPSS Assistant at EOF (ifneeded):
-4. Direct REC Area Staff to sign in at TSC Personnel Status Board.
__ 5. Dispatch and coordinate offsite radiation survey teams, via Field Team Communicator in accordance with the following:
0 F3-15 Airborne Release - Team #1 Plume Search &Sampling
- Team #2 Plume Search &Sampling OR 0 F3-16 Liquid Release - Team #1 Lock and Dam #3 Sampling
- Team #2 Recycle and Discharge Canal Sampling
_6. Dispatch 4 NPSA from OSC to go to EOF (Field Team Drivers & Sample Couriers).
-7. IFthis is an off hour activation, THEN direct Field Team Communicator to question Field Team Personnel per Fitness for Duty Program (5AWl 3.15.1).
__ 8. IF plume is traveling towards Minnesota OQR Wisconsin bluffs, THEN direct Field Team Communicator to consider effects of plume diversion along valley and look for plume up and down the valley.
.9. Evaluate survey rad data and supervise development of plume map. REC should decide ifand when plume search and sampling in upwind sectors is necessary.
- 10. Direct MIDAS Operator to obtain meteorological data (F3-13.4 and F3-13.5) and radiation monitor readings.
- 11. Direct MIDAS Operator to obtain offsite dose projection calculations for actual monitored releases and potential releases (F3-13). (Follow-up Message, PINGP 582, should be sent to states approximately 30-minute intervals.)
- 12. Request alternate meteorological data and weather forecast information as necessary (F3-13.5 and F3-113.6).
J:\TEMPLATE\0572 Radiotogical Emergency Coordinator Checklist.dot Page 53 of 74
PRAIRIE ISLAND NUCLEAR GENERATING PLANT EMERGENCY PLAN IMPLEMENTING PROCEDURES S.... NUMBER:
RESPONSIBILITIES DURING AN ALERT, F3-4 F3 SITE AREA, OR GENERAL EMERGENCY 28SREV:
EXAMPLE ONLY Figure 4 PINGP 572 USE CURRENT REVISION PING P 572, Rev. 32 Page 2 of 6 RADIOLOGICAL EMERGENCY COORDINATOR CHECKLIST
-13. Ensure MN &WI state health departments are updated and the 'Emergency Notification Follow-up Messages" (PINGP 582) are facsimiled to the state EOCs at approximately 30-minute intervals.
__ 14. Instruct, dispatch and coordinate onsite radiation survey teams (inplant surveys, smears and air sampling)
(F3-14.1).
- 15. Ensure radiological monitoring is initiated for the TSC environment utilizing the CAM and R-68 and/or AM-2 (F3-6).
- 16. Direct OSC Rad Coordinator to monitor the habitability of the Auxiliary Building and report results to you.
__ 17. lF the general area rad levels in Auxiliary Building are > 100 mR/hr OR based on Rad Protection Group surveys, THEN recommend to the Emergency Director that the Aux Operators evacuate to the OSC.
__ 18. Considering met data, release data and surveys; determine potential dose rates or contamination for non-rad controlled areas of plant (e.g., NPD, Environ. Lab, Gate House, Badge Issue, etc).
- 19. Direct Assembly Point Coord which assembly point to use based on rad hazards and wind direction (F3-9).
-20. Consider need for any special plant wide announcements concerning "avoiding hazardous areas onsite". (TSC Coordinator ensures plant wide emergency announcements are conducted, as needed.)
-21. Consider recommending early release of nonessential personnel per F3-9, if time allows.
- 22. Establish and maintain communications with the RPSS.
- 23. Turnover the offsite survey responsibilities to the RPSS:
- 23.1. Inform RPSS when next "Emergency Notification Follow-up Messages" to State EOCs are due.
.23.2. Remind RPSS to notify Minnesota and Wisconsin State Health Departments when EOF takes over all offsite functions.
Determine RPS sampling prioritiesS24.per F3-23 (number in order) and specify protective clothing/respiratory protection requirements for sampling team.
- Primary Sample - Pressurized/Unpressurized
- Secondary Sample
- Shield Building Stack Sample (consider sample plateout, F3-20)
- Containment Air Sample
"* Liquid
"* Filtered Gas
"* Iodine/Particulate
"* Other
__ 25. IF radioactive release is in progress QR radioactive samples are being shipped to EOF, THEN Initiate PINGP 1405, Request for Emergency Exemption for Transporting of Radioactive Samples.
-26. Provide periodic updates to the Emergency Director & TSC Staff (See TSC Update Guide, PINGP 1395).
-27. WHEN the NRC requests the establishment of the HPN, THEN staff the HPN phone and call one of the listed HPN phone numbers.
- 28. Consider the need for outside contractors for additional services.
-29. IF this was an NUE classification and the plant is escalating to an Alert emergency classification, THEN ensure PINGP 577, "Emergency Notification Report Form", is completed and taken to ED for review and approval.
... 30. IF plant is escalating the emergency classification, THEN proceed to the SITE AREA or GENERAL emergency section of this checklist.
, A Page 54 of 74
PRAIRIE ISLAND NUCLEAR GENERATING PLANT EMERGENCY PLAN IMPLEMENTING PROCEDURES SSITE FRESPONSIBILITIES DURING AN ALERT, NUMBER:
F3-4 AREA, OR GENERAL EMERGENCYREV: 28 EXAMPLE ONLY Figure 4 PINGP 572 USE CURRENT REVISION PINGP 572, Rev. 32 Page 3 of 6 RADIOLOGICAL EMERGENCY COORDINATOR CHECKLIST
- 1. Use F3-8 and PINGP 585, PAR Checklist for formulating all Protective Action Recommendations.
-2. Ensure "Casino Precautionary Shutdown" is considered for Site Area Emergency declaration per PINGP 585.
(The Indian Community Stutter Siren will be activated at SAE level by the SEC to alert the Indian Community of the SAE, regardless of Casino shutdown decision.)
-3. Complete PINGP 577, "Emergency Notification Report Form", and take to ED for review and approval.
- 4. Direct setup of TSC Contamination Control, IF necessary.
__ 4.1. Establish contamination control for TSC by setting up a frisker station (frisker &step-off pad) at TSC entrance (F3-6) (Frisker &step-off pad near Ventilation Control Cabinet).
.. 4.2. Station a person (NPSA) at TSC entrance to maintain TSC dosimetry and monitor frisking using PINGP 652.
__ 4.3. Post 'Step Off Pad Notice" signs at the TSC outer entrance, East End Lower Level (stored inVentilation Control Cabinet).
- 4.4. Direct all TSC Personnel to go through OSC for all plant activities to ensure proper rad. control and monitoring.
-5. Consider the necessity of Secondary Access Control (F3-21).
__ 6. IF the general area rad levels in Auxiliary Building are > 100 mR/hr OR based on Rad Protection Group surveys, THEN recommend to the Emergency Director that the Aux Operators evacuate to the OSC.
.. 7. Considering met data, release data and surveys; determine potential dose rates or contamination for non-rad controlled areas of plant (e.g., NPD, Environ. Lab, Gate House, Badge Issue, etc).
- 8. Consider need for any special plant wide announcements concerning "avoiding hazardous areas onsite". (TSC Coordinator ensures plant wide emergency announcements are conducted, as needed.)
-9. Recommend overexposure limits and/or use of KI for emergency workers, field teams, work teams, to Emergency Director, as necessary (F3-12, F3-18 & F3-19).
__ 10. IF dose rate is detected in the TSC, THEN request TSC Staff to check their SRDs. (Protective actions for the TSC are found in F3-6.)
. 11. Provide periodic updates to the Emergency Director &TSC Staff (See TSC Update Guide, PINGP 1395).
- 12. Periodically update the Assembly Pt. Coordinator.
-13. Periodically update the RPSS.
- 14. WHEN the NRC requests the establishment of the HPN, THEN staff the HPN phone and call one of the listed HPN phone numbers.
Page 55 of 74
PRAIRIE ISLAND NUCLEAR GENERATING PLANT EMERGENCY PLAN IMPLEMENTING PROCEDURES F3B4 RESPONSIBILITIES DURING AN ALERT, F SITE AREA, OR GENERAL EMERGENCY REV: 28 EXAMPLE ONLY Figure 4 PINGP 572 USE CURRENT REVISION PINGP 572, Rev. 32 Page 4 of 6 RADIOLOGICAL EMERGENCY COORDINATOR CHECKLIST
- 15. Determine RPS sampling priorities per F3-23 (number in order) and specify protective clothing/respiratory protection requirements for sampling team.
_ Primary Sample - Pressurized/Unpressurized
- Secondary Sample
- Shield Building Stack Sample (consider sample plateout, F3-20)
- Containment Air Sample
"* Liquid
"* Filtered Gas
"* Iodine/Particulate
"* Other
__ 16. IF plant is escalating to a General Emergency, THEN proceed to the GENERAL emergency section of this checklist.
S17. IF plant is terminating the emergency, THEN proceed to the RECOVERY procedure, F3-30.
- '.'* *~ ~~Tr' 7*.----E * ... 3 . . _ -_-*, * *,.
- .. , . .'*+*
. %I
.r. -** .
- S...
-- :* -
- 7 '3 Page 56 of 74
PRAIRIE ISLAND NUCLEAR GENERATING PLANT EMERGENCY PLAN IMPLEMENTING PROCEDURES FRESPONSIBILITIES F DURING AN ALERT, NUMBER:
F3-4 SITE AREA, OR GENERAL EMERGENCY EXAMPLE ONLY Figure 4 PINGP 572 USE CURRENT REVISION PINGP 572, Rev. 32 Page 5 of 6 RADIOLOGICAL EMERGENCY COORDINATOR CHECKLIST
- 1. Use F3-8 and PINGP 585, PAR Checklist for formulating all Protective Action Recommendations.
-2. Complete PINGP 577, 'Emergency Notification Report Form", and take to ED for review and approval.
__ 3. IF NOT already done AND itis necessary, THEN direct setup of TSC Contamination Control.
__ 3.1. Establish contamination control for TSC by setting up a frisker station (frisker &step-off pad) at TSC entrance (F3-6) (Frisker &step-off pad near Ventilation Control Cabinet).
-3.2. Station a person (NPSA) at TSC entrance to maintain TSC dosimetry and monitor frisking using PINGP 652.
-3.3. Post 'Step Off Pad Notice" signs at the TSC outer entrance, East End Lower Level (stored in Ventilation Control Cabinet).
-3.4. Direct all TSC Personnel to go through OSC for all plant activities to ensure proper rad. control and monitoring.
-4. Consider the necessity of Secondary Access Control (F3-21).
__ 5. IF the general area rad levels in Auxiliary Building are > 100 mR/hr OR based on Rad Protection Group surveys, THEN recommend to the Emergency Director that the Aux Operators evacuate to the OSC.
.. 6. Considering met data, release data and surveys; determine potential dose rates or contamination for non-rad controlled areas of plant (e.g., NPD, Environ. Lab, Gate House, Badge Issue, etc).
__ 7. Consider need for any special plant wide announcements concerning "avoiding hazardous areas onsite'. (TSC Coordinator ensures plant wide emergency announcements are conducted, as needed.)
__8. Recommend overexposure limits and/or use of KI for emergency workers, field teams, work teams, to Emergency Director, as necessary (F3-12, F3-18 &F3-19).
_9. IF dose rate is detected inthe TSC, THEN request TSC Staff to check their SRDs. (Protective actions for the TSC are found inF3-6.)
-10. Provide periodic updates to the Emergency Director &TSC Staff (See TSC Update Guide, PINGP 1395).
- 11. Periodically update the Assembly Pt. Coordinator.
__ 12. Periodically update the RPSS.
-13. WHEN the NRC requests the establishment of the HPN, THEN staff the HPN phone and call one of the listed HPN phone numbers.
__14. Determine RPS sampling priorities per F3-23 (number inorder) and specify protective clothing/respiratory protection requirements for sampling team.
- Primary Sample - Pressurized/Unpressurized
- Secondary Sample
- Shield Building Stack Sample (consider sample plateout, F3-20)
- Containment Air Sample
"* Liquid
"* Filtered Gas
"* Iodine/Particulate
"* Other
__ 18. IF plant is terminating the emergency, THEN proceed to the RECOVERY procedure, F3-30.
Page 57 of 74
PRAIRIE ISLAND NUCLEAR GENERATING PLANT EMERGENCY PLAN IMPLEMENTING PROCEDURES PRAIRIE ISLAND NUCLEAR GENERATING PLANT EMERGENCY PLAN IMPLEMENTING PROCEDURES NUMBER:
F3 RESPONSIBILITIES DURING AN ALERT, SITE AREA, OR GENERAL EMERGENCY REV:
F3-4 28 EXAMPLE ONLY Figure 4 PINGP 572 USE CURRENT REVISION PINGP 572, Rev. 32 Page 6 of 6 RADIOLOGICAL EMERGENCY COORDINATOR CHECKLIST 7'r 7:41 ý7 7 LL7- 41"?
COMMENTS:
Radiological Emergency Coordinator DateiTlme
- 1. IF turnover of REC responsibilities is required, THEN the on-coming REC should perform a thorough review of the sequence of events and conditions before turnover. They should use PINGP 572 as a guide.
-2. WHEN the off-going shift has been relieved, THEN Inform the ED and update the TSC Personnel Status Board.
Relief Radiological Emergency Coordinator Date/Time 2: ..
Page 58 of 74
PRAIRIE ISLAND NUCLEAR GENERATING PLANT EMERGENCY PLAN IMPLEMENTING PROCEDURES NUMBER:
RESPONSIBILITIES DURING AN ALERT, F3-4 F3 SITE AREA, OR GENERAL EMERGENCY SjREV: F3:28 4
Figure 5 PINGP 573 EXAMPLE ONLY 573 Rev 48USE CURRENT REVISION PING PING P 573, Rev. 48 Page I of 5 TECHNICAL SUPPORT CENTER COORDINATOR CHECKLIST Retention: Life of Plant Document Type: 7.36G Time of Classification
__ 1. Direct all TSC staff to sign inat TSC Personnel Status Board.
__ 2. Assign a TSC Coord Assistant to assist with TSC Coordinator tasks (PINGP 573). (Use of an assistant allows you to be available to the ED while your assistant completes remote EOF Coordinator duties.)
__ 3. Designate the following positions and refer them to the referenced procedure: (See front of EP Phone Directory for PI ERO Roster.)
"* Assign a TSC Coord Assistant (PINGP 573):
"* Record Keeper (F3-24, ED Log & PINGP 1108):
"* Technical CR-TSC-EOF 3-Way Communicators:
"* Control Room (PING P 1338) (Use TSC engineer for initial setup, turn over to Licensed Ops Staff person, ifavailable.):
"* In Tech Support Center (PINGP 1338):
"* Event Status Board Recorder (F3-26.1):
"* Emerg Work Status Board Keeper (Posted Job Aid):
"* ERCS Operator (F3-26.1):
"* Event Evaluator (SM or ED) (F3-2):
"* ENS Communicator (Red Phone &PINGP 1296):
"* Plant Wide Announcements (PINGP 1106):
__ 4. AFTER all the green positions on the TSC Personnel Status Board are filled, THEN notify the ED that the TSC is operational.
- 5. Verity the OSC is operational by calling the OSC Coordinator (x4424) and report results to ED.
- 6. Direct TSC Coordinator Assistant to:
-6.1. Post magnetic chain *no entr" & "no exit" signs on upper TSC West door to minimize SEC disturbance from personnel entering upper West door. (Signs are stored inTSC Supply Cabinet.)
-6.2. Verity TSC clocks are synchronized with ERCS clock within 1 minute.
.. 6.3. Procure copy of completed PINGP 577 from SEC for TSC ED to review upon arrival.
-6.4. Direct TSC Logistics person (Business Support, PINGP 1188) to ensure transport of the Electrical Transition Files from New Admin. Building to TSC lower level Tech. Resource Area.
-6.5. Shut TSC East Door by stairs, start TSC Clean Up System (F3-6) and post magnetic signs on East and West doors (signs stored in Ventilation Control Cabinets).
-6.6. Set up TSC Sound System (microphones and Job Aid are located on East End TSC near mic system amplifier).
-6.7. Verify plant PA speaker volume settings are appropriately turned up.
.. 6.8. Remind each work group to verify that their telephones are plugged in and the ringer is on.
-6.9. Verify the Operations Group Leader is using PINGP 1337, Operations Group Leader Checklist.
J:\TEMPLATEWs73, TR*.hnke.l A-r- P=nta*r"*l*trPh=**,
('ax ~
Page 59 of 74
PRAIRIE ISLAND NUCLEAR GENERATING PLANT EMERGENCY PLAN IMPLEMENTING PROCEDURES NUMBER:
RESPONSIBILITIES DURING AN ALERT, F3 JSITE AREA, OR GENERAL EMERGENCY REV:
F3-4 28 e.4 Figure 573,Fiue5PINGP 5 PINGP 573 USE CURRENT ONLY U EXAMPLE REVISION PINGP 573, Rev. 48 Page 2 of 5 TECHNICAL SUPPORT CENTER COORDINATOR CHECKLIST
-7. Direct the Recorder (Record Keeper) to:
__ 7.1. Assist the ED by answering the ED phone when they are not able to answer it.
__ 7.2. Keep log of work priorities from TSC Emergency Work Status Board, PINGP 1108.
__ 7.3. Keep the official ED Log.
-8. Use cell phone (located by TSC window) for backup communications if major loss of communications is experienced. Keys to lock box are located in locker.
- 9. Ensure SEC has completed required notifications (PINGP 577) and is periodically updated of plant status.
-10. WHEN the EOF has been activated and taken offsite communication responsibilities, that all future offsite emergency notifications will be done by EOF and report EOF ensure the SEC knows operational to ED.
- 11. Ensure the ERDS (NRC computer link) has been established within the 60-minute time requirement (TOC ERDS & press Fl) (F3-26.3).
- 12. Ensure appropriate NRC ENS Communications are conducted:
__ 12.1. Verify that Control Room (via the TSC Technical Communicator) has notified NRC of the event via the ENS phone (Per 10CFR50.72, NRC notifications are to be completed immediately after 15 minute state and local agency notifications and < 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />).
__ 12.2. Notify Control Room when TSC is ready to assume NRC ENS communications.
- 12.3. Use PINGP 666 for future ENS notifications and updates as a guide.
- 13. WHEN decision made to conduct an early release of plant personnel, THEN assist ED and Group Leaders in determining and releasing nonessential in accordance with F3-9.
-14. Remind ED to consider need for periodic TSC updates. (Use TSC Update Guide, PINGP 1395)
- 15. WHEN ED determines the time of the next Facility Update, THEN notify the OSC Coordinator (x4424) and EOF Coordinator (x4500) of the next Facility Update time.
- 16. Direct all reclassification announcements and periodic plant wide update announcements be made using PINGP 1106.
- 17. Direct all TSC Group Leaders to review their support staff needs and direct their extra personnel to position themselves in the TSC lower level or return to their normal work area until further notice.
- 18. Determine if additional TSC staffing (e.g., Engineering, REC area, command table, ERCS SEC support) is required and request SEC to make notifications if necessary. Consider operation, and FFD questioning if more staff are called out.
-19. Coordinate the establishment of the Employee Hot Line using PINGP 1106, if deemed necessary.
-20. WHEN ED directs establishment of long-term coverage, THEN coordinate the establishment of 24-hour coverage of all key emergency organization positions. Include FFD considerations related to fatigue, stress, illness, etc, using PINGP 1339.
- 21. Instruct plant operations and engineering to review the "Operator Workarounds" list and current "Status 80 Work Orders" to evaluate current plant configuration considerations as it relates to the event and mitigating actions.
-22. AFTER a LOCA AND within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> direct I&C calibrate the Containment H Monitoring 2 System. The manufacturer recommends calibration within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after a LOCA when containment pressure and temperature stabilize.
-23. IF plant is escalating the emergency classification, THEN proceed to the SITE AREA or GENERAL emergency section of this checklist.
Page 60 of 74
PRAIRIE ISLAND NUCLEAR GENERATING PLANT EMERGENCY PLAN IMPLEMENTING PROCEDURES F 3 RESPONSIBILITIES DURING AN ALERT, NUMBER:
F3-4 SITE AREA, OR GENERAL EMERGENCY SCURRENT REVISION Figure 5 PINGP 573 EXAMPLE ONLY 1USE CURRENT REVISION PINGP573,Rev.48 PINGP 573, Rev. 48 Page 3 of 5 TECHNICAL SUPPORT CENTER COORDINATOR CHECKLIST Time of Classification
- 1. Direct the following Plant Evacuation related activities:
__ 1.1. Confirm with Security Group Leader that TSC card readers are ready for reading ID cards.
__ 1.2. Direct all TSC personnel in lower and upper TSC insert their badges into the lower or upper (as appropriate) TSC card reader. This may be assigned to the Security Group Leader (or designee).
- 1.3. Post magnetic chain "no entry" & "no exit" signs on lower TSC West door. (Signs are stored in TSC Supply Cabinet.)
-1.4. Post "All Personnel Exiting the TSC..." sign on the I&C Cage (stored in Ventilation Control Cabinet).
-2. AFTER personnel accountability is complete, THEN ensure the Control Room, OSC, Assembly Point, and EOF are informed of the results:
- 2.1. Direct TSC Technical Communicator to inform Control Room of results.
__ 2.2. Notify OSC Coordinator (x4424) of results.
__ 2.3. Notify Assembly Pt. Coordinator (x 4292 N Warehouse, x5255 Rec. Warehouse) of results.
__ 3. IF EOF has NOT taken offsite communications, THEN ensure SEC has completed required notifications (PINGP 577).
_4. Procure copy of completed PINGP 577 from SEC for TSC ED to review.
__ 5. WHEN the EOF has been activated and taken offsite communication responsibilities, ensure the SEC knows that all future offsite emergency notifications will be done by EOF.
__ 6. IF radiological shine or release may affect TSC, THEN
_. 6.1. Issue pencil dosimetry (and TLDs if person does not have TLD) to all upper and lower TSC personnel per (F3-6 and PINGP 652).
-6.2. Ensure that issued dosimeters are less than 25% of full scale. (Re-Zero as necessary to approximately the first mark.)
-7. Assist non-plant representatives (e.g., NRC, Westinghouse, etc.) inTSC as necessary.
__ 7.1. May use PING P 1297 for initial NRC Team briefing.
__ 7.2. Utilize Logistics Coordinator assistance as necessary.
__ 8. AFTER a LOCA AND within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> direct I&C calibrate the Containment H2 Monitoring System. The manufacturer recommends calibration within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after a LOCA when containment pressure and temperature stabilize.
__ 9. IFplant is escalating to a General Emergency, THEN proceed to the GENERAL emergency section of this checklist.
__ 10. IFplant is terminating the emergency, THEN proceed to the RECOVERY procedure, F3-30.
""Mow" Page 61 of 74
PRAIRIE ISLAND NUCLEAR GENERATING PLANT EMERGENCY PLAN IMPLEMENTING PROCEDURES PRII SADNCEA EEAIGPAT MREC LNIPLMNIGPOEUE RESPONSIBILITIES DURING AN ALERT, F SREV:AREA, OR GENERAL EMERGENCY SITE F3-4 28 PINGP Figre5,INPR73eSEEX Pg 4 oCURRENT REVISION Page 4 of 5 TECHNICAL SUPPORT CENTER COORDINATOR CHECKLIST MPE4NL Time of Classification
- 1. J.Fnot already done, THEN direct the following Plant Evacuation related activities:
-1.1. Confirm with Security Group Leader that TSC card readers are ready for reading ID cards.
- 1.2. Direct all TSC personnel in lower and upper TSC insert their badges into appropriate) TSC card reader. This may be assigned to the Security the lower or upper (as Group Leader (or designee).
-1.3. Post magnetic chain "no entry" &"no exir signs on lower TSC West door.
(Signs are stored inTSC Supply Cabinet.)
__ 1.4. Post "All Personnel Exiting the TSC..." sign on the I&C Cage (stored in Ventilation Control Cabinet).
-2. AFTER personnel accountability is complete, THEN ensure the Control Room, OSC, Assembly Point, and EOF are informed of the results:
-2.1. Direct TSC Technical Communicator to inform Control Room of results.
- 2.2. Notify OSC Coordinator (x4424) of results.
- 2.3. Notify Assembly Pt. Coordinator (x 4292 N Warehouse, x5255 Rec.
Warehouse) of results.
-3. IFEOF has NOT taken offsite communications, THEN ensure SEC has completed required notifications (PINGP 577).
-4. Procure copy of completed PINGP 577 from SEC for TSC ED to review.
- 5, WHEN the EOF has been activated and taken offsite communication responsibilities, ensure the SEC knows that all future offsite emergency notifications will be done by EOF.
-6. IF NOT already done, THEN ensure TSC dosimetry is issued in TSC.
6.1. Issue pencil dosimetry (and TLDs if person does not have TLD) to all upper and lower TSC personnel per (F3-6 and PINGP 652).
- 6.2. Ensure that issued dosimeters are less than 25% of full scale. (Re-Zero as necessary to approximately the first mark.)
-7. Assist non-plant representatives (e.g., NRC, Westinghouse, etc.) inTSC as necessary.
__ 7.1. May use PING P 1297 for initial NRC Team briefing.
-7.2. Utilize Logistics Coordinator assistance as necessary.
.. 8. AFTER a LOCA AND within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> direct I&C calibrate the Containment manufacturer recommends calibration within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after a LOCA when H2 Monitoring System. The containment pressure and temperature stabilize.
_9. IF plant is terminating the emergency, THEN proceed to the RECOVERY procedure, F3-30.
C -i: A Page 62 of 74
PRAIRIE ISLAND NUCLEAR GENERATING PLANT EMERGENCY PLAN IMPLEMENTING PROCEDURES NUMBER:
RESPONSIBILITIES DURING AN ALERT, F3-4 F SITE AREA, OR GENERAL EMERGENCY V
Figure 5 PINGP 573 PINGP 573, Rev. 48 Page 5 of 5 ISEXAMPLE USE CURRENT REVISION ONLY TECHNICAL SUPPORT CENTER COORDINATOR CHECKLIST 1."Uif V -0.
COMMENTS:
TSC Coordinator Date/Time
__ 1. IF turnover of TSC Coordinator responsibilities is required, THEN the on-coming TSC Coordinator should perform a thorough review of the sequence of events and conditions before turnover. They should use PINGP 573 as a guide.
-2. WHEN the off-going shift has been relieved, THEN inform the ED and update the TSC Personnel Status Board.
Relief TSC Coordinator Date/Time TUNOERRE4FTrUFINWMER woo.F RELIEFJUNýW EE~AO~~ .I~~o Page 63 of 74
PRAIRIE ISLAND NUCLEAR GENERATING PLANT EMERGENCY PLAN IMPLEMENTING PROCEDURES RESPONSIBILITIES DURING AN ALERT, F3 SITE AREA, OR GENERAL EMERGENCY REV:
F3-4 28 Figure 6 PINGP 574 USE EAPEOL CURRENT REVISION PINGP 574, Rev. 28 Page 1 of 6 OPERATIONAL SUPPORT CENTER COORDINATOR CHECKLIST Document Type: Life of Plant Retention: 7.36J
__ 1. Assign an OSC Coordinator Assistant to perform several of the Activation Steps so you can keep the overall over sight of the OSC operation.
__ 2. Setup OSC sound system per job aid.
_3. Verify the green positions on the personnel board are staffed:
"* OSC RP COORDINATOR (PINGP 1245 checklist)
- RPS - RWp- Dosimetry
- RPS-Status-comm.
" ELECTRICAL SUPV/TECH
"* MECHANICAL SUPV.
"* I&C SUPV/TECH
- ...-'*-,, The Emergency Work Status Board is used for tracking emergency work repair groups and operational S . work activities. See the posted job aid for guidance.
__ 4. Designate an individual (preferably from the Operations Support Pool) to update the OSC Emergency Work Status Board. (PINGP 1108)
_5. As more personnel arrive, designate individuals to perform the following:
_5.1. Maintain the OSC Status Board.
- 5.2. Maintain the Emergency Center Narrative Log (PINGP 598) located in a 3-ring binder.
- 5.3. Control the use of equipment located in the emergency lockers.
- 5.4. Post the Enter and Exit signs on Control Room and OSC doors to control entrance and exit to OSC.
__ 5.5. Obtain portable radios from TSC for OSC use. Leave one radio in TSC for Field Team Communicator.
_6. Direct Operations Business Support Specialist (if available) to bring list of additional operators to the OSC Coordinator, to forward to Control Room.
__ 7. Direct the OSC Supervisors to verify telephones are plugged in, the ringer is on, and they have communication capability with their counterparts in TSC and or Control Room.
JT, ,AnATE\05VIONperational Support Ct Coordinator:rre-ki .0 .. .... *Ae... . ,W J:\TEMPLATE\0574 Operational Support Center Coordinator Checklist~dot Page 64 of 74
PRAIRIE ISLAND NUCLEAR GENERATING PLANT EMERGENCY PLAN IMPLEMENTING PROCEDURES NUMBER:
RESPONSIBILITIES DURING AN ALERT, F3-4 SITE AREA, OR GENERAL EMERGENCYREV:
28 CURETREVISION Figure 6 PINGP 574 EXAMPLE ONLY PINGP574,Rev.28 1USE CURRENT REVISION PING P 574, Rev. 28 Page 2 of 6 OPERATIONAL SUPPORT CENTER COORDINATOR CHECKLIST WINlowA., QJC zý",ý,tMVWIWTI
... 8. AFTER Step 1-3 of this checklist have been completed, THEN update OSC personnel of plant status and declare OSC operational.
- 8.1. Before you start your announcement:
"* Verify the OSC wall clock is correct with ERCS clock with in one minute.
"* Put on OSC Coordinator hat. - Test the microphone. - Stand. e Direct OSC Bell to be rung.
- tN1* - The following is an example of the expectations announcement.
8.2. Announcement
"=Attentionto all OSC workers."
8.3. Wait for all to be quiet and IF necessary, THEN repeat announcement.
"I am and I am the OSC Coordinator.
(Given Name)
The OSC is Operational at this time.
The main function of the OSC is to control the flow of people into and out of the plant. It is critical that your name appear on the Available OSC Personnel Board and that you check in and out with the Work Status Board Keeper.
I want all to follow my expectations so we can have clear, concise integrated communications. My expectations are:
"* All communications should follow the plant's communications standard using 3-way communication and the phonetic alphabet.
"* All personnel SHALL stop your conversations and listen to plant page announcements and facility updates.
", All personnel who participate in updates should follow the Facility Update Guide and keep your report as brief as possible.
- Communicate only that information that all should hear.
- Do not repeat information from previous updates unless the issue has not been addressed or pertinent update information is available.
- Let us keep facility updates to less than 8 minutes.
"* All OSC workers synchronize your watches with the OSC wall clock now.
All personnel are to check in with Radiation Protection and Work Status Board before leaving the OSC.
The present situation is:
Does anyone have anything to add?
Our Current Priorities are:
All personnel who are NOT assigned an OSC task please move to the OPS Lounge on the other side of the Control Room. I will notify you of the next update.
This is the end of the OSC update"
~~GIAIN 4ACTtVVNW "AIH. L~jV VATI -t:-p-WiV M. T70111 If Page 65 of 74
PRAIRIE ISLAND NUCLEAR GENERATING PLANT EMERGENCY PLAN IMPLEMENTING PROCEDURES
~NUMBER: Si[
RESPONSIBILITIES DURING AN ALERT, F3-4 FSITE S : :REV: AREA, OR GENERAL EMERGENCY 28
- PINGP 574, Rev. 28 PINGR 54,ev.28I Figure 6 PINGP 574 UEEXAMPLE ONLY USE CURRENT REVISION I Page 3 of 6 OPERATIONAL SUPPORT CENTER COORDINATOR CHECKLIST
_ 9. Notify TSC (ED 4332 or TSC Coordinator 4461) that the OSC is operational
-10. Notify Control Room (4250 or 4251) that the OSC is operational.
1 1) The REC SHALL be responsible for over all control and direction of the RPSs located in the OSO.
- 2) The Control Room SHALL be responsible for direction of the Operators located in the OSC
__ 11. Remind the Control Room that the Operators need to Check-in with Radiation Protection for current radiological information each time before going out.
__ 12. Call upon the OSC Operations Advisor for assistance. The Operations Advisor should use PINGP 1095, Ops Advisor Checklist.
- 13. Direct Maintenance Supervisors to have sufficient staffing of the respective trades and to consider FFD questioning ifmore staff are called out.
__ 14. IF ED orders an Early Release, THEN direct supervisors to determine and report to their TSC counterpart the names of personnel that can be released from the site.
__ 15. IF intake of smoke OR airborne contamination from turbine deck is probable, THEN direct the shutdown of the OSC ventilation makeup and exhaust fans (controls operated on west wall).
__ 16. Disallow use of food in OSC until directed otherwise by Radiation Protection Group.
__ 17. WHEN requested by TSC to establish 24-hour coverage, THEN complete PINGP 1376, OSC 24 Hour Coverage Worksheet and Include FFD considerations related to fatigue, stress, illness, etc.
- 18. IF plant is terminating the emergency, THEN follow the direction of the ED on deactivation of the OSC.
- 19. IF plant is escalating the emergency classification, THEN proceed to the SITE AREA or GENERAL Emergency section of this checklist.
P;4ATIIVATION I AT 1O ATATS ;
Page 66 of 74
PRAIRIE ISLAND NUCLEAR GENERATING PLANT EMERGENCY PLAN IMPLEMENTING PROCEDURES NUMBER:
RESPONSIBILITIES DURING AN ALERT, F3-4 F3 SITE AREA, OR GENERAL EMERGENCY REV: 28 PINGP 574, Rev. 28 Page 4 of 6 Figure 6 PINGP 574 OPERATIONAL SUPPORT CENTER COORDINATOR CHECKLIST I USE EXAMPLECURRENT ONLY REVISION I
- 1. WHEN a plant evacuation occurs, THEN ensure out plant teams are contacted to verify their well-being and activities for accountability purposes.
Security personnel will bring the personnel accountability list to the OSC for your verification of names.
1.1. WHEN accountability list is received in the OSC, THEN ensure all out plant repair team members are accounted for. The Emergency Work Status Board may be used as a tool in this process.
- 1.2. IF directed by the Emergency Director to conduct a manual accountability, THEN use PINGP 601 to list personnel names in OSC, Ops Lounge, Control Room and out plant personnel
... 1.2.1. Direct list to be delivered to Security Manager or Designee in the TSC.
__ 2. IF intake of smoke OR airborne contamination from turbine deck is probable, THEN direct the shutdown of the OSC ventilation makeup and exhaust fans (controls operated on west wall).
-3. IF plant is terminating the emergency, THEN follow the direction of the ED on deactivation of the OSC.
__ 4. IF plant is escalating the emergency classification, THEN proceed to the GENERAL Emergency section of this checklist.
Page 67 of 74
PRAIRIE ISLAND NUCLEAR GENERATING PLANT EMERGENCY PLAN IMPLEMENTING PROCEDURES NUMBER:
RESPONSIBILITIES DURING AN ALERT, F SITE AREA, OR GENERAL EMERGENCY REV:
F3-4 28 Figure 6 PINGP 574 EXAMPLE ONLY USE CURRENT REVISION PINGP 574, Rev. 28 I
Page 5 of 6 OPERATIONAL SUPPORT CENTER COORDINATOR CHECKLIST
__ 1. WHEN a plant evacuation occurs, THEN ensure out plant teams are contacted to verify their well-being and activities for accountability purposes.
- '* nmsSecurity personnel will bring the personnel accountability list to the OSC for your verification of N#names.
S1.1. WHEN accountability list is received in the OSC, THEN ensure all out plant repair team members are accounted for. The Emergency Work Status Board may be used as a tool in this process.
-1.2. IF directed by the Emergency Director to conduct a manual accountability, THEN use PING P 601 to list personnel names in OSC, Ops Lounge, Control Room and out plant personnel
__ 1.2.1. Direct list to be delivered to Security Manager or Designee in the TSC.
__ 2. IF intake of smoke OR airborne contamination from turbine deck is probable, THEN direct the shutdown of the OSC ventilation makeup and exhaust fans (controls operated on west wall).
__ 3. IF plant is terminating the emergency, THEN follow the direction of the ED on deactivation of the OSC.
[I%5G7ENR .Ž GEIR A. , GN19E -
Page 68 of 74
PRAIRIE ISLAND NUCLEAR GENERATING PLANT EMERGENCY PLAN IMPLEMENTING PROCEDURES NUMBER:
RESPONSIBILITIES DURING AN ALERT, F3-4 SITE AREA, OR GENERAL EMERGENCY PINGP 574, Rev. 28 Page 6 of 6 Figure 6 PINGP 574 OPERATIONAL SUPPORT CENTER COORDINATOR CHECKLIST CURRENT REVISION I USE EXAMPLE ONLY I 0 .. , BN0 COMMENTS:
OSC Coordinator Date/Time
_ 1. IF turnover of OSC Coordinator responsibilities are required, THEN the oncoming OSC Coordinator should perform a thorough review of the sequence of events and conditions before turnover. They should use PINGP 574 as a guide.
-2. Check with the following on turnover status.
"* ELECTRICAL SUPERVISOR
"* MECHANICAL SUPERVISOR
"* I&C SUPERVISOR
"* Ops Advisor (if available)
__ 4. WHEN turnover completed, THEN inform the TSC (ED 4332 or TSC Coordinator 4461).
Relief OSC Coordinator Date/Time
,~RU ftJNQ I! ~ EURN -I -VW4ELURNMIER Page 69 of 74
PRAIRIE ISLAND NUCLEAR GENERATING PLANT EMERGENCY PLAN IMPLEMENTING PROCEDURES I I NUMBER:
F3 RESPONSIBILITIES DURING AN ALERT, SITE AREA, OR GENERAL EMERGENCY F3R4 28 REV:
V Figure 7 PINGP 911 [ EXAMPLE ONLY USE CURRENT REVISION PINGP 911, Rev. 6 I I
Page 1 of 2 Retention: Life of Plant Document Type: 7.36N Assembly Point Coordinator Checklist
- 1. Proceed to the designated assembly area and set up the area as directed in F3-9.
Ask for a RPS, NPSA, Labor or other person who has had experience using meters and frisking to assist in radiological monitoring and frisking.
- 2. Ensure radiological monitoring is established per F3-14.1. Notify REC (4350 or 4834) of any radiological problems.
Name
- 3. Establish communications with the REC (4350 or 4834). (Designate an Assembly Point Communicator if desired.)
Name
- 4. IF weather conditions that may be hazardous to health (i.e. extreme heat or cold) exist, THEN recommend to REC (4350 or 4834) to move Assembly Point to the EOF or release personnel.
- 5. Designate a responsible individual for the control of Emergency Locker Equipment.
Name
- 6. Assist the Emergency Director with the Personnel Accountability Check, as necessary, per F3-10.
- 7. Determine the need for personnel wholebody frisking. Refer to F3-19 for guidance.
Frisking Monitor 8.
IF personnel OR vehicle deconning is necessary, THEN call REC and request a RPS.
- 9. Coordinate activities of all personnel located at the Assembly Point.
- 10. Give periodic updates to Assembly Point personnel as directed by the REC.
J:\TEMPLATE\0911 Assembly Point Coordinator Checklist.dot Page 70 of 74
PRAIRIE ISLAND NUCLEAR GENERATING PLANT EMERGENCY PLAN IMPLEMENTING PROCEDURES NUMBER:
RESPONSIBILITIES DURING AN ALERT, F3-4 SITE AREA, OR GENERAL EMERGENCY Figure 7 PINGP 911 PINGP 911, Rev. 6 I EXAMPLE ONLY USE CURRENT REVISION I
I Page 2 of 2
- 11. Provide instructions to personnel when they are released from the Assembly Point for re-entry or transport offsite.
- 12. IF a large number of vehicles (>50) are being released from site, THEN inform the RPSS (ext. 4443) of this fact, so this information can be passed onto Goodhue County EOC.
- 13. IF contamination is highly likely AND personnel have been released, THEN monitor all vehicles departing from the site per F3-19.
- 14. Ensure all Emergency Locker Equipment returned to proper place when Assembly Point is no longer required per the REC.
Assembly Point Coordinator Date/Time Page 71 of 74
PRAIRIE ISLAND NUCLEAR GENERATING PLANT EMERGENCY PLAN IMPLEMENTING PROCEDURES RESPONSIBILITIES DURING AN ALERT, F3-4 F SREV: SITE AREA, OR GENERAL EMERGENCY R-2 28 Figure 8 PINGP 1245 EXAMPLE ONLY USE CURRENT REVISION PINGP 1245, Rev. 3 Page 1 of 1 Document Type: 7.36ZB Retention: Life of Plant OSC RADIATION PROTECTION COORDINATOR CHECKLIST (1) Notify OSC Coord when you have 4 RPS personnel present.
(2) Verify R-65 operational or setup an OSC Area Monitor.(F3-7, Section 7.2)
Periodically check readings.
(3) Startup OSC/Control Room CAM. (F3-7, Section 7.2) Periodically verify proper operation of CAM.
(4) Determine radiological aspects of plant emergency and update OSC Coordinator; as necessary. Communicate with REC in TSC frequently to assess radiological aspects of the plant emergency and necessary RP Group priorities.
(5) If radiological event:
- b. Evaluate eating and drinking in the OSC/Control Room and make recommendations to OSC Coord. (F3-14.1)
- c. Establish radiological monitoring for OSC/Control Room and keep OSC Coord informed of habitability.
(6) Designate RPS to:
- a. Determine work groups inside Radiological Controlled Area and post on Work Status Board.
NAME (7) Designate RPS to setup RWP/Dosimetry issue desk. (F3-7, Section 7.3)
NAME (8) Designate person to communicate with Onsite survey teams and update Rad Status Board.
NAME (9) Direct RP Group personnel activities, as requested by REC. Keep REC informed of Radiological Surveys completed and those in progress. (F3-14.1)
(10) When a Plant Evacuation occurs:
- a. Designate RPS personnel to assist with radiological monitoring of personnel at Assembly Point. (F3-9)
NAME NAME
- b. Contact Onsite survey teams and verify activities for Accountability purposes. Assist with Accountability as necessary. (F3-10)
(11) Participate in periodic OSC Emergency Center updates (use OSC Update Guide).
(Should include a reminder to have RP Group included in Pre-Job Briefs.)
OSC RP Coordinator Date/Time J:\TEMPLATE%1245 OSC Radiation Protection Group Leader Checklist.dot Page 72 of 74
PRAIRIE ISLAND NUCLEAR GENERATING PLANT EMERGENCY PLAN IMPLEMENTING PROCEDURES NUMBER:
RESPONSIBILITIES DURING AN ALERT, F3-4 F SITE AREA, OR GENERAL EMERGENCY Figure 9 PINGP 1237 EXAMPLE ONLY USE CURRENT REVISION PINGP 1237, Rev. 0 Page 1 of 1 Document Type: 7.36ZC Retention: Life of Plant SAM EVALUATION TEAM LEADER CHECKLIST T Inmost cases, the Severe Accident Management Evaluation Team Leader will be I staffed by the General Superintendent Operations.
(1) Designate the following positions:
- a. Diagnostics Flow Chart (DFC) Evaluator:
- b. Severe Challenge Status Tree (SCST) Evaluator:
- c. Other Evaluators, as necessary (2) Direct Evaluators to use ERCS, TSC engineering and Tech Communicator to continuously access, trend and update the DFC and SCST.
(3) If TSC was not functional prior to entering Severe Accident Management, inform the Control Room that the TSC is now functional and able to perform the functions of Severe Accident Management.
_ (4) Conduct a briefing with the Shift Manager to obtain detailed information on plant conditions and what plant activities have been accomplished and what activities are in progress.
(5) Announce to the TSC that the TSC has entered into Severe Accident Management activities.
(6) Formulate severe accident strategies using the Severe Accident Guidelines, Severe Challenge Guidelines and Severe Accident Computational Aids.
(7) Recommend for authorization the implementation of the Severe Accident Strategies to the SAM Group Decision Maker (Emergency Director).
SAM EVALUATION TEAM LEADER/GSPO DATE/TIME J:\TEMPLATE\1 237 Sam Evaluation Team Leader Checklist.dot Page 73 of 74
PRAIRIE ISLAND NUCLEAR GENERATING PLANT EMERGENCY PLAN IMPLEMENTING PROCEDURES S~ NUMBER:
RESPONSIBILITIES DURING AN ALERT, F3 . SITE AREA, OR GENERAL EMERGENCY REV:
F3-4 28 Figure 10 PINGP 1188 EXAMPLE ONLY PINGP 1188, Rev. 5 I USE CURRENT REVISION I
I Page 1 of 1 Document Type: 7.36ZA Retention: 2 Years TSC LOGISTICS CHECKLIST I
- 1. Report to the TSC Coordinator and assist as necessary.
- 2. Ensure a switchboard operator has staffed the TSC Switchboard in the Shift Emergency Communicator area.
- 3. Transport the Electrical Transition Files from the admin. building to the TSC Tech Resource Area (east end, 1 st floor). (Complete within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> of activation notification).
- 4. Assist in notifying "Early Release" of site employee groups as requested (F3-9).
- 5. Assist issuing TSC dosimetry to personnel in upper and lower TSC areas using PINGP 652 when requested. Ensure dosimeters <25% scale before
- issuing.
- 6. Provide administrative and document support services as necessary.
- 7. Arrange for food and beverage for the plant's emergency organization in Control Room, OSC and TSC.
a) DO NOT locate food in TSC engineering work area.
b) Lunches may be placed in I&C area if NO radioactive release is occurring or in First Floor TSC. Lunches should be distributed to TSC workers as directed by TSC Coordinator.
_ _ 8. Assist with the receipt or escorting of the NRC Site Response Team members when they arrive.
a) Check with Radiological Emergency Coord. concerning a hazardous free entrance route.
b) We expect: 3 or 4 NRC in TSC, 1 in OSC and 1 (the resident inspector) in Control Room.
1188 TSC Logistics Checklist.DOT Page 74 of 74
PRAIRIE ISLAND NUCLEAR GENERATING PLANT EMERGENCY PLAN IMPLEMENTING PROCEDURES NUMBER:
ACTIVATION AND OPERATION OF F3-6 F3 TECHNICAL SUPPORT CENTER REV: 16 REFERENCE USE
"* Proceduresegments may be performed from memory.
"* Use the procedure to verify segments are complete.
"* Mark off steps within segment before continuing.
"* Procedureshould be availableat the work location.
O.C. REVIEW DATE: OWNER: EFFECTIVE DATE 3 -/1- *M M. Werner 4' 15- 0, Page 1 of 16
PRAIRIE ISLAND NUCLEAR GENERATING PLANT EMERGENCY PLAN IMPLEMENTING PROCEDURES F3 ACTIVATION TECHNICAL AND OPERATION OF SUPPORT CENTER NUMBER:
RE:73-6 1.0 PURPOSE The purpose of this instruction is to describe the activation and monitoring requirements of the Technical Support Center.
2.0 APPLICABILITY This instruction SHALL apply to all Shift Managers, Emergency Directors and all members of the TSC technical staff.
3.0 PRECAUTIONS 3.1 All unnecessary personnel SHALL be evacuated from the Technical Support Center when the Technical Support Center has been activated.
3.2 Monitoring of the Technical Support Center for direct radiation and airborne radioactive materials (particulate and iodine) SHALL be performed to ensure the habitability of the Technical Support Center.
3.3 Protective actions for individuals located in the Technical Support Center SHALL be taken at the prescribed levels of direct radiation or airborne radioactivity.
4.0 RESPONSIBILITIES 4.1 The TSC Coordinator is responsible to implement the actions directed in this procedure as necessary.
4.2 The Radiological Emergency Coordinator (REC) is responsible to provide oversight of the radiation monitoring of personnel and the TSC as necessary.
5.0 DISCUSSION The first and second floor of the Old Administration Building Office Annex is designated as the onsite Technical Support Center (TSC). This area SHALL be used by plant management, technical and engineering groups, and NRC representatives as a center outside the main control room from which support for emergency operating conditions can be provided. The TSC SHALL be activated when an Alert, Site Area, or General Emergency is declared.
Page 2 of 16
PRAIRIE ISLAND NUCLEAR GENERATING PLANT EMERGENCY PLAN IMPLEMENTING PROCEDURES NUMBER:
ACTIVATION AND OPERATION OF F3-6 F3 TECHNICAL SREV: SUPPORT CENTER 16 6.0 PREREQUISITES An Alert, Site Area, or General Emergency has been declared.
7.0 PROCEDURE 7.1 Activation of TSC 7.1.1 Activation of the Technical Support Center SHALL occur whenever an Alert, Site Area or General Emergency is declared. Activation of the TSC may occur during normal work hours or during off normal work hours:
A. During normal work hours, the Technical Support Center SHALL be activated whenever an Alert, Site Area, or General Emergency is declared, as announced over the public address system. All members of the Operations Committee and other designated engineers and staff members SHALL report to the TSC.
B. Ifactivation of the Technical Support Center occurs during off normal hours, the Emergency Director SHALL designate the Shift Emergency Communicator (SEC) to contact all Emergency Organization personnel, in accordance with F3-5.
7.1.2 All nonessential personnel SHALL evacuate the TSC area when the TSC has been activated.
7.1.3 Additional personnel should be notified and requested to report to the Technical Support Center as deemed necessary.
Nonessential personnel from the OSC may position NOTE: themselves in the TSC lower level due to high radiation dose rates or airborne contamination in the OSC.
7.1.4 As the emergency proceeds from the initial phase, (the period immediately following the emergency initiation) into the recovery phase, all Protective Actions for radiological hazards in the Technical Support Center SHALL be consistent with the plant Radiation Protection Program.
7.1.5 The Technical Support Center SHALL remain activated until the emergency situation has been terminated or as otherwise directed by the Emergency Director.
Page 3 of 16
PRAIRIE ISLAND NUCLEAR GENERATING PLANT EMERGENCY PLAN IMPLEMENTING PROCEDURES PRIIISADNCEA EEAIGPAT MREC LNIPLMNIGPOEUE NUMBER:
ACTIVATION AND OPERATION OF F3 TECHNICAL SUPPORT CENTER R:
F3-6 7.1.6 The TSC Coordinator SHALL ensure proper activation and operation of the TSC by completing the duties listed on PINGP 573, TSC Coordinator Checklist.
7.2 Radiological Monitoring of TSC NOTE: If airborne radioactivity is present in the TSC, manual dampers can be manipulated to route 100% of the upper level OR lower level return air through the filter unit per 7.3.4 of this procedure.
7.2.1 Monitor radiation dose rates on the TSC Area Monitor, R-68.
7.2.2 If R-68 fails, or is not working, set up the AM-2 for monitoring:
A. Obtain the AM-2 from the TSC Locker.
B. Plug the AM-2 in.
C3. Verify the green power light is on.
D. Source check the AM-2 with the button source in the TSC Locker and verify an upscale reading of meter.
E. If the AM-2 fails (power loss, incorrect reading, etc.), contact the Radiation Protection Group for additional radiation monitoring.
7.2.3 Establish operation of the TSC CAM The CAM, located in the Turbine Bldg. near the west NOTE: entrance to the TSC, is in a hot standby condition with the electronics energized and the blower, chart, and filter paper off.
A. Turn the blower switch to the ON position (located next to the recorder) to start the blower, strip chart recorder, and the filter paper drive.
B. Adjust the blower flow rate to 3 SCFM using the toggle switch located on the right side of the CAM.
C. Verify the CAM is in operation (i.e., verify the blower, filter, strip charts are operating; meters are on scale, etc.).
Page 4 of 16
PRAIRIE ISLAND NUCLEAR GENERATING PLANT EMERGENCY PLAN IMPLEMENTING PROCEDURES NUMBER:
ACTIVATION AND OPERATION OF F3-6 TECHNICAL SUPPORT CENTER REV: 16 D. Ifthe CAM fails to operate properly, contact the Radiation Protection Group for additional sampling.
7.2.4 Routinely monitor R-68 and/or VAMP (if.set up) for direct radiation levels, and the CAM for airborne particulate and iodine activity.
7.2.5 Take the following Protective Actions based on readings from the R-68, AM-2 or CAM.
A. DIRECT RADIATION
- 1. at about 15mR/hr consider evacuating all nonessential personnel from the Tech Support Center
- 2. at 1 R/hr evacuation to the Control Room is recommended A 1 mrem/hr background will cause a CAM reading of about NOTE: 20 cpm.
B. CAM - Particulate
- 1. < 1 x 10p.tCi/cc no protective action necessary
- 2. > 1 x 10.9 but< 1 x 10-6 O Ci/cc consider use of respiratory protection and evacuation of all unnecessary personnel
- 3. > 1 x 10-6 tACi/cc evacuation to the Control Room is recommended Page 5 of 16
PRAIRIE ISLAND NUCLEAR GENERATING PLANT EMERGENCY PLAN IMPLEMENTING PROCEDURES NUMBER:
ACTIVATION AND OPERATION OF F3-6 F3 TECHNICAL SUPPORT CENTER F316 NOTE: A 1 mremnhr background will cause a CAM reading of about 100 cpm.
- 1. If CAM alarms for iodine (5 x 109 giCi/cc), establish program of regular portable air samples by the Radiation Protection Group.
- 2. If portable air sample results > 1 DAC, consider evacuation of unnecessary personnel and limit exposures to less than 40 DAC - hours/week if possible.
- 3. Ifportable air sample results > 10 DAC, consider evacuation to the Control Room.
The Radiological Emergency Coordinator (REC) should recommend the use of potassium iodide pills (thyroid NOTE: blocking agent) if the projected thyroid exposure approaches 25 REM CDE. See F3-18, Thyroid Iodine Blocking Agent (Potassium Iodide), for determining projected thyroid exposures.
Page 6 of 16
PRAIRIE ISLAND NUCLEAR GENERATING PLANT EMERGENCY PLAN IMPLEMENTING PROCEDURES NUMBER:
ACTIVATION AND OPERATION OF F3-6 F3 TECHNICAL SUPPORT CENTER REV: 16 7.3 TSC Ventilation System For information regarding TSC electrical supply, See Figure 2, TSCIWCC HVAC, Lighting and Receptacles Electrical Block Diagram.
Following a loss of offsite power the TSC ventilation should be repowered as follows:
BKR 422-60 TSC Clean up Unit Fan on MCC 2FA1 is repowered automatically via D4 BKR 422-62 TSC Lower Level HVAC Unit on MCC 2FA1 is repowered automatically via D4 NOTE: BKR 263-8 TSC Upper Level HVAC Unit on MCC 2F2 is manually repowered via D3 using procedure C20.16 AOP1 which is invoked in procedure 1 ES-0.1 or 2ES-0.1 to restore battery room cooling and TSC cooling following a loss of offsite power, OR via D4 using procedure C20.16 AOP2 which is invoked in procedure 1 ES-0.1 or 2ES-0.1 to restore battery room cooling and TSC cooling following a loss of offsite power.
The upper and lower TSC HVAC supply air ducts are connected with a crossover duct and manual dampers, NOTE: located on the roof of the TSC. By changing the normal position of the manual dampers, either HVAC unit can be used to supply either floor. See Appendix A for damper lineup using the crossover duct.
7.3.1 System Startup A. CLOSE or ensure the TSC doors are closed.
B. At the TSC Control Panel, turn ES-82601 from "NORMAL" to "EMERGENCY."
Page 7 of 16
PRAIRIE ISLAND NUCLEAR GENERATING PLANT EMERGENCY PLAN IMPLEMENTING PROCEDURES NUMBER:
ACTIVATION AND OPERATION OF F3-6 TECHNICAL SUPPORT CENTER C. Place the Upper and Lower Level HVAC thermostat "AUTO-ON" fan switches in the "ON" position. (Thermostats located on north wall of upper level and south wall of lower level.)
D. Check the following indications at the TSC Control Panel (refer to Figure 1, TSC Clean-up System):
- 1. Filter Unit Trouble Light (Red light should be OFF).
NOTE: TSC Return Fans should stop.
- 2. The Clean Up Unit Fan (253-60) should start (Red "ON" Light should be lit).
- 3. Lower Level Outside Air Isolation Damper (MD-34602) should close (Green "CLOSE" light should be lit).
- 4. Upper Level Outside Air Isolation Damper (MD-34603) should close (Green "CLOSE" light should be lit).
- 5. Outside Air Volume Control Damper (MD-34604) should modulate open. (This damper modulates and will normally show mid position-both lights may or may not be lit).
- 6. Clean Up Unit Volume Control Damper (MD-34605) should modulate open. (This damper modulates and will normally show mid position-both lights may or may not be lit).
- 7. In the "EMERGENCY" mode, the TSC should be at a positive pressure greater than 0.125 inches W.G. as indicated by TSC Positive Pressure Gauge 82608.
E. Place the magnetic signs "TSC Vent System in Operation" on both sides of the 3 TSC Boundary doors (two on the West end and one on the East end).
Page 8 of 16
PRAIRIE ISLAND NUCLEAR GENERATING PLANT EMERGENCY PLAN IMPLEMENTING PROCEDURES NUMBER:
ACTIVATION AND OPERATION OF F3-6 F3 TECHNICAL SUPPORT CENTER REV: 16 7.3.2 Trouble Alarm The filter unit red trouble light will also be lit for a few seconds right after startup until the cleanup fan comes up to speed.
A. The Filter Unit Red Trouble light will be lit if the following occurs:
- 1. TSC Cleanup Unit Fan Differential Pressure Lo (DPIS-82321).
- 2. Prefilter Differential Pressure Hi (DPIS-82331).
- 3. 1st HEPA Filter Differential Pressure Hi (DPIS-82332).
- 4. Charcoal Filter Differential Pressure Hi (DPIS-82333).
- 5. 2nd HEPA Filter Differential Pressure Hi (DPIS-82334).
B. If the above light should be lit, investigate the local dp filter indications. The filters may require changing.
7.3.3 System Shutdown A. At the TSC Control Panel, turn ES-82601 from "EMERGENCY" to "NORMAL."
B. CHECK the following indications at the TSC Control Panel (refer to Figure 1, TSC Clean-up System):
- 1. Filter Unit Trouble Light (Red light should be OFF).
NOTE: The TSC Return Fans should start. 1
- 2. The Clean Up Unit Fan (253-60) should stop (Green "OFF" light should be lit).
- 3. Lower Level Outside Air Isolation Damper (MD-34602) should open (Red "OPEN" light should be lit).
Page 9 of 16
PRAIRIE ISLAND NUCLEAR GENERATING PLANT EMERGENCY PLAN IMPLEMENTING PROCEDURES NUMBER:
ACTIVATION AND OPERATION OF F3-6 F3 TECHNICAL SUPPORT CENTER F6
- 4. Upper Level Outside Air Isolation Damper (MD-34603) should open (Red "OPEN" light should be lit).
- 5. Outside Air Volume Control Damper (MD-34604) should close (Green "CLOSE" light should be lit).
- 6. Clean up Unit Volume Control Damper (MD-34605) should close (Green "CLOSE" light should be lit).
C. Remove the magnetic signs "TSC Vent System in Operation" from the TSC doors and place inside the TSC Control Panel.
D. Place the upper and lower level HVAC thermostat "AUTO-ON" fan switches in the "AUTO" position. (Thermostats located on the north wall of the upper level and south of the lower level.)
7.3.4 TSC Airborne Radioactivity Fast Clean Up A. Start System per 7.3.1.
NOTE: Mark the normal position of each damper prior to changing position.
B. OPEN the Lower Level PAC Filter Return Manual Damper (Damper B) (Reference Figure 1).
C. CLOSE the Upper Level PAC Filter Return Manual Damper (Damper D).
D. CLOSE the Lower Level Return Manual Damper (Damper A).
E. Partially CLOSE MD 34604 by changing the set point on FC 82103 to 10.
F. Monitor airborne levels in the Lower Level TSC until they are acceptable.
G. OPEN the Upper Level PAC Filter Return Manual Damper (Damper D).
H. Return the Lower Level Return Manual Damper to the Normal Position (Damper A).
I. CLOSE the Lower Level PAC Filter Return Manual Damper (Damper B)
Page 10 of 16
PRAIRIE ISLAND NUCLEAR GENERATING PLANT EMERGENCY PLAN IMPLEMENTING PROCEDURES NUMBER:
ACTIVATION AND OPERATION OF F3-6 F3 SREV: TECHNICAL SUPPORT CENTER 16 J. CLOSE the Upper Level Return Manual Damper (Damper C).
K. Monitor Airborne Levels in the Upper Level TSC until they are acceptable.
L. Return the Upper Level Return Manual Damper (Damper C) to the Normal Position.
M. Return the Lower Level PAC Filter Return Manual Damper (Damper B) to the Normal position.
N. Return the Upper Level PAC Filter Return Manual Damper (Damper D) to the Normal position.
- 0. Return the setpoint on FC 82103 to 25.
7.4 Dosimetry Issue The purpose of issuing dosimetry to personnel in the TSC is NOTE: - to provide a record of exposures received while working in the TSC and exposures received while in transit to and from the emergency centers.
7.4.1 If the event is a radiological event or has potential to be a radiological event, issue personnel dosimetry to each individual in the TSC and log initial dosimeter readings on PINGP 652, Emergency Center Activation Exposure Records.
7.4.2 As more personnel report to the TSC throughout the radiological event, ensure they are issued personnel dosimetry.
7.4.3 If a Secondary Access Control Point is set up for plant access, PINGP 652 should be forwarded to the Secondary Access Control Point for dosimetry logging when personnel leave the site at the end of their shift.
7.4.4 All entrances to the Aux Bldg SHALL be made via the OSC where each individual is advised of current radiological conditions in the Aux Bldg and issued additional dosimetry, as appropriate.
7.4.5 In those cases when an individual is going to an area where there exists high potential for the presence of radioactive contamination or radiation, the individual should go to the OSC first to be advised of current conditions and issued additional dosimetry, as appropriate.
Page 11 of 16
PRAIRIE ISLAND NUCLEAR GENERATING PLANT EMERGENCY PLAN IMPLEMENTING PROCEDURES AIR LERG F3 ACTIVATION AND OPERATION OF TECHNICAL SUPPORT CENTER REV:
F3-6 16 7.5 Set up of TSC Frisking Station 7.5.1 Place portable frisker on shelf, located in hallway outside east TSC entrance door, and plug into AC outlet.
7.5.2 Place Step-Off-Pad on floor outside East TSC entrance door.
7.5.3 All personnel entering TSC should frisk hands and feet (at a minimum) and areas of concern for possible radiological contamination.
7.5.4 If no radiological contamination exists outside the normal radiological controlled areas, the TSC frisking station may be removed.
Page 12 of 16
PRAIRIE ISLAND NUCLEAR GENERATING PLANT EMERGENCY PLAN IMPLEMENTING PROCEDURES II FN NUMBER:
ACTIVATION AND OPERATION OF F3-6 F3 TECHNICAL SUPPORT CENTER REV: 16 Figure 1 Page 13 of 16
PRAIRIE ISLAND NUCLEAR GENERATING PLANT EMERGENCY PLAN IMPLEMENTING PROCEDURES NUMBER:
ACTIVATION AND OPERATION OF R:3-6 TECHNICAL SUPPORT CENTER Figure 2 0
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Page 14 of 16
PRAIRIE ISLAND NUCLEAR GENERATING PLANT EMERGENCY PLAN IMPLEMENTING PROCEDURES NUMBER:
ACTIVATION AND OPERATION OF F3-6 F TECHNICAL SREV: SUPPORT CENTER 16 Appendix A The upper and lower TSC HVAC supply air ducts are connected with a crossover duct and manual dampers, located on the roof of the TSC. The crossover damper is normally closed. By changing the normal position of the manual dampers, either HVAC unit can be used to supply either floor.
- 1. Manual damper lineup for normal operations:
TSC UPPER LVL SUPPLY UPSTRM DAMPER - OPEN TSC UPPER LVL SUPPLY DWNSTRM DAMPER - OPEN TSC X-OVER DAMPER - CLOSED TSC LOWER LVL SUPPLY UPSTRM DAMPER - OPEN TSC LOWER LVL SUPPLY DWNSTRM DAMPER - OPEN
- 2. Manual damper lineup for UPPER LVL HVAC Unit OUT OF SERVICE, LOWER LVL HVAC Unit supplying air to upper level:
TSC UPPER LVL SUPPLY UPSTRM DAMPER - CLOSED TSC UPPER LVL SUPPLY DWNSTRM DAMPER - OPEN TSC X-OVER DAMPER - OPEN TSC LOWER LVL SUPPLY UPSTRM DAMPER - OPEN TSC LOWER LVL SUPPLY DWNSTRM DAMPER - CLOSED to force all air to upper level as desired OR ADJUST to provide air to both levels.
Page 15 of 16
PRAIRIE ISLAND NUCLEAR GENERATING PLANT EMERGENCY PLAN IMPLEMENTING PROCEDURES NUMBER:
ACTIVATION AND OPERATION OF F3-6 TECHNICAL SUPPORT CENTER16 Appendix A The TSC CAM sample tube is located in the TSC LOWER LVL NOTE: HVAC return air duct. If the TSC LOWER LVL HVAC unit is out of service, use grab samples to monitor TSC air quality.
- 3. Manual damper lineup for LOWER LVL HVAC Unit OUT OF SERVICE, UPPER LVL HVAC Unit supplying air to lower level:
TSC UPPER LVL SUPPLY UPSTRM DAMPER - OPEN TSC UPPER LVL SUPPLY DWNSTRM DAMPER - CLOSED to force all air to lower level as desired OR ADJUST to provide air to both levels.
TSC X-OVER DAMPER - OPEN TSC LOWER LVL SUPPLY UPSTRM DAMPER - CLOSED TSC LOWER LVL SUPPLY DWNSTRM DAMPER - OPEN Page 16 of 16
PRAIRIE ISLAND NUCLEAR GENERATING PLANT PLANT SAFETY PROCEDURE NUMBER:
F3 CORE DAMAGE ASSESSMENT REV:
F3-17 9
REFERENCE USE
"* Proceduresegments may be performed from memory.
"* Use the procedure to verify segments are complete.
"* Mark off steps within segment before continuing.
"* Procedureshould be availableat the work location.
O.C. REVIEW DATE: OWNER: EFFECTIVE DATE 7 ýcM.
ýz-Z Werner q-I og9 Page 1 of 39
PRAIRIE ISLAND NUCLEAR GENERATING PLANT PLANT SAFETY PROCEDURE 1.0 PURPOSE The purpose of this procedure is to provide a means to best estimate the degree of reactor core damage from the measured fission product concentrations in water and gas samples taken for the primary system and containment under accident conditions.
2.0 APPLICABILITY This procedure SHALL apply to the Nuclear Engineering Staff.
3.0 PRECAUTIONS 3.1 The numbers obtained using this procedure are at best, estimates only.
3.2 When making core damage calculations as per this procedure, considerations should be given to other plant indicators, for example:
3.2.1 Incore Thermocouples.
3.2.2 Reactor Coolant Loop Radiation Monitors (R70/71).
3.2.3 Containment Radiation Monitors (R48/49).
3.2.4 Hydrogen Concentration in the Containment Atmosphere 3.3 Spiking may occur after a shutdown or significant power change, usually during the 2 to 6 hour6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> period following the power change. Iodine spiking is a characteristic of the condition where an increase in the normal primary coolant activity is noted, but no damage to the cladding has occurred.
4.0 RESPONSIBILITIES The Nuclear Engineering Group is responsible to estimate the degree of reactor core damage according to the guidance provided in this procedure.
Page 2 of 39
PRAIRIE ISLAND NUCLEAR GENERATING PLANT PLANT SAFETY PROCEDURE NUMBER:
F3 CORE DAMAGE ASSESSMENT REV:
F3-17_
9 5.0 DISCUSSION The approach utilized in this methodology of core damage assessment is measurement of fission product concentrations in the primary coolant system, and containment, when applicable, utilizing the post accident sampling system.
Certain nuclides have been selected to be associated with each particular core damage state, i.e., clad damage, fuel overheat and fuel melt. These nuclides reach equilibrium quickly within the fuel cycle. Once equilibrium condition are reached, a fixed inventory of the nuclides is assumed to exist within the fuel pellet. For these nuclides which reach equilibrium, their relative ratios within the fuel pellet can also be considered to be constant. During operation, certain volatile fission products collect in the gap. The relative ratios in the gap can also be considered to be constant, however, the distribution of the nuclides in the gap is not in the same proportion as the fuel pellet inventory since the migration of each nuclide into the gap is dependent on its particular diffusion rate. The relative ratios of the nuclides analyzed during an accident may be compared to the predicted relative ratios existing in the gap and fuel pellet to determine the source of the fission product release, i.e., gap release or fuel pellet.
Clad damage is characterized by the release of these fission products, i.e., isotopes of the noble gases, iodine, and cesium which have accumulated in the gap and during the operation of the plant. When the cladding ruptures, it is assumed that the fission product gap inventory of the damaged fuel rods is instantaneously released to the primary system. For this methodology it is assumed that the noble gases will escape through the break of the primary system boundary to the containment atmosphere and the iodines will stay in solution and travel with primary system water during the accident.
Fission product release associated with overtemperature fuel conditions arises initially from the portion of the noble gas, cesium and iodine inventories that was previously accumulated in grain boundaries. In addition, small amounts of the more refractory elements, barium-lanthanum, and strontium are also released.
Fuel pellet melting leads to rapid release of many noble gases, halides, and cesiums remaining in the fuel after overheat conditions. Significant release of the strontium, barium-lanthanum chemical groups is perhaps the most distinguishing feature of melt release conditions.
Auxiliary indicators such as core exit thermocouples, reactor vessel water level, reactor coolant loop radiation monitors, containment radiation monitors, and the containment hydrogen concentration are available for estimating core damage. These indications should confirm the core damage estimates which in turn are based on the radionuclide analysis.
Page 3 of 39
PRAIRIE ISLAND NUCLEAR GENERATING PLANT PLANT SAFETY PROCEDURE NUMBER:
F3 CORE SREV:
DAMAGE ASSESSMENT F3-17 9
6.0 PREREQUISITES An emergency of an Alert, Site Alert, or General Emergency has been declared.
7.0 PROCEDURE NOTE: The program B8ODAMASS may be used whenever core damage estimates are desired.
7.1 Request the Radiation Protection Group to obtain the applicable samples to enable an adequate assessment of core damage. See Table 1 for suggested sampling locations.
7.2 Obtain the following plant data at the approximate sample time:
7.2.1 Incore Thermocouple Map 7.2.2 Containment Pressure 7.2.3 Containment Temperature 7.2.4 Containment Hydrogen Concentration 7.2.5 Containment Radiation Level 7.2.6 Containment Sump Level 7.2.7 RVLIS Level 7.3 Perform B80DAMASS according to the instructions in SWI-NE-5 (23) to obtain core damage estimates. Continue with Step 7.15 of this procedure when the B80DAMASS run is complete.
Page 4 of 39
PRAIRIE ISLAND NUCLEAR GENERATING PLANT PLANT SAFETY PROCEDURE NUMBER:
F3 CORE DAMAGE ASSESSMENT REV:
F3-17 9
NOTE: Ifthe computer is not available, perform the following manual calculations to obtain core damage estimates.
7.4 Decay correct the specific activities determined by the sample analysis, back to the time of reactor shutdown, as follows:
The decay correction may have been accomplished by the NOTE: computer during the spectrum analysis. Therefore, this step may not need to be completed.
A Ao=e -X it Where:
A = measured specific activity, gCi/gm or gCi/cc Xi = decay constant of isotope i, sec1 t = time elapsed from reactor shutdown to time of sampling, sec.
A0 = decay corrected specific activity .ICi/gmor pCi/cc 7.5 If a parent-daughter relationship exists for a specific isotope, the following steps should be followed to calculate the fraction of the measured activity due to the decay of the daughter that was released and then to calculate the activity of the daughter released at shutdown.
7.5.1 Calculate the hypothetical daughter concentration (QB) at the time of the sample analysis assuming 100 percent release of the parent and daughter source inventory:
XB QOAl(e-X't _e-ABt)+ QOB e-XBt QB (t) = XBXK
- XAi Where:
Q0 Ai = 100% source inventory (Ci) of parent i, Table 2 or Table 4.
Q0 B 100% source inventory (Ci) of daughter, Table 2 or Table 4.
QB (t) = hypothetical daughter activity (Ci) at sample time.
K= if parent has 2 daughters, Ki is the branching factor, Table 3.
XA = decay constant of parent i, sec XB = daughter decay constant, sec-1 t = time period from shutdown to time sample, sec.
Page 5 of 39
PRAIRIE ISLAND NUCLEAR GENERATING PLANT PLANT SAFETY PROCEDURE NUMBER:
F3 CORE DAMAGE ASSESSMENT REV:
F3-17 9
7.5.2 Determine the contribution of only the decay of the initial inventory of the daughter to the hypothetical daughter activity at sample time:
QOB eBt Fr =
Q5 (t) 7.5.3 Calculate the amount of decay corrected sample specific activity associated with just the daughter that was released.
MOB = Fr X A0 Where: A0 = decay corrected specific activity (RCi/gm or l.Ci/cc) as determined by the analysis.
7.6 Determine the total volume or mass of the medium which was sampled.
7.6.1 Containment Volume:
V = containment free volume (cc's)
= 3.74 X 1010 cc's 7.6.2 Liquid Mass:
A. Liquid temperature < 200°F Mass (gms) 28.3 X 103 cc
= volume (ft3) X PSTP X Where: PSTP = water density at STP = 1.0 gm/cc B. Liquid temperature > 200OF 3 CC Mass (gms) = volume (ft)3ft XP P (2) X PSTP X 28.3 Xl1O X POSTPft ft3 Where: P (2) = water density ratio at medium PSTP temperature, from Figure 1 PSTP = water density at STP = 1.0 gm/cc Page 6 of 39
PRAIRIE ISLAND NUCLEAR GENERATING PLANT PLANT SAFETY PROCEDURE NUMBER:
F3 CORE DAMAGE ASSESSMENT REV:
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7.7 Determine the total activity of each isotope in each medium.
7.7.1 Containment Atmosphere:
Curie Total containment A0 (pCi/cc) X V (cc's) X 1 X 106pCi Activity (curies)
Where: A0 = Specific activity of containment atmosphere (pCi/cc), decay corrected to time of reactor shutdown and temperature/pressure corrected.
V = containment free volume (cc's)
- 3.74 X 1010 cc's 7.7.2 Liquid Sample:
x Curie lCuri Total Liquid = Liquid x A0 (pCi/cc) 1 X10lýtCi Activity (Curies) MASS (gms)
Where: A0 = Specific activity of liquid sample (pCi/gm), decay corrected to time of reactor shutdown.
7.8 The approximate total activity of each isotope in the liquid samples can now be calculated.
Total Water Activity = RCS Activity + Sump Activity + Activity Leaked to Secondary System.
7.9 Now the total activity of each isotope released at the time of the accident can be determined:
Total Activity = Total Water + Containment Released Activity Atmosphere Activity Page 7 of 39
PRAIRIE ISLAND NUCLEAR GENERATING PLANT PLANT SAFETY PROCEDURE 7.10 Utilizing the total activity of each isotope released, calculate the activity ratios of the released fission products.
Noble Gas Activity 7.10.1 Noble Gas Ratio Xe - 133 Activity Iodine Activity 7.10.2 Iodine Ratio =
1-131 Activity Steady state power conditions may be assumed where power does not vary by more than +/- 10% of rated power level from time averaged value.
I 7.11 Determine the power history prior to reactor shutdown.
7.12 Using the power history, determine a power correction factor for each isotope, in accordance with the following guidelines:
Steady state power condition is assumed where the power does not vary by more than +/- 10% of rated power level from time averaged value.
I 7.12.1 Steady State power prior to shutdown.
A. Half-life of nuclide < 1 day Average Power Level (Mwt) for Prior 4 Days Power Correction Factor =
Rated Power Level (Mwt)
B. Half-life of nuclide > 1 day Power Correction Factor = Average Power Level (Mwt) for Prior 30 Days Rated Power Level (Mwt)
C. Half-life of nuclide - 1 year Average Power Level (Mwt) for Prior 1 year Power Correction Factor =
Rated Power Level (Mwt)
Page 8 of 39
PRAIRIE ISLAND NUCLEAR GENERATING PLANT PLANT SAFETY PROCEDURE NUMBER:
F3 CORE DAMAGE ASSESSMENT REV:
F3-17 9
7.12.2 Transient power history in which the power has not remained constant prior to reactor shutdown.
For the majority of the selected nuclides, the 30-day power NOTE: history prior to shutdown is sufficient to calculate a power correction factor.
A. Power Correction Factor (
RP Pj= average power level (Mwt) during operating period ti RP = rated power level of the core (Mwt) t = operating period in days at power Pj where power does not vary more than +/-10 percent power of rated power level from time averaged value (P1).
X = decay constant of nuclide i in inverse days.
tj= time between end of period j and time of reactor shutdown in days.
B. For the few nuclides with half-lives around one year or longer, a power correction factor which ratios effective full power days to total calendar days of cycle operation is applied.
Power Correction Factor = Actual Operating EFPD of equilibrium cycle Total expected EFPD of equilibrium cycle operation Where: Equilibrium Cycle = three (3) cycles of core operation (approximately 1050 EFPD) 7.12.3 For Cs-1 34, Figure 2 is used to determine the power correction factor.
To use Figure 2, the average power during the entire operating period is required.
7.13 The total inventory of fission products available for release at reactor shutdown are calculated by applying the power correction factors to the equilibrium, end-of-life core inventories.
Equilibrium Inventory at Power Corrected Inventory = end - of - life (Ci) X Correction (Table 2) Factor Page 9 of 39
PRAIRIE ISLAND NUCLEAR GENERATING PLANT PLANT SAFETY PROCEDURE NUMBER:
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F3-17 9
7.14 Determine the percentage of inventory released, for each isotope.
Release Total Activity Released (Ci) X 100 Corrected Inventory (Ci)
Percentage (%)
7.15 The results of radionuclide analysis may now be used to determine an estimate of the extent of core damage.
7.15.1 From Figure 3 thru 15, estimate the extent of core damage by categorizing the percentage of clad damage, fuel over-temperature, and fuel melt.
7.15.2 Compare the calculated activity ratios with those listed in Table 5.
Measured relative ratios greater than the gap activity ratios listed in Table 5 are indicative of more severe failures, e.g., fuel overheat.
7.16 To verify the conclusion of the radionuclide analysis, other indicators should now be used to provide verification of the estimate of core damage.
7.16.1 Containment Hydrogen Concentration:
A. Obtain the containment hydrogen concentration (%).
Within the accuracy of this methodology, it isassumed that NOTE:- recombiners will have an insignificant effect on the hydrogen concentration when it is indicated that extensive zirconium-steam reaction could have occurred.
B. From Figure 16, determine the percentage (%) zirconium water reaction.
C. Table 6 can be used to validate the extent of core damage estimate.
Page 10 of 39
PRAIRIE ISLAND NUCLEAR GENERATING PLANT PLANT SAFETY PROCEDURE NUMBER:
F3 CORE DAMAGE ASSESSMENT REV:
F3-17 9
7.16.2 Core Exit thermocouple Readings:
A. Obtain as many core exit thermocouple readings as possible for evaluation of core temperature conditions.
If a thermocouple reads greater than 1650OF or is reading NOTE: considerably different than neighboring thermocouples, thermocouple failure should be considered.
B. Compare the thermocouple readings with those in Table 7 to confirm the core damage estimate.
Radiation Monitors in containment may experience errors NOTE: during first 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> after a DBA LOCA due to thermally induced errors. See Attachment 1 for more information.
7.16.3 Containment Radiation Monitor:
A. Obtain the containment dome monitor readings, R/Hr, from R-48 and/ or R-49.
B. From Figure 17, verify core damage estimate. The exposure rate in Figure 17 is based on the release of only noble gases to the containment. Halogens and other fission products were not considered to be significant contributors to the containment monitor reading.
7.16.4 Reactor Coolant Loop Radiation Monitor:
A. Obtain the reactor coolant loop radiation monitor readings, R/Hr, from R-70 and/or R-71.
B. From Figure 18, determine estimated core damage.
7.17 All indicators should confirm any core damage estimates. If radio-nuclide analysis and auxiliary indicators do not agree on core damage estimates, then recheck of indications may be performed, or certain indicators may be discounted, based on engineering judgment.
Page 11 of 39
PRAIRIE ISLAND NUCLEAR GENERATING PLANT PLANT SAFETY PROCEDURE NUMBER:
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Table 1 Suggested Sampling Locations Principal Other Scenari o Sampling Locations Sam ling Locations Small Break LOCA Reactor Power > 1*1,A RCS Hot Leg, Containment Atmosphere Reactor Power < l%* RCS H-ot Leg**
Large BreaR LOCA Reactor Power > I%* Containment Sump, Containment Atmosphere, RCS Hot Leg Reactor Power < I'k* Containment Sump, Containment Atmosphere Steam Line Break RCS Hot Leg, Containment Atmosphere Steam Generator Tube RCS Hot Leg, Secondary Rupture System indication of Signif- Containment Sump, Containment icant Containment Sump Atmosphere Inventory Containment Building Containment Atmosphere, Radiation Monitor Alarm Containment Sump Safety Inj:ection RCS Hot Leg Actuated Indication of High RCS Hot Leg Radiation Level in RCS Assume operating at that level for some appreciable time.
Page 12 of 39
PRAIRIE ISLAND NUCLEAR GENERATING PLANT PLANT SAFETY PROCEDURE NUMBER:
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Table 2 Fuel Pellet Inventory Fuel Pellet Inventory*
Nuclide Half Life Inventory Curies**
Kr 85m 4.4 1.0 x 107 Kr 87 76 m 1.85 x 107 Kr 88 2.8 h 2.69 x 107 Xe 131m 11.8 d 2.94 x 105 Xe 133 5.27 d 9.26 x 107 Xe 133m 2.26 d 1.35 x 107 Xe 135 9.14 h 1.77 x 107 1 131 8.05 d 4.54 x 107 1 132 2.26 h 6.65 x 107 1 133 20.3 h 9.26 x 107 1 135 6.68 h 8.33 x 107 Rb 88 17.8 m 2.69 x 107 Cs 134 2 yr 1.09 X 107 CS 137 30 yr 4.96 x 106 Te 129 68.7 m 1.51 X 107 107 Te 132 77.7 h 6.65 X Sr 89 52.7 d 3.70 x 107 90 28 yr 3.36 x 106 Sr Ba 140 12.8 d 7.91 x 107 La 140 40.22 h 8.33 x 107 92.5 m 7.07 x 107 La 142 17.27 m 5.81 x 107 Pr 144 Inventory based on ORIGEN run for equilibrium, end-of-life core.
Westinghouse, 2-Loop, 1650 Mwt Plant Page 13 of 39
PRAIRIE ISLAND NUCLEAR GENERATING PLANT PLANT SAFETY PROCEDURE I
NUMBER:
F3 CORE [)AMAGE ASSESSMENT REV:
F3-17 9
Table 3 Parent-Daughter Relationships Parent Daughter Parent Half-Life* Daughter K**
Half-Life*
Kr-88 2.8 h Rb-88 17.8 m 1.00 1-131 8.05 d Xe-131m 11.8 d .008 1-133 20.3 h Xe-133m 2.26 d .024 1-133 20.3 h Xe-133 5.27 d .976 Xe-133m 2.26 d Xe-133 5.27 d 1.00 1-135 6.68 h Xe-135 9.14 h .70 Xe-135m 15.6 in Xe-135 9.14 h 1.00 1-135 6.68 h Xe-135m 15.6 m .30 Te-132 77.7 h1 1-132 2.26 h 1.00 Sb-129 4.3 h Te-129 68.7 m .827 Te-129m 34.1 d Te-129 68.7 m .680 Sb-129 4.3 h Te-129m 34.1 d .173 Ba-140 12.8 d La-140 40.22 h 1.00 Ba-142 11 m La-142 92.5 m 1.00 Ce-144 284 d Pr-144 17.27 m 1.00
- Table of Isotopes, L.ederer, Hollander, and Perlman, Sixth Edition Branching decay factor Page 14 of 39
PRAIRIE ISLAND NUCLEAR GENERATING PLANT PLANT SAFETY PROCEDURE NUMBER:
F3 CORE DAMAGE ASSESSMENT REV:
Table 4 Source Inventory of Related Parent Nuclides Nuclide Half-Life Inventory, Curies xe-135m 15.6 m 1.97 x 107 Sb-129 4.3 h 1.49 x 107 Te-129m 34.1 d 3.74 x 106 Ba-142 1l m 7.65 x 107 Ce-144 284 d 4.83 x 107 Page 15 of 39
PRAIRIE ISLAND NUCLEAR GENERATING PLANT PLANT SAFETY PROCEDURE NUMBER:
F3 CORE DAMAGE ASSESSMENT REV:
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Table 5 Isotopic Activity Ratios of Fuel Pellet and Gap Isotopic Activity Ratios of Fuel Pellet and Gap*
Nuclide Fuel Pellet Activity Ratio Gap Activity Ratio Kr-85m 0.11 0.022 Kr-87 0.22 0.022 Kr-88 0.29 0.045 Xe-131m 0.004 0.004 Xe- 133 1.0 1.0 Xe-i33m 0.14 0.096 Xe-135 0.19 0.051 1-131 1.0 1.0 1-132 1.5 0.17 1-133 2.1 0.71 1-135 1.9 0.39 Noble Gas Ratio - Noble Gas Isotope Inventory Xe-133 Inventory Iodine Ratio = Iodine Isotope Inventory 1-131 Inventory
- The measured ratios of various nuclides found in reactor coolant during normal operation is a function of the amount of "tramp" uranium on fuel rod cladding, the number and size of "defects" (i.e., "pin holes"), and the location of the fuel rods containing the defects in the core. The ratios derived in this report are based or, calculated values of relative concentrations in the fuel or in the gap. The use of these present ratios for post accident damage assessment is restricted to an attempt to differentiate between fuel overtemperature conditions and fuel cladding failure conditions. Thus the ratios derived here are not related to iu-e!
defect levels incurred during normal operation.
Page 16 of 39
Percent and Type Containment Radiogas Monitor Core Exit Mn WA Core of fission Fission Hydrogen R/hr 10 hrs Thermocouples Core Damage Products Product after Readings Monitor Uncovery Category Released Rat io* shutdown.6 (Vol % H2)
(Deg F) Indication -I Wn rI d AV 4 8ge . *1*
-KR-a? , -* 3 x 10 Not Applicablc - 750 KegI igibl I.ounrcovery C" Xt-I 33 1-131 xl 1-133 < I x 3D 0-50% clad damage Kr-87 Xe-133 ;0 1-131 10" - 0 01 103 0.1 0.3 Kr-97 = 0.022 1-133 = 0.71 0 - 50 750 - 1300 Core uncovery 0 - 6 0 00 1-133 1O" O.
- r
-5 m
50-100% clad damage Kr-87 0.01 - 0.02 Kr-87 = 0.022 50 to 100 1300 - 1650 Core uncovery 6 - 13 Xe-133 0.1 - 0.2 0 1-131 0.3 - 0.5 1-133 = 0.71 1-133 0.1 - Q) 0.2 m
0-50. fuel pellet overtemperstore Xe-Kr, Cc, I I - 20 Xr-87 = 0,22 100 to 1.15E4 > 1650 Core uncovery 6 - 13 0 0
M Sr-Ba 0 - 0.1 1-133 = 2.1 50-100% fuel pellet 0-h Xe-Kr, Cs, I Xr-87 = 0.22 1.15E4 to 2.3E4 > 1650 Core uncovery 6 - 13 overtempersture 20 - 40 C)
Sr-Ba 0.1 - 0.2 1-133 = 2.1 p.
CD 0
cn 0-50% fuel melt 6 - 13 C,)
Xe, Kr, Cc, I 40-70 Sr-Bs 0.2 - 0.8 Kr-87 = 0.22 2.3E4 to 2.7E4 > 1650 Core uncovery (U) 0 Pr 0.1 - 0.8 1-133 = 2.1 CA Co 50-100% fuel melt Xe, Xr, Cs, I, Te Kr-87 = 0.22 > 2.7E4 > 1650 Core urcovecy 6 - 13 0
> 70 C
Sr, Is > 24 Pr 3 0.8 1-133 = 2.1 z
(D -i Characteristics of Cateeoriem of Fuel Dsazae'*
This table is intended to supplement the methodology outlined in this report and should not be used without referring to this report and without considerable engineering judgement.
These values are from Figure 17 and should be revised for times other than 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br />, z
C 9
Kr-87 1-133 ca Xe-133' r-131 m m
Co CD 0 "C,,
CD CA)
PRAIRIE ISLAND NUCLEAR GENERATING PLANT PLANT SAFETY PROCEDURE NUMBER:
F3 CORE DAMAGE ASSESSMENT REV:
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Table 7 Expected Fuel Damage Correlation With Fuel Rod Temperature Fuel. Damage Temperature *F*
No Damage
< 1300 Clad Damage 1300 - 2000 Balloon:ing of zircaloy cladding
> 1300 Burst of zircaloy cladding 1300 - 2000 Oxidation of cladding and hydrogen generation
>. 1600 Fuel Overtemperature 2000 3450 Fissioni product fuel lattice mobility 2000 2550 Grain boundary diffusion release of fission 2450 3450 produ Is Fuel Melt > 3450 Dissolution and liquefaction of U0 in 2 > 3450 the Z:ircaloy - ZrO2 eutectic Melting of remaining U02 5100
- These ten.eratures are material property characteristics and are non-spec-fic with respect to locations within the fuel and/or fuel cladding.
Page 18 of 39
PRAIRIE ISLAND NUCLEAR GENERATING PLANT PLANT SAFETY PROCEDURE NUMBER:
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Figure 1 Water Density Ratio (Temperature vs. STP) 800-Z L.
i.]
I-J L
oc) 0/0 STP Page 19 of 39
PRAIRIE ISLAND NUCLEAR GENERATING PLANT PLANT SAFETY PROCEDURE NUMBER:
F3 CORE DAMAGE ASSESSMENT REV:
F3-17 9
Figure 2 Power Correction Factor For CS-134 Based on Average Power During Operation 0
0 Li z
0 0
L.J
-J Li Li
- 0. C> C o CS Lj L-J*
ce Page 20 of 39
PRAIRIE ISLAND NUCLEAR GENERATING PLANT PLANT SAFETY PROCEDURE NUMBER:
F3 CORE DAMAGE ASSESSMENT RE:F3-17 REV: 9 Figure 3 Relationship of % Clad Damage With % Core Inventory Released of XE-133
]
L.........
...... ..... . . . : o....
S.....
.. ... .. . i. .... .. "... .. '.. . , S t ...... .
S..
[Legend ]
S LOW BURNUP HIGH BURNUP I
A.
0" S
Li
-J LJ .......... ...... i S Z
>I z
Id Li ...... ...
0 ou 0.01. I K-i ...-. . .. .
U.Vu1-t 0.1 10 100 CLAD DAMAGE (.)
Page 21 of 39
PRAIRIE ISLAND NUCLEAR GENERATING PLANT PLANT SAFETY PROCEDURE NUMBER:
F3 FREV:
CORE DAMAGE ASSESSMENT F3-17 9
Figure 4 Relationship of % Clad Damage With % Core Inventory Released of 1-131 1'
~ il S...........
S...........
. .. .3... :... .. ..
.. ....... ....¢ S....:...... .... .....
S.......... . ...... *.... . .
.. . .... ...: ...L . ... :.
.:*' .i - ..
.. ...... .i , " ! i i ! S Legend 0.1.
K;HIGH AVERAGE BRU I
/..
...... /
7: ...........
.... I. i LA.J
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LdJ S.
- ,1 0 001.
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... . - . i.-. -.-
S7 ? i~7
- 1, .
S..........*... .... !..... .. ... 4 K
00 001 - -' - - - -r -r-r'
.......... .. ..... .. . ; ; .: ,. S.. . . . . . . . . *... .. . ;.. . . . .. . .: .
0.000..........
O. ;
10 100 CLAD DAMAGE (%)
Page 22 of 39
PRAIRIE ISLAND NUCLEAR GENERATING PLANT PLANT SAFETY PROCEDURE NUMBER:
F3 CORE DAMAGE ASSESSMENT REV:
F3-17 9
Figure 5 Relationship of % Clad Damage With % Core Inventory Released of 1-131 W/Spiking
.... i-i . .i. .. .. .
i ... .. .. . .. " .- . .. . .. ."*. ; 2 *"-*[*;
. . . ".-. . .I
... . ... ._... +.. .- . ! . .. * .... .. .. ... ... ..... . . ' -.. - - '
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. . . ]
../....
V) 0
-cc ...X.
La
-j 0.01. J 0 ............... ..... ... ..................
. d . : . .
..... . .... .... .. ... . ... 7/.
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.. .......... 7. . .. .. . .. .. . . .. .. .
0 n nni =:, ,o......... . . . .*....,... -.. .2 1 .. ;
.. . .. . .. . .. .. . .. . *... . .. ..,.. , .... .. .. . .. . . .. ... ..... : " ".. ':.. : :". ,.......... :....... !...-..,ý.,..- . .. .,
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. . .. , . . I.;
ii 0 .1 ... ~I ,. . . . , - . . . . .
CLADAMAGE- (/-.)10 16o Page 23 of 39
PRAIRIE ISLAND NUCLEAR GENERATING PLANT PLANT SAFETY PROCEDURE NUMBER:
F3 CORE DAMAGE ASSESSMENT REV:
Figure 6 Relationship of % Clad Damage With % Core Inventory Released of KR-87 0,.1
.":.......... ...... .----.. . -. .F *........ ./.. * -.- :.... '* " S..........
............ ...... ..L .L.. I.-. ........... ;...... - - "..... . ..-... .
0.01- VEAG
.......... ...... !:ii!*
...; "' = :
0 ........
i")
LJ
.- J ..... ..... .!.......-.* .
1 u
.- 0.00 1 . ...... ; .
I. . . .. . .
I-. .C .... ......... . (.. .....
z L..J 0>
C.
J ..
S........... _.i......-. ".. -... . . .o. ..
o.0001, .K... ....
................ * ... f .., !
0.0ooooi.
100 Page 24 of 39
PRAIRIE ISLAND NUCLEAR GENERATING PLANT PLANT SAFETY PROCEDURE NUMBER:
F3 Fige7 RREV:
CORE DAMAGE ASSESSMENT F3-17 9
Figure 7 Relationship of % Clad Damage With % Core Inventory Released of XE-1 31 M I
........... ....... i-i--ii i...... ......
S S
Legend S......... ....... i ii i ........ .
/*
"AVERAGE LOW BURNUP HIGH BURNUP 0.1- , , , * #
-AJ .................. ..... ..... ......
LJI
.......... . ... 7
............... .. . ..... . 7 z
wL ........... ..........7 ..... .
z 7... /
wj f ,, *
- 0)
S...........
S i.....
- -! i iii .
. .....*........... I TT I ...........
S S........... ..... .L .... :...-.. . . . .
0.001-1
- 0. 1 10 100 CLAD DAMAGE (%)
Page 25 of 39
PRAIRIE ISLAND NUCLEAR GENERATING PLANT PLANT SAFETY PROCEDURE Figure 8 Relationship of % Clad Damage With % Core Inventory Released of 1-132
. . . **!i....
- 3!i i~: -:
S.
Legend 0.01- AVERAGE LOW BURNUP .. ......... ........ .... .........
t..............
HIGH BURNUP
.. ..... ... i Xg...........
0.000 .. .....
..... .... IL .. .. .. . .. . .. . . .. ..
K. 7...
S.... .. .. ..- . ... .. .; .. ,- ; : -,
0.000 .. ... ..
S........... i.......i -- -i --i i !..
S....... . ...... . -. - - -" - - -F .
- ...... S......... -...... !....-...?...:.
..... ......... ii~
, ... -- ;....-. - o ; -'
- rrnrr 100 CLAD DAMAGE(%
Page 26 of 39
PRAIRIE ISLAND NUCLEAR GENERATING PLANT PLANT SAFETY PROCEDURE NUMBER:
F3 CORE DAMAGE ASSESSMENT REV:
F3-17 9
Figure 9 Relationship of % Clad Damage With % Core Inventory Released of 1-133 j ..*,
4---
, .r!..*..... ".... . ...
. .-. ? ""* .........................
....... *::!-.....!-i I S..... .. .. . *;... . ....
.............. ..............! . S S
... Legen.d .
I Ii 0
AVERAGE iLOW BURNUP .......... ....
HIGH BLo RNUP
........ .. --! -!-?-. S 0n 0. I LUJ 'I. 01.
L.J S. . .... .. .. ! ...... : . . .,,... .e........ "....... -... *- - - . ! "*
LiJ
> 17'.2 ..j .-T.... i:...
a 6 o2 LJ .. ... .. .. .:........... !!
3 ............ ... I .. . .. . .: . .. .......
0 r9 S..........
S.......... -......:
0.0001 ..........
. ] .. . .* '.. ........... * . .. : .... :..
S........... ..... .'........ .*' .: .. *..-.
S.........
.. :;.. .: i :*!i w
0.000014 100 0.1 CLAD DAMAGE (%)
Page 27 of 39
PRAIRIE ISLAND NUCLEAR GENERATING PLANT PLANT SAFETY PROCEDURE F.3 NUMBER:
CORE DAMAGE ASSESSMENT F3-17 REV: 9 Figure 10 Relationship of % Clad Damage With % Core Inventory Released of 1-135
...................... 7777FT ...................
I ............
ý.. .......... ......
0.1 r7Legend ........... ..............
... AVERAGE ........... .................
.. LOW BURNUP ........... ....... ................
HIGH BURNUP x Lai 0.01- J, iri
-j
... :...7..
La
.7.......
LaJ
(-)
A AAL+/-
z * .:::: .. ...... :...:: ::::::: S.....
A -tI ..........
LJ
. * -. 7...q- . .. .
0 -I.... .... ..' .t'i i i'i .... .......... ..............
A A Z'~
o .....
o.o 0.1.- .
I. ..... ..........
0.1 m I0 (0) 100 CLAD DAMAGE Page 28 of 39
PRAIRIE ISLAND NUCLEAR GENERATING PLANT PLANT SAFETY PROCEDURE NUMBER:
F3 CORE DAMAGE ASSESSMENT SREV:
F3-17 9
Figure 11 Relationship of % Fuel Over Temperature With % Core Inventory Released of XE, KR, I, or CS PW*
. ......... S...............
...... . ................ L .;. . .. . . . . . .;...
...... ............ ....i--i -i-. -i S............ ..... !......... *...... ...... ! -. - - ---
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S................. ......... i...... i..... "... * " "#'.
.. -.... !....... i....i....
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Legend S.............
.. ... . 7.
NOMINAL ....
MINIMUM MAXIMUM Q 10-LU .... .. . ........... .... ....... ..
z........................7........... ..............
LI) 7.
-j Li S...........
/..........................
/ ............ 7-.........i...... ..........
0I-z LU z
LIJ 0
(_)
I S................. ......... ,...... ..... ... ' . . ... .
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Page 29 of 39
PRAIRIE ISLAND NUCLEAR GENERATING PLANT PLANT SAFETY PROCEDURE NUMBER:
F3 CORE DAMAGE ASSESSMENT REV:
F3-17 9
Figure 12 Relationship of % Fuel Over Temperature With % Core Inventory Released of BA or SR 1 S..........
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Page 30 of 39
PRAIRIE ISLAND NUCLEAR GENERATING PLANT PLANT SAFETY PROCEDURE NUMBER:
F3 CORE DAMAGE ASSESSMENT REV:
F3-17 9
Figure 13 Relationship of % Fuel Melt With % Core Inventory Released of XE, KR, I, CS or TE finn S................ .......... i........ : -*-.. -.. -- *! -.
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............... . ...., : ! S I t I I 1 g i 10 1O0 FUEL MELT (7.)
Page 31 of 39
PRAIRIE ISLAND NUCLEAR GENERATING PLANT PLANT SAFETY PROCEDURE Figure 14 Relationship of % Fuel Melt With % Core Inventory Released of BA or SR I1,1l lu 7 . ....... . ... .. .. . . . ..... .. . ." "" ' " - -: " I S......
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Page 32 of 39
PRAIRIE ISLAND NUCLEAR GENERATING PLANT PLANT SAFETY PROCEDURE NUMBER:
F3 CORE DAMAGE ASSESSMENT REV:
F3-17 9
Figure 15 Relationship of % Fuel Melt With % Core Inventory Released of PR 1*1"1, Inn_
4 S................. . . . . ... . . . . .° ,. . . .. . .
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Page 33 of 39
PRAIRIE ISLAND NUCLEAR GENERATING PLANT PLANT SAFETY PROCEDURE NUMBER:
F3 CORE DAMAGE ASSESSMENT REV:
F3-17 9
Figure 16 Containment Hydrogen Concentration Based on Zirconium Water Reaction 25 20 20 ......
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Page 34 of 39
66 0o 96 efBed (saIOH) IN30100V *31-1V 31f.i OOOL 001 Ol I I- 4 -LO'[O
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PRAIRIE ISLAND NUCLEAR GENERATING PLANT PLANT SAFETY PROCEDURE NUMBER:
F3 CORE DAMAGE ASSESSMENT REV:
F3-17 9
Figure 18 RICS Dose Rate vs. RCS Activity Concentrations 1 Hour After Shutdown 1.0E+05 II!t I I~~~ - 1 !
IH
............................................................ _1 -I IZ l.LLJJ DBA: WOG 100% NG 50% lodines I I 100% DBA RELEASE L1IttII21f1221ZH21IEI:fzWtttIl 1.0E+04 Gap: WOG 100% Radionuclides in Fuel Gap I
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F 2.R S o. .88c 1.0E+01 based on 0.025 R/hr 1.0E+00 1.OE+02 1.OE+03 1.OE+04 1.0E+05 1.0E+06 1.0E RCS CONCENTRATION (Total uCi/ml)
Page 36 of 39
PRAIRIE ISLAND NUCLEAR GENERATING PLANT PLANT SAFETY PROCEDURE NUMBER:
F3 CORE DAMAGE ASSESSMENT REV:
F3-17 9
Attachment 1 Thermally Induced Current Errors in Containment Radiation Monitors
- 1. R-48/R-49 & R-70/R-71 Thermally Induced Errors R-48/49 or R-70/71 signals may experience errors during the first 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> after a DBA LOCA. Industry testing of high range radiation monitor (HRRM) systems has revealed that signal errors or the loss of signal are the result of thermally induced current (TIC) and/or moisture intrusion into the coaxial connectors. Based on the EPRI Plant Support Engineering study, worst case estimated errors are summarized below:
Time After Postulated Estimated Errors in DBA Readings
- 1 minute > 3000 R/hr
- 2 minutes - 100 R/hr m inutes - 15 S8 R/hr
- 2 to 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> - - 9 R/hr
> 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> No Effect from TIC More background information concerning thermally induced current in high range radiation monitors is described in Section II1.
Please note that errors in the range of +/-10 R/hr one hour after a postulated DBA has minimal effect on our assessment of fission product release to containment when we are considering magnitudes of > 100 R/hr reading to be confirmation of fission producttreleased to containment.
- 2. Background on Thermally Induced Current (TIC) on High Radiation Monitors
Background
Excerpts from: PINGP Response to High Range Radiation Monitor Cable Study: Phase II, Report No. TR-1 12582 November 2000.
Transient signal errors have been observed in industry testing of the high range radiation monitor (HRRM) system. At PINGP, these are plant radiation monitors RE-48 and RE-49.
The investigation into this issue revealed that signal errors or the loss of signal are the result of thermally induced currents (TICs) and/or moisture intrusion into the coaxial connectors. Information Notices, IN 97-45 and IN 97-45 Supplement 1, were issued by the NRC to alert licensees to these potential issues.
Page 37 of 39
PRAIRIE ISLAND NUCLEAR GENERATING PLANT PLANT SAFETY PROCEDURE NUMBER:
F3 CORE DAMAGE ASSESSMENT REV:
F3-17 9
Attachment 1 Thermally Induced Currrent Errors in Containment Radiation Monitors EPRI Plant Support Engineering (PSE) was tasked to study the significance of this issue, which resulted in the issuance of TR-1 12582, "High Range Radiation Monitor Cable Study: Phase I1". This study was focused on the thermally induced current phenomena since moisture intrusion issues are well understood within the industry and have more generic applications. Phase I of the EPRI study confirmed that TIC existed and was significant under thermal transients. Phase II of the study identified the sources of the TIC and developed a mathematical model for cable responses to thermal transients.
Study Results and Analysis Using the developed profiles in the Phase II study, the actual amplitude, duration, and sign of HRRM signal errors to be expected could be determined. From this data, PINGP was able to ascertain the approximate expected signal error for the HRRMs during the postulated DBA. The expected radiation readings due to the TIC phenomena, based upon the worst case cable length, are as follows:
50 seconds 3872 R/hr 100 seconds 88 R/hr 500 seconds 13.2 R/hr 8000-15000 seconds -8.8 R/hr
>15000 seconds no effect from TIC From 8000 seconds until 15000 seconds, the HRRMs could provide a "fail" alarm, based on the required "keep alive" signal current of 1E-1 1 amps since the current may drop to 8.8E-1 1 amps. It should be noted that any significant radiation releases would drive the current back up and the HRRMs would function properly, except for the-8.8 R/hr error that may be present. After 15000 seconds (4.1 hours1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />), there would be no TIC effects on the HRRMs.
The installed HRRM cable at PINGP is the worst case tested cable, Rockbestos RSS-6-104, and is in greater lengths than were tested, 130 feet tested vs. 290 feet installed (worst case). Other variables that could significantly effect the TIC phenomena are, 1) the tested cable was not installed within conduit whereas the PINGP cable runs are installed entirely within conduit, 2) the temperature differential of the test samples, 100 degc, is greater than the temperature differential from the PINGP accident profile, 68 degc, 3) the EPRI mathematical model was developed based on hypothetical LOCA profiles, which are more severe than the PINGP LOCA profile, and 4) consideration regarding whether the test methodology of immersion of the test samples into a ice bath and then to a boiling water plunge is representative of what the cable would experience during an actual transient.
Page 38 of 39
PRAIRIE ISLAND NUCLEAR GENERATING PLANT PLANT SAFETY PROCEDURE NUMBER:
F _3 CORE DAMAGE ASSESSMENT REV:
F3-17 9
Attachment 1 Thermally Induced Currrent Errors in Containment Radiation Monitors PINGP Response to HRRM Signal Error During the initial phase of any postulated accident, it would not expected to see indication of actual fuel damage for the first 10-15 minutes. If indeed the alarms would come in for RE-48 and RE 49, Operations would be occupied with accident mitigation and monitoring tasks during this time period and this alarm, even though acknowledged, would be ignored during this period. Other parameters would be available for alarm validation, i.e., core exit temperatures, RVLIS, radiation monitors located in the Auxiliary Building, etc. Due to the nature of the TIC phenomenon, the radiation level readings, even if the alarms have come in, would be decreasing. Again, this is validation of an erroneous signal and not actual core damage.
For emergency plan response and possible SAMG considerations, the TIC phenomenon would no longer be affecting the radiation monitors and/or due to the earlier alarms and decreasing readings that were noted, it would be confirmed that no fuel damage had occurred and these were indeed erroneous readings. A general site emergency alarm would be activated at 1000 R/hr, but as cited previously, this is well after the expected error signal has been significantly reduced. Other variables would be available to verify possible fuel damage and any possible actions required within the emergency plan procedures would not occur until after the TIC phenomena has either passed or has been verified to be erroneous.
Page 39 of 39
PRAIRIE ISLAND NUCLEAR GENERATING PLANT EMERGENCY PLAN IMPLEMENTING PROCEDURES NUMBER:
F3 TRANSITION TO RECOVERY REV:
7F3-30 6
REFERENCE USE
"* Proceduresegments may be performed from memory.
"* Use the procedure to verify segments are complete.
"* Mark off steps within segment before continuing.
"* Procedureshould be availableat the work location.
O.C. REVIEW DATE: OWNER: EFFECTIVE DATE 0-'o o M. Werner /-7/- 0 Page 1 of 11
PRAIRIE ISLAND NUCLEAR GENERATING PLANT EMERGENCY PLAN IMPLEMENTING PROCEDURES NUMBER:
F3 TRANSITION TO RECOVERY SREV:
F3-30 6l 1.0 PURPOSE The purpose of this procedure is to provide instructions for transition to the Recovery phase. This procedure specifies the criteria for termination of the emergency phase and guidance for the Emergency Director and Technical Support Center Group Leaders conducting on-site recovery planning.
2.0 APPLICABILITY This Instruction SHALL apply to all Emergency Directors, TSC Group Leaders, and TSC Record Keepers.
3.0 PRECAUTIONS NONE 4.0 RESPONSIBILITIES 4.1 Emergency Director is responsible to implement this procedure and assist the Emergency Manager in the decision to terminate the event or enter the Recovery phase.
4.2 Emergency Manager is responsible for the decision to terminate the event or enter the Recovery phase.
4.3 Plant Manager is responsible to coordinate the on-site recovery activities and assist the Recovery Manager.
4.4 The Recovery Manager (named the Event Manager in 5AWI 1.13.0) is responsible for the overall management of the recovery activities necessary to return the plant to a normal operational or shutdown status.
4.5 TSC Group Leaders are responsible to assess plant conditions in the respective areas and identify short and long term recovery actions necessary to return the plant to a normal operational or shutdown status.
4.6 TSC Record Keeper is responsible to assist the Emergency Director in compiling all identified recovery actions.
Page 2 of 11
PRAIRIE ISLAND NUCLEAR GENERATING PLANT EMERGENCY PLAN IMPLEMENTING PROCEDURES NUMBER:
F3 TRANSITION TO RECOVERY REV:
F3-30 6
5.0 DISCUSSION 5.1 Recovery In general, an Unusual Event or Alert may be terminated without transition to Recovery phase while a Site Area Emergency or General Emergency will probably require a planned transition to a Recovery phase and the establishment of a Recovery Organization. Whether a Unusual Event/Alert is terminated or a Site Area Emergency/General Emergency transition to a Recovery phase the site procedure 5AWI 1.13.0, "Plant Event Investigation and Recovery", should be reviewed and a determination to implement this procedure completed.
Termination of an Alert classification includes the dismissal of the site Emergency Response Organization. Any necessary in-plant or on-site follow-up activities should be coordinated and managed by the normal plant site organization in accordance with 5AWI 1.13.0. Post-Alert conditions may require the establishment of a Recovery Organization. The Emergency Director and Emergency Manager should make this determination based on the extent of damage or other considerations and implementation of 5AWI 1.13.0 should be completed.
If a Site Area Emergency does not require significant repairs or analysis beyond the capabilities of the normal plant site organization, the event may be terminated without transition to a Recovery phase, however, implementation of 5AWI 1.13.0 should be completed.
5.2 NRC Post-Accident Assessment It is expected that the NRC will, as a minimum, send an Augmented Inspection Team (AIT) to the plant to perform a thorough investigation of the incident. As a result, the AIT will request the following:
5.2.1 Any failed equipment not necessary for safe shutdown or operation of the plant be quarantined. No work on failed equipment should be performed unless absolutely necessary for plant safety. The NRC will want to perform failure analysis on the failed equipment.
5.2.2 No written or electronic records be destroyed or erased.
5.2.3 Schedules be adjusted or additional resources be made available for NRC interviews with plant personnel involved with the incident. A court reporter may be present to transcribe the interviews.
5.2.4 Office space (or trailer space) be made available for the NRC AIT.
Page 3 of 11
PRAIRIE ISLAND NUCLEAR GENERATING PLANT EMERGENCY PLAN IMPLEMENTING PROCEDURES NUMBER:
F3 TRANSITION TO RECOVERY RE:F3-30 REV: 6 6.0 PREREQUISITES 6.1 An Unusual Event or Alert has been declared and conditions indicate that plans for event termination may begin, or 6.2 A Site Area Emergency or General Emergency has been declared and conditions indicate the immediate emergency phase is over and transition plans to the Recovery phase may begin.
7.0 PROCEDURE 7.1 Termination of an Unusual Event or Alert 7.1.1 As plant conditions stabilize during an Unusual Event or Alert, consider terminating the event classification.
7.1.2 Assess plant and environmental conditions. PINGP 1102 or PINGP 1103, Shutdown Safety Assessment, may be used as an assessment guide.
7.1.3 The emergency classification may be terminated once the following criteria are met:
A. The plant is in stable condition with at least one fission product barrier intact, and B. Radioactive gaseous and liquid effluent are being controlled within the following limits:
- 1. Gaseous effluent release rates (or resulting dose rates) are within plant limits as defined in Section 3.1 of H4, ODCM.
- 2. Liquid effluent release rates (or resulting concentrations) are within the plant limits as defined in Table II, Column 2 of H4, ODCM, Table 4.3 Old 10CFR20 Appendix B (April 1992).
and C. The potential for future degradation of plant conditions is small (PINGP 1102 or PINGP 1103, Shutdown Safety Assessment, may be used as an assessment guide).
Page 4 of 11
PRAIRIE ISLAND NUCLEAR GENERATING PLANT EMERGENCY PLAN IMPLEMENTING PROCEDURES NUMBER:
F3 TRANSITION TO RECOVERY REV:
F3-30 6
7.1.4 When the criteria for termination are met, terminate the Unusual Event or Alert.
A. Termination of an Unusual Event classification may be performed by the Shift Manager.
B. Termination of an Alert classification SHALL be performed by the Emergency Director if the EOF is not activated.
C. Once the EOF is activated and responsible for off-site communications, the Emergency Manager SHALL terminate the Alert classification.
7.1.5 Upon event termination, ensure a message is announced and repeated over the public address system advising personnel of the termination.
Additional comments may be added as necessary to update plant personnel.
"ATTENTION PLANT PERSONNEL. THE (Specify Emergency Class) IS TERMINATED."
For Alert termination add:
"SECURE THE EMERGENCY RESPONSE CENTERS AND RESUME NORMAL DUTIES."
7.1.6 Review and approve the Notification Report Form (PINGP 577) and designate the SEC to complete the notifications of state, local and site personnel in accordance with F3-5, "Emergency Notifications." If the EOF has the responsibility for off-site communication, the Emergency Manager will coordinate the event close-out and off-site notifications.
7.1.7 Initiate outage planning if equipment and plant systems are significantly damaged and extended plant shutdown is required in accordance with 5AWI 1.13.0.
7.1.8 Direct plant Document Control group to collect all emergency checklists, documentation and records generated during the event.
7.1.9 Provide the necessary resources to support the requirements of the NRC post-event investigative team.
7.1.10 Develop and submit NRC follow-up reports through normal administrative procedures.
Page 5 of 11
PRAIRIE ISLAND NUCLEAR GENERATING PLANT EMERGENCY PLAN IMPLEMENTING PROCEDURES NUMBER:
F3 TRANSITION TO RECOVERY I RE:F3-30 REV: 61I 7.2 Transition to the Recovery Phase 7.2.1 WHEN the event has stabilized and plant conditions warrant possible transition to the Recovery phase, THEN assess plant and environmental conditions. PINGP 1102 and PINGP 1103, Shutdown Safety Assessment, may be used with other assessment tools to assess the condition of the plant.
7.2.2 Direct the TSC Group Leaders to assess conditions in their respective areas and identify recovery actions necessary to return the plant to a normal operational or shutdown status. TSC Group Leaders should use the Recovery Action Item Form, PINGP 1017 to identify their action items.
7.2.3 Direct the TSC Record Keeper to compile all identified recovery actions on the Recovery Action Item Form, PINGP 1017. The list should identify short-term corrective actions (to be completed within hours or days) and long-term actions which may require an extended plant outage (i.e., weeks or months).
If the Site Area Emergency event does not require significant repairs or analysis beyond the capabilities of the normal NOTE: plant site organization and the conditions of 7.1.3.A, 7.1.3.B,
& 7.1.3.C are met, then the Site Area Emergency may be terminated without a transition to Recovery phase.
7.2.4 When the following plant and environmental criteria are met, consider transition to the Recovery phase.
A. The plant is in a stable condition with at least one fission product barrier intact, and B. Radioactive gaseous and liquid effluent are being controlled within the following limits:
- 1. Gaseous effluent release rates (or resulting dose rates) are within plant limits as defined in Section 3.1 of H4, ODCM.
- 2. Liquid effluent release rates (or resulting concentrations) are within the plant limits as defined in Table II, Column 2 of H4, ODCM, Table 4.3 Old 10CFR20 Appendix B (April 1992),
and Page 6 of 11
PRAIRIE ISLAND NUCLEAR GENERATING PLANT EMERGENCY PLAN IMPLEMENTING PROCEDURES SP MMER N
F3 TRANSITION TO RECOVERY R REV:
F3-30 6
C. If Severe Accident Management (SAM) was implemented, the SAM termination criteria per SAMG Diagnostic Flow Chart are met, and D. The potential for future degradation of plant conditions is small (PINGP 1102 or PINGP 1103, Shutdown Safety Assessment, may be used as an assessment guide),
and E. NRC Headquarters (or the Director of Site Operations of the on-site response team) concurs with the transition to Recovery.
7.2.5 When the criteria for transition to Recovery are met, contact the Emergency Manager and discuss the conditions.
7.2.6 If the Emergency Manager concurs that conditions for Recovery are met, request the Emergency Manager discuss the proposed transition with the N RC . ........
If the NRC Incident Response Team has not been mobilized or has not arrived, the proposal to enter the Recovery phase NOTE: should be discussed with NRC Headquarters. If the NRC Response Team is on-site, the Emergency Manager and Emergency Director should meet with the Director of Site Operations to discuss the transition to Recovery.
7.2.7 If the NRC concurs that transition to Recovery is appropriate, obtain the compiled list of short and long term corrective actions on the Recovery Action Item Form PINGP 1017. Review the list with the TSC Group Leaders to ensure all items have been addressed.
7.2.8 Upon review completion, inform the Emergency Manager. Arrange a meeting with Emergency Manager and Recovery Manager to discuss the Recovery actions and the transition to Recovery. Implementation of 5AWI 1.1.13 should be started at this time also.
7.2.9 When the transition to Recovery is made, ensure a message is announced and repeated over the public address system advising personnel of the transition to the Recovery Phase. Additional comments may be made as to the current emergency center staffing needs or closeout of selected emergency centers.
Page 7 of 11
PRAIRIE ISLAND NUCLEAR GENERATING PLANT EMERGENCY PLAN IMPLEMENTING PROCEDURES NUMBER:
F3 TRANSITION TO RECOVERY SREV: F3-30 6 F "ATTENTION ALL PLANT PERSONNEL. THE (Specify the Emergency Class) IS TERMINATED. THE PLANT IS GOING TO RECOVERY. A RECOVERY ORGANIZATION IS BEING ESTABLISHED."
7.2.10 Initiate outage planning if equipment and plant systems are significantly damaged and extended plant shutdown is required.
7.2.11 Direct Document Control to collect all emergency checklists, documentation and records generated during the event.
7.2.12 Provide the necessary resources to support the requirements of the NRC post-event investigative team.
7.2.13 Develop and submit NRC follow-up reports through normal administrative procedures.
Page 8 of 11
PRAIRIE ISLAND NUCLEAR GENERATING PLANT EMERGENCY PLAN IMPLEMENTING PROCEDURES NUMBER:
FF3 TRANSITION TO RECOVERY REV:
- 3-30 6
Figure 1 Example Of Recovery Action Item Form I EXAMPLE ONLY USE CURRENT REVISION]
PINGP 1017, Rev. 4 Page 1 of 3 Retention: Life RECOVERY ACTION ITEM FORM Short Term = Hours or Days Long Term = Weeks or Months PERFORM ASSESSMENTS IN THE FOLLOWING AREAS AND IDENTIFY POTENTIAL SHORT TERM AND LONG TERM RECOVERY ACTION ITEMS:
A. Current operational status of plant systems and equipment involved in the emergency. Include reviews of "Operator Workarounds" and current "in progress" Work Orders to evaluate current plant configuration.
B. Current Operational status of the unaffected unit and its effect on the affected unit.
C. Identification of all systems, components, or equipment damaged or made inoperable during the event.
D. Estimate of necessary repairs, parts and tools to restore all affected systems and equipment back to a fully operational state.
E. Identification of special tools or equipment that may be required during the restoration period.
F. Estimate of additional personnel resources that may be required during the restoration period.
G. Identification of applicable plant surveillance tests and procedures required for post maintenance testing.
H. Identification of applicable system operability tests and procedures to restore plant systems to normal operational or shutdown configuration.
I. Estimate of liquid and solid radioactive waste generated during the event and recommendations on management and disposal.
J. Identification of special radiological considerations for personnel entry into affected areas with elevated dose rates or contamination levels (i.e.; temporary shielding engineering evaluations, robotics, etc.)
K. Estimate of the decontamination and monitoring activities necessary to restore affected areas inside and outside the plant site to pre-accident levels.
L. Identification of continued offsite radiological sampling and potential assistance to state or local agencies in the area of sampling, monitoring, and decontamination.
M. Following a containment LOCA, sample and adjust (if necessary) the pH of the primary coolant/recirculation loop within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> after the accident. Acceptable range is between 7.0 and 10.5 to minimize the potential for stress corrosion cracking of the stainless steel piping and components.
1017 Recovery Action Item Form.OT Page 9 of 11
PRAIRIE ISLAND NUCLEAR GENERATING PLANT EMERGENCY PLAN IMPLEMENTING PROCEDURES NUMBER:
F3 TRANSITION TO RECOVERY REV:
F3-30 6
Figure 1 Example Of Recovery Action Item Form I USE EXAMPLE CURRENT ONLY REVISION PINGP 1017, Rev. 4 Page 2 of 3 RECOVERY ACTION ITEM FORM SHORT-TERM RECOVERY ACTIONS (Hours or Days)
PREPARED BY __ TSC ___ EOF (Check one)
ITEM # ACTION ITEM DESCRIPTION COMMENTS Reviewed by: Date:
Emergency Director/Emergency Manager Page 10 of 11
PRAIRIE ISLAND NUCLEAR GENERATING PLANT EMERGENCY PLAN IMPLEMENTING PROCEDURES NUMBER:
F3-30 F3 TRANSITION TO RECOVERY REV: 6 Figure 1 Example Of Recovery Action Item Form I USE EXAMPLE REVISION] I CURRENT ONLY PINGP 1017, Rev. 4 Page 3 of 3 RECOVERY ACTION ITEM FORM LONG-TERM RECOVERY ACTIONS (Weeks or Months)
PREPARED BY _ TSC _ EOF (Check one)
ITEM # ACTION ITEM DESCRIPTION COMMENTS Reviewed by: Date:
Emergency Director/Emergency Manager Page 11 of 11