L-PI-18-016, Enclosure 1, Form 1576 Emergency Action Level (EAL) Matrix, Revision 10 and Emergency Action Level Technical Bases, Revision 13

Enclosure 1, Form 1576 Emergency Action Level (EAL) Matrix, Revision 10 and Emergency Action Level Technical Bases, Revision 13

ML18113A054
Person / Time
Site:	Prairie Island
Issue date:	04/17/2018
From:	Northern States Power Co, Xcel Energy
To:	Office of Nuclear Reactor Regulation
Shared Package
ML18113A047	List: L-PI-18-016, Enclosure 1, Form 1576 Emergency Action Level (EAL) Matrix, Revision 10 and Emergency Action Level Technical Bases, Revision 13 ML18113A053
References
L-PI-18-016
Download: ML18113A054 (41)

Text

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. ENCLOSURE 1

• PRAIRIE ISLAND NUCLEAR GENERATING FORM (PINGP) 1576

. EMERGENCY ACTION LEVEL (EAL) MATRIX, REVISION 10 AND EMERGENCY ACTION LEVEL TECHNICAL BASES, REVISION 13

• 'PINGP 1576 Emergency Action Level (EAL) Matrix (Transmittal Group IDs 1020 for:

Document Control Desk (two sets of EAL pages)_

Nuclear Material Safety/Safeguards (one set of EAL pages); and .

Chief of Security and Preparedness Region Ill with one CD-ROM (two sets of EAL pages)

F3-2.1 Emergency Action Level Technical Bases (Transmittal Group ID 1018 for:

Document Control Desk (partial update) F3-2.1 (Page i thru iii) (6-F-1 thru .6-F-17) double sided Chief of Security and Prepar,edness Region Ill (partial update) F3-2.1

.(Page i thru iii) (6-F-1 thru 6-F-17) double sided and one CD-ROM 5 Sets of EAL Matrixes (8 pages in each set) and One CD-ROM

• 2 Set of Pages of F3-2-1 (12 pages double sided for each set) and One CD-ROM

CD-0676 Controlled Document Transmittal REV.2 Report Date: 12/18/2017 To US NRC C/0 PAM JOHNSON (P.I.) From C-DOC CNTRL-PI Facility Pl Address 1717 WAKONADE DR Address PAMELA JOHNSON WELCH, MN 55089 DOCUMENT CONTROL DESK PARTIAL UPDATE US NRC Vital NO Transmittal 12/18/2017 Transmittal Group ID 1018 Date Vital Ack Facility Doc Type Sub Type Document Number Status Revision Status Date Copy Holder . Media Copies Req

• Pl PRO EP F3-2.1 ISSUED 013 12/18/2017 515 HC 1 Marked (*) documents require your acknowledgement.

Acknolwedgement Date: _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ Signature: - - - - - - - - - - - - - - - -

If documents no longer required for this copyholder, complete QF2122 Request for Service, and submit to Document Control.

PRAIRIE ISLAND NUCLEAR GENERATING PLANT EMERGENCY PLAN IMPLEMENTING PROCEDURE NUMBER:

. EMERGENCY ACTION LEVEL F3-2.1

• TECHNICAL BASES REV: 13

~fffifl~t

*****,

• Procedure segments may be performed from memory.

Use. the procedure to verify segments are complete. .

Mark off steps within segment before continuing.

Procedure should be available at the work location.

PORC REVIEW DATE: APPROVAL:

PCR #: 602000001144 12/1/17 Page i

PRAIRIE ISLAND NUCLEAR GENERATING PLANT EMERGENCY PLAN IMPLEMENTING PROCEDURE NUMBER:

EMERGENCY ACTION LEVEL F3-2.1

• TECHNICAL BASES REV: 13 Record of Revision Date of Revision Revision Number Title Page 2017 13 Record of Revision 2017 13 Emergency Action Level Technical Bases Document (22 pages) 2015 11 Table R-0 Category R - Abnorma_l Rad Levels/Radiological 2014 10 Effluent (19 pages)

Table C-0 Category C - Cold Shutdown/Refueling System 2017 12 Malfunction (29 pages)

Table E-0 Category E - Independent Spent Fuel Storage 6 Installations (ISFSI) (4 pages)

Table F-0 Category F - Fission Product Barrier Degradation 2017 13 I_

(17 pages)

Table H-0 Category H - Hazards (28 pages) 2014 10 Table S-0 Category S - System Malfunction (30 pages)

• 2017 12 Page ii

PRAIRIE ISLAND NUCLEAR GENERATING PLANT EMERGENCY PLAN IMPLEMENTING PROCEDURE

• EMERGENCY ACTION LEVEL TECHNICAL BASES NUMBER:

REV:*

F3-2.1 13 Table 1 Significant Changes From the Previous Revision Section Change Table F-0 Incorporate changed values for subcooling and RVLIS for Table F-1 basis information for Fission Product Barriers .

• Page iii

• Table F-0 Recognition Category F F.ission Product Barrier Degradation INITIATING CONDITION MATRIX UE ALERT SITE AREA EMERGENCY GENERAL EMERGENCY FU1 ANY Loss or ANY Potential Loss FA1 ANY Loss or ANY Potential Loss FS1 Loss or Potential Lo~s of ANY FG1 Loss of ANY Two Barriers AND of Containment of EITHER Fuel Clad OR RCS Two Barriers Loss or Potential Loss of Third Barrier Op. Modes: Power Operation, Op. Modes: Power Operation, Op. Modes: Power Operation, Hot Standby, Startup, Hot Hot Standby, Startup, Hot Hot Standby, Startup, Hot Op. Modes: Power Operation, Shutdown Shutdown Shutdown Hot Standby, Startup, Hot Shutdown NOTES

1. The logic used for these initiating conditions reflects the following considerations:

• The Fuel Clad Barrier and the RCS Barrier are weighted more heavily than the Containment Barrier. UE ICs associated with RCS and Fuel Clad Barriers are addressed under Sy~tem Malfunction ICs.

• At the Site Area Emergency level, there must be some ability to dynamically assess how far present conditions are from the threshold for a General Emergency. For example, if Fuel Clad and RCS Barrier "Loss" EALs existed, that, in addition to offsite dose assessments, would require continual assessments of radioactive inventory and containment integrity. Alternatively, if both Fuel Clad and RCS Barrier "Potential Loss" EALs existed, the Emergency Director would have more assurance that there was no immediate need to escalate to a General Emergency.

• The ability to escalate to higher emergency classes as an event deteriorates must be maintained. For example, RCS leakage steadily increasing

• would represent an increasing risk to public health and safety.

2. Fission Product Barrier ICs must be capable of addressing event dynamics .. Thus, the EAL Reference Table F-1 states that imminent (i.e., within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />) Loss or Potential Loss should result in a classification as if the affected threshold(s) are already exceeded, particularly for the higher emergency classes.

• PINGP 6-F-1 F3-2.1, Rev. 13

This page intentionally blank.-

PINGP 6-F-2 F3-2.1, Rev. 13

TABLE F-1 PINGP Emergency Action Level Fission Product Barrier Reference Table Thresholds For LOSS or,POTENTIAL LOSS of Barriers*

• Determine which combination of the three barriers are lost or have a potential loss and use the following key.to classify the event. Also an event for multiple events could occur which result in the conclusion that exceeding the Loss or Potential Loss thresholds is imminent (i.e., within 1 to 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />). In this imminent loss situation use judgment and classify as if the thresholds are exceeded.

• UNUSUAL EVENT ALERT SITE AREA EMERGENCY GENERAL EMERGENCY FU1 ANY loss or ANY Potential Loss of FA1 ANY Loss or ANY Potential Loss of FS1 Loss or Potential Loss of ANY two FG1 Loss of ANY two Barriers AND Containment EITHER Fuel Clad OR RCS Barriers Loss or Potential Loss of Third Barrier Fuel Clad Barrier EALS RCS Barrier EALS Containment Barrier EALS LOSS POTENTIAL LOSS LOSS POTENTIAL LOSS LOSS POTENTIAL LOSS

1. Criticai Safety Function Status . 1. Critical Safety Function Status 1. Critical Safety Function Status Core-Cooling Red Core Cooling-Orange Not Applicable RCS Integrity-Red Not Applicable Containment-Red OR OR Heat Sink-Red Heat Sink-Red OR OR OR

2. Primary Coolant Activity Level 2. RCS Leak Rate 2. Containment Pressure Coolant Activity GREATER Not Applicable GREATER THAN available Unisolable leak exceeding Rapid unexplained 46 PSIG and increasing THAN 300 µCi/gm 1-131 makeup capacity as 60gpm decrease following initial OR equivalent indicated by a loss of RCS increase Containment hydrogen subcooling LESS THAN 21

[40]* degree F OR Containment pressure or sump level response not concentration GREATER THAN OR EQUAL TO 6%

. OR I*

"Adverse containment consistent with LOCA Containment pressure conditions are defined as a . conditions GREATER THAN 23 psig containment pressure with LESS THAN one full greater than 5 psig or train of depressurization containment radiation level equipment operating greater than 1E4 R/Hr.

During adverse

• containment conditions use iCCM to determine RCS subcoollng.

PINGP 6-F-3 F3-2. 1, Rev. 13

TABLE F-1 PINGP Emergency Action Level Fission Product Barrier Reference Table Thresholds For LOSS or POTENTIAL LOSS of Barriers"

• Determine which combination of the three barriers are lost or have a potential loss and use the following key to classify the event. Also an event for multiple events could occur which result in the conclusion that exceeding the Loss or Potential Loss thresholds is imminent (i.e., within 1 to 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />). In this imminent loss situation use judgment and classify as if the thresholds are exceeded.

• UNUSUAL EVENT ALERT SITE AREA EMERGENCY GENERAL EMERGENCY FU1 ANY loss or ANY Potential Loss of FA1 ANY Loss or ANY Potential Loss of FS1 Loss or Potential Loss of ANY two FG1 Loss of ANY two Barriers AND Containment EITHER Fuel Clad OR RCS Barriers Loss or Potential Loss of Third Barrier

~

Fuel Clad Barrier EALS RCS Barrier EALS Containment Barrier EALS LOSS POTENTIAL LOSS LOSS POTENTIAL LOSS LOSS

• POTENTIAL LOSS OR

3. Core Exit Thermocouple Readings 3. Core Exit Thermocouple Readings GREATER THAN 1200 GREATER THAN 700 Not applicable Core exit thermocouples in degree F degree F excess of 1200 degrees F and restoration procedures not effective within 15 minutes OR Core exit thermocouples in excess of 700 degrees F with reactor vessel level below 40% RVLIS Full Range and restoratlol]

procedures not effective within 15 minutes PINGP 6-F-4 F3~2.1, Rev. _13

TABLE F-1 PINGP Emergency Action Level Fission Product Barrier Reference Table Thresholds For LOSS or POTENTIAL LOSS of Barriers*

• Determine which combination of the three barriers are lost or have a potential loss and use the following key to classify the event. Also an event for multiple events could occur which result in the conclusion that exceeding the Loss or Potential Loss thresholds is imminent (i.e., within 1 to 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />). In this imminent loss situation use judgment and classify as if the thresholds are exceeded. *

• UNUSUAL EVENT ALERT SITE AREA EMERGENCY GENERAL EMERGENCY FU1 ANY loss or ANY Potential Loss of FA1 ANY Loss or ANY Potential Loss of FS1 Loss or Potential Loss of ANY two FG1 Loss of ANY two Barriers AND Containment EITHER Fuel Clad OR RCS Barriers Loss or Potential Loss of Third Barrier Fuel Clad Barrier EALS RCS Barrier EALS Containment Barrier EALS LOSS POTENTIAL LOSS LOSS** POTENTIAL LOSS LOSS POTENTIAL LOSS OR OR OR

4. Reactor Vessel Water Level 3. SG Tube Rupture 4. SG Secondary Side Release with P-to-S Leakage Not Applicable Level LESS THAN: SGTR that results in an Not Applicable RUPTURED S/G is also Not applicable

• 40% RVLIS Full Range ECCS (SI) Actuat!on FAULTED outside of (no RCPs) containment

• 30% RVLIS Dynamic OR Head Range (1 RCP) . Primary-to-Secondary leak

• 60% RVLIS Dynamic rate GREATER THAN 10 Head Range (2 RCPs) gpm with nonisolable steam release from affected S/G to the environment OR

5. CNMT Isolation Valves Status After CNMT Isolation Containment Isolation Not Applicable Valve(s) not closed AND Direct pathway to the

-environment exists after Containment Isolation signal OR OR OR

5. Containment Radiation Monitoring 4. Containment Radiation Monitoring 6. Significant Radioactive Inventory in Containment Containment rad monitor Not Applicable Containment rad monitor Not Applicable Not Applicable Containment rad monitor 1 (2) R-48 or 49 reading 1 (2) R-48 or 49 reading reading.1 (2) R-48 or 49 GREATER THAN 200 R/hr GREATER THAN 7 R/hr GREATER THAN 800 R/hr' PINGP 6-F-5 F3-2.1, Rev. 13

TABLE F-1 PINGP. Emergency Action Level Fission Product Barrier Reference Table Thresholds For LOSS or POTENTIAL LOSS of Barriers*

• Determine *which combination of the three barriers are lost or have a potential loss and use the following key to classify the event. Also an event for multiple events could occur which result in the conclusion that exceeding the Loss or Potential Loss thresholds is imminent (i.e., within 1 to 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />). In this imminent loss situation use judgment and classify as if the thresholds are exceeded.

UNUSUAL EVENT ALERT SITE AREA EMERGENCY GENERAL EMERGENCY FU1 ANY loss or ANY Potential Loss of FA1 ANY Loss or ANY Potential Loss of FS1 Loss or Potential Loss of ANY two FG1 Loss of ANY two Barriers AND Containment EITHER Fuel Clad OR RCS Barriers

• Loss or Potential Loss of Third Barrier Fuel Clad Barrier EALS RCS Barrier EALS Containment Barrier EALS LOSS POTENTIAL LOSS LOSS POTENTIAL LOSS LOSS POTENTIAL LOSS OR OR OR

6. Other Indications 5. Other) Indications 7. Other Indications Not Applicable Not Applicable

• Not Applicable Not Applicable Not Applicable Not Applicable OR OR OR

7. Emergency Director Judgment 6. Emergency Director Judgment 8. Emergency Director Judgment Any condition* in the opinion of the Emergency Director that Any condition in the opinion of the Emergency Director that Any condition in the opinion of the Emergency Director indicates Loss or Potential Loss of the Fuel Clad Barrier indicates Loss or Potential Loss of the RCS Barrier that indicates Loss or Potential Loss of the Containment Barrier PINGP 6-F-6 F3-2.1, Rev. 13

• Basis Information For Table F-1 PINGP Emergency Action Level Fission Product Barrier Reference Table FUEL CLAD BARRIER EALs: (1 or 2 or 3 or 4 or 5 or 6 or 7)

The Fuel Clad Barrier is the zircalloy or stainless steel tubes that contain the fuel pellets.

1. Critic~! Safety Function Status RED path indicates an extreme challenge to the safety function. ORANGE path indicates a severe challenge to the safety function.

Core Cooling - ORANGE indicates subcooling has been lost and that some clad damage may occur. Core Cooling-ORANGE path is entered if core exit TCs are less than 1200°F, RCS subcooling based on core exit TCs is less than 21 F [40F] and either:

• No RCPs are running and either core exit TCs are less than 700°F and RVLIS full range is greater than 40%, or core exit TCs are greater than 700°F and RVLIS full range is less than 40%. .

• At least one RCP is running* and RVLIS Dynamic Head Range is less than 60% (2 RCPs) or 30% (1 RCP). [Ref. 1] .

Heat Sink - RED indicates the ultimate heat sink function is under extreme challenge and thus these two items (Core Cooling - ORANGE or Heat Sink - RED) indicate potential loss of the Fuel Clad Barrier. Heat Sink-Red path is entered if wide range level in both S/Gs is less than 50% and total feedwater flow to S/Gs is less than 200 gpm. [Ref. 2] (Note that if feedwater flow to S/Gs is reduced less than 200 gpm due to operator action, the Heat Sink-Red Path is NOT valid and consistent with the 1 (2)FR-H.1 procedure caution, Ref. 17)

Core Cooling - RED indicates significant superheating and core uncovery and is considered to indicate loss of the Fuel Clad Barrier. Core Cooling-RED path is entered if:

• • Core exit TCs are greater than 1200°F, or

• Core exit TCs are greater than 700°F with RCS subcooling based on core exit TCs less than 21 F [40F], RVLIS full range is less than 40% and no RCPs are running Critical Safety Function Status Tree (CSFST) setpoints enclosed in brackets (e.g., [40°F], etc.) are used under adverse containment conditions. Adverse containment condition thresholds apply when containment pressure is greater than 5 psig or containment radiation exceeds 1E+4 R/hr.

[Ref. 1, 8] During adverse containment conditions ERGS Subcooling does not adequately account for instrument uncertainties and the ICCM is to be used when checking RCS Subcooling.

The barrier loss/potential loss occurs when the plant parameter associated with the CSFST path is reached (not when the operator reads the CSFST in the EOP network).

• 2. Primary Coolant Activity Level This vah,.1~ is 300 µCi/gm 1-131 equivalent. Assessment by the NUMARC EAL Task Force indicates that this amount of coolant activity is well above that expected for iodine spikes and corresponds to less than 5% fuel clad damage. This amount of radioactivity indicates significant clad damage and thus the Fuel Clad Barrier is considered lost.

There is no equivalent "Potential Loss" EAL for this item.

PINGP 6-F-7 F3-2.1, Rev. 13

3. Core Exit Thermocouple Readings Core Exit Thermocouple Readings are included in addition to the Critical Safety Functions to include conditions when the CSFs may not be in use (initiation after SI is blocked).

The "Loss" EAL 1200 degrees F reading corresponds to significant superheating of the coolant.

This value correspqnds to the temperature reading that indicates core cooling - RED in Fuel Clad Barrier EAL #1 which is 1200 degrees F. [Ref. 1]

The "Potential Loss" EAL 700 degrees F reading corresponds to loss of subcooling. This value corresponds to the temperature reading that indicates core cooling - ORANGE in Fuel Clad Barrier EAL #1 which is 700 degrees F. [Ref.1] *

'4. Reactor Vessel Water Level There is no "Loss" EAL corresponding to this item because it is better covered by the other Fuel Clad Barrier "Loss" EALs.

The RVLIS values for the "Potential Loss" EAL corresponds to the top of the active fuel under various RCP configurations (2 RCPs running, 1 RCP running, or no RCPs running).

The "Potential Loss EAL is defined by the Core Cooling - ORANGE path. [Ref.1, 2]

5. Containment Radiation Monitoring The 200 R/hr reading is a value which indicates the release bf reactor coolant, with elevated activity indicative of fuel damage, into the containment. [Ref. 9] The reading .is calculated

• assuming the instantaneous release and dispersal of the reactor coolant noble gas and iodine inventory associated with a concentration .of 300 µCi/gm dose equivalent 1-131 into the containment atmosphere. [Ref. 4, 5] Reactor coolant concentrations of this magnitude are several times larger than the maximum concentrations (including iodine spiking) allowed within technical specifications and are therefore indicative of fuel damage. This value is higher than that specified for RCS barrier Loss EAL #4. Thus, this EAL indicates a loss of both the fuel clad barrier and a loss of RCS barrier.

There is no "Potential Loss" EAL associated with this item.

6. Other Indications Not Applicable

7. Emergency Director Judgment

(

This EAL addresses any other factors that are to be used by the Emergency Director in determining whether the Fuel Clad barrier is lost or potentially lost. Such a determination should include imminent barrier degradation, barrier monitoring capability and dominant accident sequences.

PINGP 6-F-8 F3-2.1, Rev. 13

Imminent barrier degradation exists if the degradation will likely occur within two hours based on a projection of current safety system performance. The term "imminent" refers to recognition of the inability to reach safety acceptance *criteria before completion of all checks.

/

Barrier monitoring capability is decreased if there is a loss or lack of reliable indicators.

This assessment should include instrumentation operability concerns, readings from portable instrumentation and consideration of offsite monitoring results.

• Dominant accident sequences lead to degradation of all fission product barriers and likely entry to the CSFSTs. The Emergency Director should be mindful of the Loss of AC power (Station Blackout) and AT_WS EALs to assure timely emergency classification declarations.

The additional bulleted items in the basis for Emergency Director judgment are* a combination of bases information from NEI 99-01 revision 4. The first bulleted item comes from the notes on Table 5-F-1 as well as sectiqns 3.9 and 3.10 of the NEI document regarding "imminent"*barrier loss. The second bulleted item is from the bases of IC SG1, loss of all AC,* regarding degraded barrier monitoring capability that must be considered in this EAL. The third bulleted item also comes from the IC SG2 as well as SG2 (ATWS) regarding the importance of the use of Emergency Director judgment to make anticipatory declarations based on FPB monitoring .

• PINGP . 6-F-9 F3-2.1, Rev. 13

RCS BARRIER EALs: (1 or 2 or 3 or 4 or 5 or 6)

The RCS Barrier includes the RCS primary side and its connections up to and including the pressurizer safety and relief valves, and other connections up to and including the primary isolation valves.

1.. Critical Safety Function Status RED path indicates an extreme challenge to the safety function derived from appropriate instrument readings, and these CSFs indicate a potential loss of RCS barrier.

RCS Integrity-Red path is entered if cold leg temperature decreases greater than 100°F in the last 60 minutes and RCS pressure/cold leg temperature is to the left of Limit A. The combinat.ion of these two conditions indicates the RCS barrier is under extreme challenge. [Ref. 6] Heat Sink-Red path is entered if wide range level in both S/Gs is less than 50% and total feedwater flow to S/Gs is less than 200 gpm. The combination of these two* conditions indicates the ultimate heat sink function is under extreme challenge. [Ref. 2] (Note that if feedwater flow to S/Gs is reduced less than 200 gpm due to operator action, the Heat Sink-Red Path is NOT valid and consistent with the 1 (2)FR-H.1 procedure caution, Ref. 17)

The barrier potential loss occurs when the plant parameter assocjated with the CSFST path is reached (not when the operator reads the CSFST in the EOP network).

There is no "Loss" EAL associated with this item.

2. RCS Leak Rate The "Loss" EAL addresses conditions where leakage from the RCS is greater than available inventory control capacity such that a loss of subcooling has occurred. The loss of subcooling is the fundamental indication that the inventory control systems are inadequate in maintaining RCS pressure and inventory against the mass loss through the leak.

• The "Potential Loss" EAL is based. on the inability to maintain normal liquid inventory within the Reactor Coolant System (RCS) by normal operation of the Chemical and Volume Control System which is normal operation of two low capacity positive displacement variable speed charging pumps discharging to the charging header. Each charging pump has a maximum capacity of 60 gpm [Rev. 7}. An RCS leak rate exceeding the capacity of one charging pump is indicative of a substantial RCS leak. Sixty gpm is used to indicate the Potential Loss which is readily determined by control room staff using either the plant process computer leak rate calculations or control board charging and letdown flow indications.

3. SG Tube Rupture This EAL is intended to address the full spectrum of Steam Generator (SG) tube rupture events in conjunction with Containment Barrier "Loss" EAL #4 and Fuel Clad Barrier EALs. Tt,e "Loss" EAL addresses RUPTURED SG(s) for which the leakage is large en_ough to cause actuation of ECCS 1

(SI). ECCS (SI) actuation is caused by:

• PRZR pressure less than 1830 psig PINGP Either SG pressure less than 530 psig Containment pressure greater than 3.5 psig 6-F-10 F3-2.1, Rev. 13

• • This is consistent to the RCS Barrier "Potential Loss" EAL #2. This condition is described by "entry into E..'.3 required by EOPs". By itself, this EAL will result in the declaration of an Alert. However, if the SG is also FAULTED (i.e., two barriers failed), the declaration escalates to a Site Area Emergency per Containment Barrier "Loss" EAL #4. [Ref. 8]

• There is no "Potential Loss" EAL.

• 4. Containment Radiation Monitoring The 7 R/hr reading is a value which indicates the release of reactor coolant to the containment.

The reading is calculated assuming the instantaneous release and dispersal of the reactor coolant

• noble gas and iodine inventory associated with normal operating concentrations (i.e., within Technical Specifications) into the containment atmosphere. [Ref. 4, 5] This reading is less than that specified for Fuel Clad Barrier EAL #5. Thus, this EAL would be indicative of a RCS leak only.

If the radiation monitor reading increased to that specified by Fuel ~lad Barrier EAL #5, fuel damage would also be indicated. *

• The physical location of the containment radiation monitors is such that radiation from a cloud of released RCS gases can be distinguished from radiation from nearby piping and components containing elevated reactor coolant activity, making the use of these monitors for this EAL classification appropriate.

• There is no "Potential Loss" EAL associated with this item.

• 5. Other Indications Instrumentation used for this EAL is consistent with that used in the RCS integrity EOP. There is no additional applicable indication to use for RCS barrier EALs. [Ref. 6]

.6. Emergency Director Judgment

• This EAL addresses any other factors that are to be used by the Emergency Director in determining whether the RCS barrier is Jost or potentially lost. Such a determination should include imminent barrier degradation, barrier monitoring capability and dominant accident sequences.

• Imminent barrier degradation exists if the degradation will likely occur within two hours based on a projection of current safety system performance. The term "imminent" refers to recognition of the inability to reach safety acceptance criteria before completion of all checks. * *

• Barrier monitoring capability is decreased if there is a loss or lack of reliable indicators.

Thfs assessment should include instrumentation operability concerns, readings from portable ir.istrumentation and consideration of offsite monitoring results.

• Dominant accident sequences lead to degradation of all fission product barriers and likely entry to the CSFSTs. The Emergency Director should be mindful of the Loss of AC power (Station Blackout) and ATWS EALs to assure timely emergency classification declarations .

• PINGP 6-F-11 F3-2.1, Rev. 13

j The additional bulleted items in the basis for Emergency Director judgment are a combination of

• bases information from NEI 99-01 revision 4. The first bulleted item comes from the notes. on Table 5-F-1 as well as sections 3.9 and 3.10 of the NEI document regarding "imminent" barrier loss. The second bulleted item is from the bases of IC SG1, loss of all AC, regarding degraded barrier monitoring capability that must be considered in this EAL. The third bulleted \item also comes from the IC SG2 as well as SG2 (ATWS) regarding the importance of the use of Emergency* Director judgment to make pnticipatory dectarations based on FPB monitoring.

PINGP 6-F-12 F3-2.1, Rev. 13

• CONTAINMENT BARRIER EALs: (1 or 2 or 3 or 4 or 5 or 6 or 7 or 8)

The Containment Barrier includes the containment building, its connections up to and including the outermost containment isolation valves. This barrier also includes the main steam, feedwater, and blowdown line extensions outside the containment building up to and including the outermost secondary side isolation valve.

1. Critical Safety Function Status RED path indicates an \

extreme challenge to the safety function. Containment-Red path is entered if containment pressure is greater than 46 psig. This pressure is the containment design pressure, and thus represents a potential loss of containment. Conditions leading to a containment RED

• path result from RCS barrier and/or Fuel Clad Barrier Loss. Thus, this EAL is primarily a discriminator between Site Area Emergency and General Emergency representing a potential loss of the third barrier. [Ref. 9, 1O]

The barrier potential loss occurs when the plant parameter associated with the CSFST path is reached (not when the operator reads the CSFST in the EOP network).

• There is no "Loss" EAL associated with this item.

2. Containment Pressure Rapid unexplained loss of pressure (i.e., not attributable to containment spray or condensation effects) following an initial pressure increase indicates a loss of containment integrity. USAR Appendix K describes containment pressure response for a bounding LOCA. [Ref. 16]

Containment pressure and sump levels should increase as a result of the mass and energy release into containment from a LOCA: Thus, sump level or pressure not increasing indicates containment bypass and a loss of containment integrity.

• The 46 PSIG for potential loss of containment is based on the containment design pr~ssure. [Ref.

1~ . .

If hydrogen concentration reaches or exceeds 6% in Containment, an explosive mixture exists. If the combustible mixture ignites, loi;s of the *containment barrier could occur. To generate such levels of. combustible gas, an inadequate core cooling situation must already have existed. As described above, this EAL is primarily a discriminator between Site Area Emergency and General Emergency representing a potential loss of the third barrier. [Ref. 3]

The third potential loss EAL represents a potential loss of containment in that the containment heat removal/depressurization system (but not including containment venting strategies) are either lost or performing in a degraded manner, as indicated by containment pressure greater than the setpoint (23 psig) at which the equipment was supposed to have actuated. A full train of depressurization equipment is one containment spray pump and two containment fari coil units.

• This equipment will provide 100% of the required cooling capacity during post-accident conditions.

Each internal containment spray system consists of a spray pump, spray header, nozzles, valves, piping, instruments, and controls to ensure* an operable flow path capable of taking suction I from the RWST upon an ESF actuation signal. [Ref. 11, 12] .

• PINGP 6-F-13 F3-2.1, Rev. 13 L ---

3. Core Exit Thermocouples In this EAL, the restoration procedures are those emergency operating procedures that address .

the recovery of the core cooling critical safety functions. The procedure is considered effective if the temperature is decreasing or if the vessel water level is increa$ing.

Severe accident analyses (e.g., NUREG-1150) have concluded that function restoration procedures can arrest core degradation within the reactor vessel in a significant fraction of the core damage scenarios, and that the likelihood of containment failure is very small in these events.

Given this, it is appropriate to provide a reasonable period to allow function restoration procedures to arrest .the core melt sequence. Whether or not the procedures will be effective should be apparent within 15 minutes. The Emergency Director should make the declaration as so.on as it is determined that the procedures have been, or will be ineffective. The reactor vessel l~vels chosen are consistent with the emergency response guides (EOPS) for PINGP [Ref. 1, 3]

Core exit thermocouple readings of 1200°F represent significant superheating of the coolant. This value corresponds to the temperature reading that indicates core cooling - RED in Fuel Clad Barrier EAL #1. Core exit thermocouple readings in excess of 700°F with reactor vessel level below 40% RVLIS Full Range indicate core exit superheating and core uncovery.

• The conditions in this potential loss EAL represent an imminent core, melt sequence which, if not corrected, could lead to vessel failure and an increased* potential for containment failure. In conjunction with the Core Cooling and Heat Sink criteria in the Fuel and RCS barrier columns, this .

EAL would result in the declaration of a General Emergency -- loss of two ba.rriers and the potential loss of a third. If the function restoration procedures are ineffective, there is no "success" path. [Ref. 1, 3] *

• There is no "Loss" EAL associated with this item.

4. SG Secon.dary Side Release With Primary To Secondary Leakage This "loss" EAL recognizes that SG tube leakage can represent a bypass of the containment barrier as well as a loss of the RCS barrier. The first "loss" EAL addresses the condition in which a RUPTURED steam generator is also FAULTED. This condition represents a bypass of the RCS and containment barriers. In conjunction with RCS Barrier "loss" EAL #3, this would always result in the declaration of a Site Area Emergency. A faulted SIG means the existence of secondary side leakage that results in an uncontrolled lowering in steam generator pressure or the steam generator being completely depressurized. A ruptured SIG means the existence of primary-to-secondary leakage of a magnitude sufficient to require or cause a reactor trip and safety injection.

Confirmation should be based on diagnostic activities consistent with E-0, Reactor Trip or Safety Injection. [Ref. 8]

The second "loss" EAL addresses SG . tube leaks that exceed 1O *gpm in conjunction with a nonisolable release path to the environment from the affected steam generator. The threshold for establishing the nonisolable secondary side release is intended to be a prolonged release of '-

radioactivity from the RUPTURED steam generator directly to the environment. This could be expected to occur when the main condenser is unavailable to accept the contaminated steam (i.e.,

SGTR with concurrent loss of offsite power and the RUPTURED steam generator is required for plant cooldown or a stuck open relief valve). If the main condenser is available, there may be releases via air ejectors, gland seal exhausters, and other similar controlled, and often monitored, pathways.

• These pathways do not meet the- intent of a nonisolable release path to the environment. These minor releases are assessed using Abnormal Rad Levels I Radiological

• Effluent ICs. [Ref. 8]

PINGP 6-F-14. F3-2.1, Rev. 13

____J

• It should be realized that the two "loss" EALs described above could be considered redundant.

This was recognized during *the development process. The inclusion* of an EAL that uses

• Emergency Procedure commonly used terms like "ruptured and faulted" adds to the ease of the classification process and has been included based on this human factor concern.

A pressure boundary leakage of 10 gpm is used as the threshold in IC SU5.1, RCS Leakage, and is deemed appropriate for this EAL. For smaller breaks, not exceeding the normal charging capacity threshold in RCS Barrier "Potential Loss" EAL #2 (RCS Leak Rate) or not resulting in ECCS actuation in EAL #3 (SG Tube Rupture), this EAL results in a UE. For larger breaks, RCS barrier EALs #2 and #3 would result in an Alert. For SG tube ruptures which may involve multiple steam generators or unisola!>le secondary line breaks, this EAL would exist in conjunction* with RCS barrier "Loss" EAL #3 and would result in a Site Area Emergency. Escalation to General Emergency would be based on "Potential Loss" of the Fuel Clad Barrier.

5. Containment Isolation Valve Status After Containment Isolation This EAL is intended to address incomplete containment isolation that allows direct release to the environment. It represents a loss of the containment barrier.

Irregardless of the reason for the containment isolation signal, if a containment isolation signal does not result in Containment Isolation Valve(s) to close and a direct pathway to the environment exists after Containment Isolation signal,* then FPB EAL Containment Loss 5 conditions are met and will result in at *(east an UE Classification. For example, an unsuccessful automatic containment isolation signal would result in a loss of the containment barrier. If the failure of the automatic containment isolation signal is followed by a successful manual containment isolation

• signal, subsequent escalations would have the containment barrier intact.

The use of the modifier "direct" in defining the release path discriminates against release paths through interfacing liquid systems. The existence of an in-line charcoal filter does not make a release path indirect since the filter is not effective at removing fission noble gases. Typical filters have an efficiency of 95-99% removal of iodine. Given the magnitude of the core inventory of iodine, significant releases could still occur. In addition, since the fission product release would be driven by boiling in the reactor vessel, tl:le high humidity in the release stream can be expected to render the filters ineffective in a short period.

• There is no "Potential Loss" EAL associated with this item.

6. Significant Radioactive Inventory in Containment The 800 R/hr reading is a value which indicates significant fuel damage well in excess of the EALs associated with both loss of Fuel Clad and loss of RCS Barriers. [Ref. 4, 5] A major release of radioactivity requiring offsite proteGtive actions from core damage is not possible unless a major failure of fuel cladding allows radioactive material to be released from the core into the reactor coolant.

• Regardless of whether containment is challenged, this amount of activity in containment, if released, could have such severe consequences that it is prudent to treat this as a potential loss of containment, such that a General Emergency declaration is warranted. NUREG-1228, "Source Estimations During Incident Response to Severe Nuclear Power Plant Accidents," indicates that such conditions do not exist when the amount of clad damage is less than 20%. Accordingly, the

• EAL threshold corresponds to clad damage of 20%. [Ref. 4, 5]

I There is no "Loss" EAL associated with this item.

PINGP 6-F-15 F3-2.1, Rev. 13

7. Other (Site-Specific) Indications Instrumentation used for this EAL is consistent with that used in the Containment integrity EOP:

There is no additional applicable indication to use that may unambiguously indicate loss or potential loss of the containment barrier. Venting of the containment during an emergency is not used as a means of preventing catastrophic failure. [Ref. 9]

8. Emergency Director Judgment This EAL addresses any other factors that are to be used by the Emergency Director in determining whether the Containment barrier is lost or potentially lost. Such a determination should include imminent barrier degradation, barrier monitoring capability .and dominant accident..

sequences.

• Imminent barrier degradation exists if the degradation will likely occur within two hours based on a projection of current safety system performance. The term "imminent" refers to recognition of the inability to reach safety acceptance criteria before completion of all checks.

• Barrier monitoring capability is decreased if there is a loss or lack of reliable indicators.

This assessment should include instrumentation operability concerns, readings from portable instrumentation and consideration of offsite monitoring results. *

• Dominant accident sequences lead to degradation of all fission product barriers and likely entry to the CSFSTs. The Emergency Director should be mindful of the Loss of AC power *

(Station Blackout) and ATWS EALs to assure timely emergency classification declarations.

The additional bulleted items in the basis for Emergency Director judgment are a combination of bases information from NEI 99-01 revision 4. The first bulleted item comes from the notes on Table 5-F-1 as well as sections 3.9 and 3:10 of the NEI document regarding "imminent" barrier loss. The second bulleted item is from the bases of IC SG1, loss of all AC, regarding degraded barrier monitoring capability* that must be considered in this EAL. The third bulleted item also comes from the IC SG2 as well as SG2 (ATWS) regarding the importance of the use of Emergency Director judgment to make anticipatory declarations based on FPB monitoring.

PINGP Basis Reference{s):

1. F-0.2 Core Cooling

2. F-0.3 Heat Sink

3. FR-C.1 Response to Inadequate Core Cooling

4. *F3-17 Core Damage Assessment

5. Memo to EAL Upgrade Project File from Mel Agen dated 7/31/04 "Containment Rad Monitors

& Fuel Cladding Damage Based on USAR".

6. F-0.4 Integrity

7. USAR Section 10.2.3

• PINGP 6-F-16 F3-2.1, Rev. 13

• 8. E-0 Reactor Trip or Safety Injection

9. F-0.5 Containment

10. USAR Section 5.2.1

11. Technical Specifications Table 3.3.2-1

12. Technical Specifications B3.6.5

13. Memo to EAL Upgrade Project File from Mel Agen dated 10/11/04 "R-9 Rad Monitors & Fuel Cladding Damage Based on USAR"

14. USARSection 10.2.3.3.7.

15. USAR Appendix D

16. USAR Appendix K

• 11. FR-H.1, Response to Loss of Secondary Heat Sink

• PINGP 6-F-17 F3-2.1, Rev. 13

CD-0676 Controlled Document Transmittal REV.2 Report Date: 12/18/2017 To US NRC C/0 PAM JOHNSON (P.I.) From C-DOC CNTRL-PI Facility Pl Address 1717 WAKONADE DR Address PAMELA JOHNSON WELCH, MN 55089 DOCUMENT CONTROL DESK US NRC Vital NO

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Prairie Island Nuclear Generating Plant EMERGENCY ACTION LEVEL MATRIX

- GENERAL EMERGENCY SITE AREA EMERGENCY ALERT UNUSUAL EVENT HOT & COLD

• RG 1 RG1 .1 NOTE :

Offsne Dose Resulting from an Actual or Imminent Release of Gaseous Radioactivity Exceeds 1000 mRem TEDE or 5000 m Rem Thyroid COE for the Actual or Projected Duration of the Release Using Actual Meteorology.

1  ! 2 3 4 5 6  ! DEF !

If dose assessment results are available at the time of declaration, the classification should be based on RG1 .2 instead of RG1 .1.

While necessary declarations should not be delayed awaiting resu lts ,

the dose assessment should be initiated / completed in order to RS1 RS1 .1 NOTE :

Offsite Dose Resulting from an Actual or Imminent Release of Gaseous Radioactivfy Exceeds 100 mRem TEDE or 500 mRem Thyroid COE for the Actual or Projected Duration of the Release.

2 3 4 5 6  ! DEF !

If dose assessment results are ava ilab le at the time of declaration ,

the classification should be based on RS1 .2 instead of RS1 .1.

While necessary declara tions should not be delayed awaiting results ,

the dose assessment should be initiated / completed in order to RA1 RA 1.1 Any UNPLANNED Release of Gaseous or Liquid Radioactivfy to the Environment that Exceeds 200 Times the Offsite Dose Calculation Manual Specification for 15 Minutes or Longer.

2  ! 3 4 5 6 I DEF VALID reading on any effluent monitor that exceeds 200 Times the alarm setpoint established by a current radioactivity discharge perm it for 15 minutes or longer.

OR I

RU1 RU1 .1 Any UNPLJ'.NNED Release of Gaseous or Liquid Radioactivfy to the Envir->nment that Exceeds Two Times the Offsne Dose Calculation Manual Specification for 60 Minutes or Longer.

2 3 4 5 setpoint established by a current radioactivity discharge permit for 60 minutes or longer.

6  ! DEF VALID reading on any effluent monitor that exceeds two times the alarm determine if the classification should be subsequently escalated . determine if the classification should be subsequently escalated .

VALID reading on effluent monitor R-18 that exceeds 900,000 cpm for 15 minutes or longer.

RU1 .2 I 2 3 4 5 6  ! DEF VALID reading on one or more monitors listed in Table R- 1 that exceeds VALID reading on one or more monitors listed in Table R-1 that exceeds VALID reading on one or more of the following radiation monitors or expected to exceed column "GE" for 15 minutes or longer: or is expected to exceed column "SAE" for 15 minutes or longer: (Table R-1) that exceeds the reading shown for 60 minutes or longer:

Offsite Rad Offslte Rad Conditions RG1 2 I 2 3 4 5  ! 6 I DEF RS1 .2 I 2 3 4 5 6  ! DEF  ! RA1 .2 2 3 4 5 6  ! DEF RU1 .3 I 2 3 4 5 6  ! DEF  ! Conditions Dose assessment using actual meteorology indicates doses GREATER Dose assessment using actual meteorology indicates doses GREATER VALID reading on one or more of the following radiation monitors Confirmed sample analysis for gaseous or liquid release indicates THAN 1000 mRem TEDE or 5000 mRem thyroid CDE at or beyond the THAN 100 mRem TEDE or 500 mRem thyroid COE at or beyond the (Table R-1 ) that exceeds the reading shown for 15 minutes or longer: concentrations or release rates , with a release duration of 60 minutes site boundary. site boundary. or longer, in excess of two times ODCM specification.

RA 1.3 I 2 3 4 5 6  ! DEF !

RG1 .3  ! 2 3 4 5 6 DEF RS1 .3 I 2 3 4 5 6  ! DEF Confirmed sample anal ysis for gaseous or liquid release indicates Field survey results indicate closed window dose rates exceeding Field survey results indicate closed window dose rates exceeding concentrations or release rates, with a release duration of 15 minutes 1000 mR/hr expected to continue for more than one hour. at or beyond 100 mR/hr expected to continue for more than one hour. at or beyond or longer, in excess of 200 Times ODCM specification.

site boundary; the site boundary; OR OR Analyses of field survey samples indicate thyroid COE of 5000 mRem for Analyses of field survey samples indicate thyroid COE of 500 mRem for one hour of inhalation. at or beyond site boundary. one hour of inhalation. at or beyond the site boundary.

RA2 Damage to Irrad iated Fuel or Loss of Water Level that Has or RU2 Unexpected Increase in Plant Radiation.

Abnormal Abnormal Will Result in the Uncovering of Irradiated Fuel Outside the Rad Reactor Vessel. Rad Release RA2.1 2 3 4 5 6  ! DEF ! RU2.1 2 3 4 5 6 DEF  ! Release A VALID alarm on one or more of the following radiation monitors : VALID indication of uncontrolled water level decrease in the reactor Rad Rad

• R-25 or R-31 SFP Air Monitor (HI Alarm ) refueling cavity, spent fuel pool. or fuel transfer canal with all irrad iated Effluent Effluent

• R-5 Fuel Handling Area Monitor reading (HI Alarm) fuel assemblies remaining covered by water as indicated by level

• R-28 New Fuel Pool Criticality Area Monitor (HI Alarm) LESS THAN SFP low water level alarm , Refueling Canal Level , or

• 1(2) R-1 1 CtmVSBV Air Particulate Monitor (HI Alarm) visual observation (752.5 feet elevation);

• 1(2) R-12 CtmVSBV Radio Gas Monitor (HI Alarm ) AND

• 1(2) R-2 Containment Vessel Area Monitor (HI Alarm ) Any UNPLANNED VALID Area Radiation Monitor reading increases as indicated by:

• R-5 Fuel Handling Area Monitor reading RA2.2 I 2 3 4 5 6  ! DEF

• R-28 New Fuel Pool Criticality Area Monitor Water level LESS THAN 10 feet above an irradiated fuel assembly

• 1(2) R-2 Containment Vessel Area Monitor for the reactor refueling cavity, spent fuel pool and fuel transfer canal

• Other Portable Area Radiation Monitoring Instrumentation Onsite Rad that will result in irrad iated fuel uncovering Onsite Rad Conditions RU2.2 I 2 3 4 5 6  ! DEF Conditions RA3 Release of Radioactive Material or Increases in Radiation Levels Any UN PLANNED VALID Area Radiation Monitor reading increases by Within the Facility That Impedes Operation of Systems Required a factor of 1000 over normal* levels .

to Maintain Safe Operations or to Establish or Maintain Cold Shutdown.

• Normal levels can be considered as the highest reading in the past RA3.1 2 3 4 5 6  ! DEF ! twenty-four hours excluding the current peak value.

VALI D radiation monitor readings GR EATER THAN 15 mR/hr in areas requiring continuous occupancy to maintain plant safety functions :

Control Room (Rad monitor R-1 );

OR Central Alarm Station (by portable radiation monitoring instrumentation).

RA3.2 I 2 3 4 5 6  ! DEF Any VALID radiation monitor reading GREATER THAN 1 R/hr in Table R-1 Effluent Monitor Classification Thresholds areas requiring infrequent access to maintain plant safety functions Monitor GE SAE Alert UE (Table H-1 ).

Gaseous CPM CPM 1(2) R-50 High Range Stack Gas Monitor 43000 mR/ hr 4300 mR/hr N/A N/A 1R-22* Shield Building Vent Rad Monitor N/A N/A 1so.000*1 1.s Es 1,600*/ 1.6 E3 ----

Table H-1 Plant Areas 2R-22* Shield Building Vent Rad Monitor N/A N/A 100,000*t 1 ES 1.000*11 E3 1R-30* & 1R- 37" Unit 1 Aux. Building Vent Rad Monitors N/A N/A 100,000*t 1 ES 1.000*11 E3 Area HU16* HU2.1* HA1 .2 HA1.3 HA1.4 HA1 .5 HA2 .1 HA3_1* HA3.2* RA3.2 2R-30* Unit 2 Aux. Building Vent Rad Monitors N/A N/A 100,000*t 1 E5 1.000*1 1 E3 2R-37" Unit 2 Aux. Building Vent Ra d Monitors N/A N/A 120,000*1 1.2 E5 1,200*1 1.2 E3 - Shield/Containment Building X X X X X X X X X R-35* Radwaste Building Vent Rad Monitor N/A N/A 100,000*t 1 ES 1,000*11 E3 - Auxiliary Building X X X X X X X X X X R-25* & R-31

• S ent Fuel Pool Vent Rad Monitors N/A N/A 800,000*1 8 ES 8,000*15 E3 - D5/D6 Diesel Generator Building X X X X X X X X X X Liquid - Plant Screenhouse X X X X X X X X X X R-18* Waste Effluent Liquid Monitor N/A N/A 900,000*/ g ES 30.000*1 3 E4 - Control Room X X X X X X X X X 1R-19* SG Slowdown Radiation Monitor N/A N/A 100,000*t 1 ES 1.000*11 E3 - Relay Room X X X X X X X X X X 2R-19* SG Slowdown Radiation Monitor N/A N/A 60,000*t 6 E4 600*/6 E2 - Turbine Building X X X X X X X X X X R-21 Circ Water Dischar e Monitor N/A N/A 800,000/ 8 ES 8,000/ 8 E3 - Condensate Storage Tanks X X X X Notes : 1) ERCS EAL Alarms indicate an EA L th res ho ld May have been exceeded . Further evaluation of the rad iation mo nitor reading is requ ired to determine if the EAL thres hold is exceeded . 2 )

• Appl ies w hen Effluent d ischarge not isolated.

• Also consider areas contiauous to these.

PINGP 1576, Rev. 10 Doc Type/Sub Type: EP/EVT Retention: Lifetime + Page 1 of 8

Prairie Island Nuclear Generating Plant EMERGENCY ACTION LEVEL MATRIX

• HA 1.2 Natural and Destructive Phenomena Affecting the Plant VITAL AREA.

1 2

2  !

3 3  !

4 4  !

5 Seismic Event GREATER THAN Operating Basis Earthquake (OBE) 5  !

Tornado or high winds GR EATER THAN 95 mph within PROTECTE D AREA boundary and resulting in VISIBLE DAMAG E to any of the 6

6 as indicated by "OBE Exceedance" alarm on Seismic Monitoring Panel.

DEF DEF

! HU1 .2 Natural and Destructive Phenomena Affecting the PROTECTED AREA.

1 I 2 2

3 3

4 4

5 5

6 Earthquake felt in plant as indicated by VALID "Event" alarm on Seismic Monitoring Panel.

!  ! 6 i Report by plant personnel of tornado or high winds GREAT ER THAN 95 mph striking within PROTECTED AR EA boundary.

DEF DEF i following plant structures / equipment or Control Room indication of degraded performance of those systems (Table H-1 ).

HU1 .3  ! 1  ! 2  ! 3  ! 4  ! 5  ! 6 i DEF  !

Vehicle crash into plant structures or systems within PROTECTED HA1.3  ! 2 3 4 5 6 I DEF AREA boundary.

Vehicle crash within PROTECTED AREA boundary and resulting in HU1 .4  ! 1  ! 2 3 4 5 6 i DEF VISIBLE DAMAGE to any of the following plant structures / equipment Report by plant personnel of an unanticipated EXPLOSION within Destructive therein or Control Room indication of degraded performance of those PROTECTED AREA boundary resulting in VISIBLE DAMAGE to Phenomenon None None systems (Table H-1) . permanent structure or equipment.

HA 1.4  ! 1 2 3 4 5 6 i DEF ! HU1 .5  ! 2 3 4 5 6 i DEF Turbine failure-generated missiles result in any VISI BLE DAMAG E Report of turbine fa ilure resulting in casing penetration or damage to to or penetration of any of the following plant areas (Table H-1). turbine or generator seals .

HA 1.5  ! 1  ! 2  ! 3  ! 4  ! 5  ! 6  ! DEF ! HU1 .6  ! 1  ! 2  ! 3 4 5 6 i DEF Uncontrolled flooding in any Table H- 1 area of the plant that results in Uncontrolled flooding in following areas of the plant that has the degraded safety system performance as indicated in the Control Room potential to affect safety related equipment needed for the current or that creates industrial safety hazards (e.g .. electric shock) that operating mode (Table H-1 ).

precludes access necessary to operate or monitor safety equipment. HU1.7  ! 1 2 3 4 5 6 i DEF i 2 3 4 5 6  ! DEF ! High or low river water level occurrences affecting the PROTECTED High or low river water level occurrences affecting the PROTECTED AREA as indicated by:

AREA as indicated by: River intake level GREATER THAN 692 ft MSL; River intake level GREAT ER THAN 698 ft MSL; OR OR River intake level LESS THAN 669.5 ft MSL.

River intake level LESS THAN 666.5 ft MSL.

HA2 FIRE or EXPLOSION Affecting the Operability of Plant Safety HU2 FIRE Within PROTECTED AREA Boundary Not Extinguished Systems Required to Establish or Maintain Safe Shutdown. Within 15 Minutes of Detection.

Fire or HA2.1  ! 1  ! 2  ! 3  ! 4 I 5  ! 6  ! DEF  ! HU2.1 2 3 4 5 6  ! DEF  !

Explosion None None FIRE or EXPLOSION in any of the following areas (Table H-1 ): FIRE in buildings or areas contig uous (in actual contact with or AND immediately adj acent) to any Table H-1 area not extinguished within Affected system parameter indications show degraded performance 15 minutes of control room notification or verification of a control or plant personnel report VISIBLE DAMAGE to permanent structures room alarm.

or equipment within the specified area.

HA3 Release of Toxic or Flammable Gases Within or Contiguous HU3 Release of Toxic or Flammable Gases Deemed Detrimental to to a VITAL AREA Which Jeopardizes Operation of Systems Normal Operation of the Plant.

Required to Maintain Safe Operations or Establish or Maintain Safe Shutdown.

Toxic and None None HA3.1 2 3 4 5 6  ! DEF  ! HU3 .1 2 3 4 5 6 DEF  ! Toxic and Flammable Report or detection of toxic gases within or contiguous to Table H- 1 Report or detection of toxic or flammable gases that has or could enter Flammable Gas areas in concentrations that may result in an atmosphere IMMEDIATELY the site area boundary in amounts that can affect NORMAL PLANT Gas DANGEROUS TO LIFE AND HEALTH (IDLH). OPERATIONS.

HA3.2  !  ! 2  ! 3  ! 4  ! 5 6 i DEF i I HU3.2 ..---,,---2--,...--3--.-4--.--5--.--6-...-D_E_F....,I I Report or detection of gases in concentration GREATE R THAN the Report by Local, County or State Officials for evacuation or sheltering LOWER FLAMMABILITY LIMIT within or contiguous to Table H-1 areas. of site personnel based on an offsite event.

Table H-1 Plant Areas Area HU1 .6* HU2.1* HA12 HA1.3 HA1.4 HA1 .5 HA2.1 HA3.1' HA3.2* RA3.

- Shield/Containment Building X X X X X X X X X

- Auxiliary Building X X X X X X X X X X

- D5/D6 Diesel Generator Building X X X X X X X X X X

- Plant Screenhouse X X X X X X X X X X

- Control Room X X X X X X X X X

- Relay Room X X X X X X X X X X

- Turbine Building X X X X X X X X X X

- Condensate Storage Tanks X X X X

• Also consider areas conti uous to these .

PINGP 1576, Rev. 10 Doc. Type/Sub Type: EP/EVT Retention: Lifetime + Page 2 of 8

Prairie Island Nuclear Generating Plant EMERGENCY ACTION LEVEL MATRIX HOSTILE ACTION Resulting in Loss of Physical Control of the Facility_

1  ! 2 3 4 5 6 A HOSTILE ACTION has occurred such that plant personnel are i DEF unable to operate equipment required to maintain safety functions.

HG1 .2  ! 1  ! 2  ! 3  ! 4  ! 5  ! 6  ! DEF A HOSTILE ACTION has caused failure of Spent Fuel Cooling Systems and IMMINENT fuel damage is likely for a freshly off-loaded reactor core in pool.

2 3 4 5  !

A HOSTILE ACTION is occurring or has occurred within the PROTECTED AREA as reported by Shift Security Supervision.

6 DEF or Airborne Attack T hreat.

2  ! 3 4 5 A HOSTILE ACTION is occurring or has occurred within the OWNER CONTROLLED AREA as reported by Security Shift Supervision.

HA4.2 ..

6  ! DEF

, _,-,---2--,,---3--,,---4--,.---5---,,---6---,I-D-EF---,I A validated notification from NRC of an airliner attack threat within HU4.2 Confirmed SECURITY CONDITION or Threat Which Indicates a Potential Degradation in the Level of Safety of the Plant.

1 1

2  ! 3 2 _ 3 _ 4 _ 5

! 4 A SECURITY CONDITION that does NOT involve a A credible PINGP security threat notification.

5  !

HOSTILE ACTION as rer orted bl Securitr Shift Supervision.

6 6

DEF DEF 30 minutes of the site. 4 5 6 DEF Security A validated notification from NRC providing information of an aircraft threat.

Hazards Continued HS2 Control Room Evacuation Has Been Initiated and Plant Control HA5 Control Room Evacuation Has Been Initiated.

Cannot Be Established .

HS2.1 2 3 4  ! 5 6 DEF HAS.1 2 3 4 5 6 DEF  !

None Control room evacuation has been initiated; Entry into 1(2)C1 .3 AOP-1 Shutdown from Outside the Control Room None AND or F-5 Appendix B Control Room Evacuation (Fire) for control room Control of the plant cannot be established per 1(2)C1 .3 AOP-1 , evacuation.

Shutdown from Outside the Control Room or F-5 Appendix B, Control Room Evacuation (Fire) within 15 minutes.

HG2 Other Conditions Existing Which in the Judgment of the HS3 Other Conditions Existing Which in the Judgment of the HA6 Other Conditions Existing Which in the Judgment of the HU5 Other Conditions Existing Which in the Judgment of the Emergency Director Warrant Declaration of General Emergency. Emergency Director Warrant Declaration of Site Area Emergency Director Warrant Declaration of an Alert. Emergenc,y Director Warrant Declaration of a UE.

Emergency .

Emergenc y HG2.1  ! 2 3 4 5 6 i DEF i HS3.1 2 3 4 5 6 i DEF  ! HA6.1 2 3 4 5 6 i DEF  ! HUS.1 i 2 3 4 5 6 DEF  ! Emergency Director Other conditions exist which in the judgment of the Emergency Director Other conditions exist which in the judgment of the Emergency Director Other conditions exist which in the judgment of the Emergency Director Other conditions exist which in the judgment of the Emergency Director Director Judgment indicate that events are in process or have occurred which involve actual indicate that events are in process or have occurred which involve actual indicate that events are in process or have occurred which involve actual indicate that events are in process or have occurred which indicate a Judgment or imminent substantial core degradation or melting with potential for or likely major failures of plant functions needed for protection of the public. or likely potential substantial degradation of the level of safety of the plant. potential degradation of the level of safety of the plant. No releases of loss of containment integrity. Releases can be reasonably expected to Any releases are not expected to result in exposure levels which exceed Any releases are expected to be limited to small fractions of the EPA radioactive material requiring offsite response or monitoring are expected exceed EPA Protective Action Guideline exposure levels offsite for more EPA Protective Action Guideline exposure levels beyond the site boundary. Protective Action Guideline exposure levels. unless further degradation of safety systems occurs.

than the immediate site area .

Table H-1 Plant Areas Area HU1 .6" HU2.1" HA1 .2 HA1.3 HA1.4 HA1 .5 HA2 .1 HA3.1" HA3.2" RA3.

- Shield/Containment Building X X X X X X X X X

- Auxiliary Building X X X X X X X X X X

- 05/06 Diesel Generator Building X X X X X X X X X X

- Plant Screenhouse X X X X X X X X X X

- Control Room X X X X X X X X X

- Relay Room X X X X X X X X X X

- Turbine Building X X X X X X X X X X

- Condensate Storage Tanks X X X X

• Also consider areas conti uous to these .

PINGP 1576, Rev. 10 Doc. Type/Sub Type: EP/EVT Retention: Lifetime + Page 3 of 8

' Prairie Island Nuclear Generating Plant EMERGENCY ACTION LEVEL MATRIX HOT

• Prolonged Loss of All Offsite Power and Prolonged Loss of All Onsite AC Power to Safeguards Buses I 2 3 4 Loss of power to or from Transformers CT-11 , CT-12 , 1RY, and 2RY that results in a loss of all offsite power to both Safeguards Buses 15 and 16 (25 and 26);

AND Failure of Diesel Generators D1 and D2 (D5 and DB) to supply power to Safeguards Buses.

2 3 4 I Loss of power to or from Transformers CT-11 , CT-12 , 1RY, and 2RY that results in a loss of all offsite power to both Safeguards Buses 15 and 16 (25 and 26);

AND Failure of both Diesel Generators D1 and D2 (D5 and DB) to supply SA5.1  !

AC power capability to Safeguards Buses reduced to a single power source for GREATER THAN 15 minutes such that any additional single failure would result in station blackout.

2 3 4 AC power capability to Safeguards Buses 15 and 16 (25 and 26) reduced to only one of the following sources for GREATER THAN 15 minutes :

• Transformer CT-11 ;

• Transformer CT-12; Loss of All Offsite Power to Safeguards Buses for GREATER THAN 15 Minutes.

2 3 4 I Loss of power to or from Transformers CT-11 , CT-12, 1RY, and 2RY that results in a loss of all offsite power to both Safeguards Buses 15 and 16 (25 and 26) for GREATER THAN 15 minutes; AND Two Diesel Generators (D1 , D2, D5, DB) are supplying power to Los s of to Safeguards Buses 15 and 16 (25 and 26); power to Safeguards Buses 15 and 16 (25 and 26);

• Transformer 1RY; Safeguards Buses 15 and 16 (25 and 26). Loss of Power AND AND

• Transformer 2RY; Power Either of the following : Failure to restore power to Safeguards Bus 15 or 16 (25 or 26) within

• Diesel Generator D1 (D5);

a. Restoration of Safeguards Bus 15 or 16 (25 or 26) within 15 minutes from the time of loss of both offsite and onsite AC power.

• Diesel Generator D2 (DB);

4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> is not likely; AND OR SS3 Loss of All Vital DC Power. Any additional single fa ilure will result in station blackout.

b. Continuing degradation of core cooling based on Fission Product Barrier monitoring as indicated by Core Cooling-RED SS3.1 i 2 3 4  !

or ORANGE path. Loss of all Safeguards DC power based on LESS THAN 112 VDC on 125VDC Panels 11 and 12 (21 and 22) for GREATER THAN 15 minutes .

SG2 Failure of the Reactor Protection System to Complete an SS2 Failure of Reactor Protection System Instrumentation to SA2 Failure of Reactor Protection System Instrumentation Automatic Trip and Manual Trip was NOT Successful and Complete or Initiate an Automatic Reactor Trip Once a to Complete or Initiate an Automatic Reactor Trip Once a There is Indication of an Extreme Challenge to the Ability Reactor Protection System Setpoint Has Been Exceeded Reactor Protection System Setpoint Has Been Exceeded to Cool the Core. and Manual Trip Was NOT Successful. and Manual Trip Was Successful.

SG2.1 I 2  ! SS2.1 i 2  ! SA2.1  ! 2 3 lndication(s) exist that automatic and manual trip were NOT lndication(s) exist that automatic and manual trip were NOT NOTE : A failed manual trip followed by a successful manual trip successful in reducing power to LESS THAN 5%; successful in reducing power to LESS THAN 5%. reducing reactor power to less than 5% meets this EAL.

AND RPS Either of the following : RPS lndication(s) exist that a Reactor Protection System setpoint was Failure a. Core cooling is extremely challenged as indicated by Failure exceeded; Core Cooling - RED path; AND None OR System RPS automatic trip did not reduce power to LESS THAN 5%; System

b. Heat removal is extremely challenged as indicated by Malfunct. AND Malfunct.

Heat Sink - RED path.

Any of the following operator actions are successful in reducing power to LESS THAN 5% , Manual Control Board :

• Reactor Trip

• AMSAC/DSS Actuation

• Turbine Trip SS4 Complete Loss of Heat Removal Capability SU2 Inability to Reach Required Shutdown Within Technical Specification Limits.

Inability to Reach or SS4.1 I 2 3 4  ! SU2.1  ! 1  ! 2  ! 3 4 Inability to Reach or None Loss of core cooling and heat sink as indicated by: None Plant is not brought to required operating mode within Technical Maintain Maintain

a. Core Cooling - RED path; Specifications LCO Action Statement Time.

Shutdown Shutdown AND Conditions Conditions

b. Heat Sink - RED path.

SS6 Inability to Monitor a SIGNIFICANT TRANSIENT in Progress. SA4 UNPLANNED Loss of Most or All Safety System Annunciation SU3 UNPLANNED Loss of Most or All Safety System Annunciation or Indication in Control Room With Either (1) a SIGNIFICANT or Indication in the Control Room for Greater Than 15 minutes.

TRANSIENT in Progress, or (2) Compensatory Non-Alarming Indicators are Unavailable.

SS6.1 i 2 3 4  ! SA41 i 2 3 4 I SU3.1 I 2 3 4  !

Loss of most (approximately >75%) or all annunciators associated UNPLANNED loss of most (approximately >75%) or all annunciators UNPLANNED loss of most (approximately >75%) or all annunciators or with safety systems : or indicators associated with safety systems for GREATER THAN indicators associated with safety systems for GREATER THAN None

• Main Control Boards A, B-1 (2), C-1 (2), D-1(2), E-1(2), F-1 (2), 15 minutes: 15 minutes:

Inst./ G-1 (2) NIS Racks I, II , Ill , IV, and ERCS Alarms;

• Main Control Boards A, B-1 (2), C-1 (2), D-1(2), E-1 (2), F-1(2),

• Main Control Board A, B- 1(2), C-1 (2), D-1 (2), E-1 (2), F-1(2), G-1(2) Inst. /

Comm. AND G-1(2) NIS Racks I, II , Ill , IV, and ERCS Alarms; NIS Racks I, II, Ill , IV, and ERCS Alarms. Comm .

A SIGNIFICANT TRANSIENT in progress; AND SUB UNPLANNED Loss of All Onsite or Offsite Communications AND Either of the following :

Capabilities.

a. A SIGNIFICANT TRANSIENT in progress; Compensatory non-alarming indications are unavailable; AND OR SU6.1  ! 1  ! 2 3 4

b. Compensatory non-alarming indications are unavailable. Loss of all Table C-1 onsite communications capability affecting the Indications needed to monitor the ability to shut down the reactor ,

ability to perform routine operations.

maintain the core cooled, maintain the reactor coolant system intact, and maintain containment intact are unavailable. sus.2 I 1 I 2 I 3 I 4 I Loss of all Table C-2 offsite communications ca abilit .

Table C-1 Onsite Communications Systems Table C-2 Offsite Communications System Sound Powered Phones Plant Telephone Network Plant Pag ing System Plant Radio System (dedicated offsite channels)

Plant Telephone Network ENS Network Plant Radio System PINGP 1576, Rev. 10 Doc. Type/Sub Type: EP/EVT Retention: Lifetime + Page 4 of 8

Prairie Island Nuclear Generating Plant EMERGENCY ACTION LEVEL MATRIX HOT

• Fuel Clad Degradation None None None SU4.1  !

fuel clad degradation.

SU4.2  ! 1  !

2 2

3 3

! 4 Radiation Monitor 1(2)R-9 GREATER THAN 1.2 R/hr indicating 4

Coolant sample activity GREATER TH AN Technical Specification 3.4.17 Condition C allowable limits indicating fuel clad degradation.

Fuel Clad Degradation SU5 RCS Leakage.

System System Malfunct.

2 3 4 I Malfunct.

RCS None None None Unidentified or pressure boundary leakage GREATER THAN 10 gpm. RCS Leakage Leakage sus.2 I 2 3 4 I Identified leakage GREATER THAN 25 gpm.

SUB Inadvertent Criticality.

Inadvertent Inadvertent Critical ity None None None 3 4 I Criticality An UNPLANNED sustained positive startup rate observed on nuclear instrumentation.

Table C-1 Onsite Communications Systems Table C-2 Offsite Communications System Sound Powered Phones Plant Telephone Network Plant Paging System Plant Radio System (dedicated offsite channels)

Plant Telephone Network ENS Network Plant Radio System MODE-NA Natural phenomena events affecting a loaded cask CO NFINEMENT BOUNDARY as indicated by VISIBLE DAMAGE to the cask:

• earthquake

• tornado (and tornado missile)

• fiood

• lightning

• snow/ ice None None None ISFSI Cask EU1.2 Cask Events Confine. Accident conditions affecting a loaded cask CONFINEMENT Confine. ISFSI Boundary BOUNDARY as indicated by VISI BLE DAMAGE to the cask: Boundary Events

• dropped cask

• tipped over cask

• cask burial

• explosion

• fire EU 1.3 Any condition in the opinion of the Emergency Director that indicates loss of loaded fuel storage cask CONFINEMENT BOUNDARY .

PINGP 1576, Rev. 10 Doc. Type/Sub Type: EP/EVT Retention: Lifetime + Page 5 of 8

Prairie Island Nuclear Generating Plant EMERGENCY ACTION LEVEL MATRIX HOT

• Barrier (Table F-1 ).

2 3 4 I Loss of ANY two Barriers AND Loss or Potential Loss of Third 2 3 4 Loss or Potential Loss of ANY two Barriers (Table F-1 ).

I Table F-1 FISSION PRODUCT BARRIER REFERENCE TABLE NOTE ANY Loss or ANY Potential Loss of EITHER Fuel Clad OR RCS (Table F-1).

2 3 4 ANY Loss or ANY Potential Loss of Containment (Table F-1 ).

Determine which combination of the three barriers are lost or have a potential loss and use the following key to classify the event. Also an event for multiple events could occur which result in the conclusion that exceeding the Loss or Potential Loss thresholds is imminent (i.e., within 1 to 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />). In this imminent loss situation use judgment and classify as if the thresholds are exceeded.

I Fuel Cladding Barrier RCS Barrier Containment Ba rrier 0 Loss D Potential Loss D Loss D Potential Loss D Loss D Potential Loss D 1. Critical Safety Function Status 1. Critical Safety Function Status 1. Critical Safety Function Status 1. Critical Safety Function Status 1. Critical Safety Function Status 1. Critical Safety Function Status Core-Cooling Red. Core Cooling-Orange; Not Applicable. RCS Integrity-Red; Not Applicable. Containment-Red.

OR OR Heat Sink-Red . Heat Sink-Red.

Fission Product Fission Product Barriers

2. Primary Coolant Activity Level D 2. Primary Coolant Activity Level 2. RCS Leak Rate 2. RCS Leak Rate 2. Containment Pressure 2. Containment Pressure Barriers Coolant Activity GREATER THAN Not Applicable. GREATER THAN available makeup Unisolable leak exceeding Rapid unexplained decrease following initial 46 PSIG and increasing ;

300 µCi/gm 1-131 equivalent. capacity as indicated by a loss of 60 gpm . increase; OR RCS subcooling LESS THAN 21 [40)" OR Containment hydrogen concentration GREATER THAN degree F. Containment pressure or sump level response OR EQUAL TO 6%;

not consistent with LOCA conditions . OR

• Adverse containment conditions are defined Containment pressure GR EATER THAN 23 psig with as a containment pressure greater than 5 psig LESS THAN one fu ll train of depressurization or conta inment rad iation level greater than eq uipment operating .

1E4 R/Hr.

3. Core Exit Thermocouple Readings D 3. Core Exit Thermocouple Readings 3. SG Tube Rupture 3. SG Tube Rupture 3. Core Exit Thermocouple Readings 3. Core Exit Thermocouple Readings GREATER THAN 1200 degree F. GREATER THAN 700 degree F. SGTR that results in an ECCS (SI ) Not Applicable. Not Applicable. Core exit thermocouples in excess of 1200 degrees F Actuation. and restoration procedures not effective within 15 minutes; OR Core exit thermocouples in excess of 700 degrees F with reactor vessel level below 40% RVLIS Full Range and restoration procedures not effective within 15 minutes.

4. Reactor Vessel Water Level D 4. Reactor Vessel Water Level 4. Containment Radiation Monitoring 4. Containment Radiation Monitoring 4. SG Secondary Side Release with P-to-S Leakage 4. SG Secondary Side Release with P-to-S Leakage Not Applicable. Level LESS THAN : Containment rad monitor 1(2)R-4 8 Not Applicable. RUPTURED S/G is also FAULTED outside of Not Applicable 40% RVLIS Full Range (no RCPs); or 49 reading GREATER THAN 7 R/hr. containment; 30% RVLIS Dynamic Head Range OR (1 RCP); Pri mary-to-Secondary leak rate GREATER THAN 60% RVLIS Dynamic Head Range 10 gpm with nonisolable steam release from affected (2 RCPs). SIG to the environment.

5. CNMT Isolation Valves Status After CNMT Isolation 5. CNMT Isolation Valves Status After CNMT Isolation

5. Containment Radiation 5. Containment Radiation Monitoring Containment isolation Valve(s) not closed; Not Applicable.

Monitoring Not Applicable. AND Containment rad monitor Direct pathway to the environment exists 1(2)R-48 or 49 reading after Containment Isolation signal.

GREATER THAN 200 R/hr.

6. Significant Radioactive Inventory in Containment 6. Significant Radioactive Inventory to Containment

6. Other Indications Not Applicable. Containment rad monitor 1(2)R-48 or 49 reading Not Applicable D 6. Other Indications 5. Other Indications 5. Other Indications GREATER THAN 800 R/hr.

Not Applicable. Not Applicable. Not Applicable.

7. Other Indications 7. Other Indications Not Applicable. Not Applicable.

7. Emergency Director Judgment D 7. Emergency Director Judgment 6. Emergency Director Judgment 6. Emergency Director Judgment Any condition in the opinion of Any condition in the opinion of the Any condition in the opinion of the Any condition in the opinion of the 8. Emergency Director Judgment 8. Emergency Director Judgment the Emergency Director that Emergency Director that indicates Emergency Director that indicates Emergency Director that indicates Any condition in the opinion of the Emergency Any condition in the opinion of the Emergency indicates Loss of the Fuel Potential Loss of the Fuel Loss of the RCS Barrier. Potential Loss of the RCS Director that indicates Loss of the Containment Director that indicates Potential Loss of the Clad Barrier. Clad Barrier. Barrier. Barrier. Containment Barrier.

PINGP 1576, Rev. 10 Doc. Type/Sub Type: EP/EVT Retention: Lifetime + Page 6 of 8

Prairie Island Nuclear Generating Plant EMERGENCY ACTION LEVEL MATRIX COLD

• AND I

Loss of All Offsite Power and Loss of All Onsite AC Power to Safeguards Buses.

I i 5 6 Failure of Diesel Generators 01 and 0 2 (05 and 0 6) to supply power to Safeguards Buses 15 and 16 (25 and 26) ;

i DEF Loss of power to or from Transformers CT-11 , CT-12 , 1RY, and 2RY i

that results in a loss of all offsite power to both Safeguards Buses 15 and 16 (25 and 26);

Loss of All Offsite Power to Safeguards Buses for GREATER THAN 15 Minutes.

s s Loss of power to or from Transformers CT-11 , CT-12 , 1RY, and 2RY that results in a loss of all offsite power to both Safeguards Buses 15 and 16 (25 and 26) for GREATER THAN 15 minutes; AND I

At least one Diesel Generator (0 1, 0 2, 0 5, 0 6) is supplying power to one of the affected safeguards buses .

Loss of None None Loss of AND Power Power Failure to restore power to Safeguards Bus 15 or 16 (25 or 26) within CU7 UNPLANNED Loss of Required DC Power for GREATER THAN 15 minutes from the lime of loss of both offsite and onsite AC power.

15 Minutes.

CU7.1 i 5 6 UNPLANNED Loss of required vital DC power based on LESS THAN 112 voe on 125 voe Panels 11 and 12 (21 and 22);

AND Failure to restore power to at least one required DC panel within 15 minutes from the time of loss .

CG1 Loss of RPV Inventory Affecting Fuel Clad Integrity with CS 1 Loss of RPV Inventory Affecting Core Decay Heat CA1 Loss of RCS Inventory. CU2 UNPLANNED Loss of RCS Inventory with Irradiated Fuel in Containment Challenged with Irradiated Fuel in the RPV. Removal Capability. the RPV.

CG 1.1 I i i I I 5 I 6 I CS 1.1 i i i 5 CA1.1 i 5 CU2.1 i i 6

1. Loss of RPV inventory as indicated by unexplained level increase With CONTAINMENT CLOSU RE not established: Loss of RCS inventory as indicated by RPV level at O inches UNPLANN ED RCS level decrease below the RPV fiange for GREATER in Containment Sumps A or C, or Waste Holdup Tank as indicated a. RPV inventory as indicated by RPV level LESS THAN Refueling Canal I RCS Narrow Range I Ultrasonic THAN OR EQ UAL TO 15 minutes .

by sump pump run times, levels, or alarms; 73% RVLI S Full Range; (at or LESS TH AN 75% RVLI S Full Range). CU2.2 i i 6 AND

2. RPV Level:

OR

b. RPV level cannot be monitored for GREATER THAN 30 minutes CA1 .2 ! i i i i 5 Loss of RPV inventory as indicated by unexplained level increase in Containment Sumps A or C, or Waste Holdup Tank as indicated Loss of RCS inventory as indicated by unexplained level

a. LESS THAN 63% RVUS Full Range for GREATER THAN 30 with a loss of RPV inventory as indicated by unexplained level increase in Containment Sumps A or C, or Waste Holdup by sump pump run times, levels, or alarms; minutes; increase in Containment Sumps A or C, or Waste Holdup Tank AND Tank as indicated by sump pump run times, levels, or alarms ;

OR as indicated by sump pump run times, levels, or alarms. AND RPV level cannot be monitored.

b. cannot be monitored , with indication or core uncovery for Cold SDI Refuel GREATER THAN 30 minutes as evidenced by one or more of the CS1 .2 I i 5 RCS level cannot be monitored for GREATER THAN 15 minutes . Cold SDI Refuel following : With CONTAINMENT CLOSUR E established: CA2 Loss of RPV Inventory with Irradiated Fuel in the RPV.

System System

• Containment Vessel Area Monitor R-2 reading a. RPV inventory as indicated by RPV level LESS THAN 63%

Malfunct. Malfunct.

GREATER THAN 1000 mR/hr RVLIS Full Range; OR CA2. 1 i 6

• Erratic Source Range Monitor Indication; Loss of RPV inventory as indicated by RPV level at O inches AND b. RPV level cannot be monitored for GREATER THAN 30 minutes with Refueling Canal I RCS Narrow Range I Ultrasonic. Reactor

3. Indication of CONTAINMENT challenged as indicated by one or more of the following :

a loss of RPV inventory as indicated by either:

• Unexplained level increase in Containment Sumps A or C, or CA2.2 i i i i i i 6 Vessel Level Loss of RCS inventory as indicated by unexplained level increase in

• Containment hydrogen concentration GREATER THAN OR EQUAL Waste Holdup Tank as indicated by sump pump run times, Containment Sumps A or C, or Waste Holdup Tank as indicated by T06% levels, or alarms Reactor sump pump run times, levels , or alarms;

• CONTAINMENT CLOSURE not established

• Erratic Source Range Monitor Indication Vessel AND

• Containment pressure GREATER THAN 1.0 psig with RPV level cannot be monitored for GREATER THAN 15 minutes .

Level CS2 Loss of RPV Inventory Affecting Core Decay Heat Removal CONTAINMENT CLOSURE established.

Capability with Irradiated Fuel in the RPV.

NOTE: CS2 .1 and CS2.2 should not be used for classification unless RPV level is below the bottom inside diameter (ID) of the RCS hot leg penetration. If level is at or above the Bottom ID , CU2 or CA2 should be used for event classification in the Refueling mode.

CS2.1 6 i With CONTAI NMENT CLOS UR E not established, and RPV level cannot be monitored, with indication of core uncovery as evidenced by one or more of the following :

• Containment Vessel Area Monitor R-2 reading GREATER THAN 1000 mR/hr

• Erratic Source Range Monitor Indication CS2.2 i i 6 With CONTAI NMENT CLOSU RE established, and RPV level cannot be monitored, with indication of core uncovery as evidenced by one or more of the following :

• Containment Vessel Area Monitor R-2 read ing GREATER THAN 1000 mR/hr

• Erratic Source Range Monitor Indication PINGP 1576, Rev. 10 Doc. Type/Sub Type: EPI EVT Retention: Lifetime + Page 7 of 8

' Prairie Island Nuclear Generating Plant EMERGENCY ACTION LEVEL MATRIX COLD

• I s W ith CONTAINMENT CLOSURE and RCS integrity not established s

an UNPLANNED event results in RCS temperature exceeding 200' F.

NOTES 1

1f an RCS heat removal system is in operation within this time frame and RCS temperature is being reduced then this EAL is not applicable.

UNPLANN ED Loss of Decay Heat Removal Capability with Irradiated Fuel in the RPV.

THAN 15 minutes.

s s

s An UNPLANNED event results in RCS temperature exceeding 200°F.

s I Loss of all RCS temperature and RPV level indication for GREATER I

RC S RCS 2 Temp. 1f the Pressurizer is solid then only the RCS temperature Temp.

None None threshold is applicable to CA4.3.

CA4.2  ! 5 6 W ith CONTAINMENT CLOSURE established and RCS integrity not established Q[ RCS inventory reduced an UNPLANNED event results in RCS temperature exceeding 200°F for GREATER THAN 1

20 minutes .

CA4.3 r-1---r----.,---.,--"T"- 5-.--6--r----,

An UNPLANNED event results in RCS temperature exceeding 200°F 1

for GREATER THAN 60 minutes or results in an RCS pressure f T CU6 UNPLANN ED Loss of All Onsite or Offsite Communications Capabilities.

CU6.1  ! l 5 6 Cold SD/ Comm . None None None Loss of all Table C-1 onsite communications capability affecting the Comm.

Refuel ability to perform routine operations.

System cus.2 j I I I s s Cold SD/

Malfunct. Loss of all Table C-2 offsite communications capability. Refuel System CU5 Fuel Clad Degradation. Malfunct.

CUS.1 j 5 6 l RCS Letdown Rad Monitor 1(2)R-9 or portable radiation monitoring Fuel Clad None None None instrumentation GREATER THAN 1.2 R/hr indicating fuel clad Fuel Clad Degradation degradation. Degradation CUS.2 '"!---,.---...----,--,....-5-..--6 -..---,

Coolant sample activity GREATER THAN Technical Specification 3.4.17 Condition C allowable limits indicating fuel clad degradation.

CU1 RCS Leakage.

RC S cu1.1 i I s RCS Leakage None None None Unidentified or pressure boundary leakage GREATER THAN 10 gpm . Leakage cu1.2 i I s Identified leakage GREATER THAN 25 gpm.

CUB Inadvertent Criticality.

Inadvertent Inadvertent Criticality None None None cua.11 s s I Criticality An UNPLANNED sustained positive startup rate observed on nuclear instrumentation.

Table C-1 Onsite Communications Systems Table C-2 Offsite Comm unications System Sound Powered Phones Plant Telephone Network Plant Paging System Plant Radio System (dedicated offsite channels )

Plant Telephone Network ENS Network Plant Radio System PINGP 1576, Rev. 10 Doc. Type/Sub Type: EP/EVT Retention: Lifetime + Page 8 of 8

Prairie Island Nuclear Generating Plant EMERGENCY ACTION LEVEL MATRIX

- GENERAL EMERGENCY SITE AREA EMERGENCY ALERT UNUSUAL EVENT HOT & COLD

• RG 1 RG1 .1 NOTE :

Offsite Dose Resulting from an Actual or Imminent Release of Gaseous Radioactivity Exceeds 1000 mRem TEDE or 5000 mRem Thyroid COE for the Actual or Projected Duration of the Release Using Actual Meteorology.

2 3 4 5  ! 6 i DEF i If dose assessment results are available at the time of declaration ,

the classification should be based on RG1 .2 instead of RG1 .1.

While necessary declarations shou ld not be delayed awaiting results ,

the dose assessment should be initiated I completed in order to RS1 RS1 .1 NOTE:

Offsrte Dose Resulting from an Actual or Imminent Release of Gaseous Radioactivity Exceeds 100 mRem TEDE or 500 mRem Thyroid COE for the Actual or Projected Duration of the Release.

2 3 4 5 6  ! DEF !

If dose assessment results are available at the time of declaration ,

the classification should be based on RS1 .2 instead of RS1 .1.

While necessary declarations should not be delayed awaiting results ,

the dose assessment should be initiated / completed in order to RA1 RA 1.1 Any UNPLANNED Release of Gaseous or Liquid Radioactivity to the Environment that Exceeds 200 Times the Offsrte Dose Calculation Manual Specification for 15 Minutes or Longer.

2 3  ! 4 5 setpoint established by a current radioactivity discharge permit for 15 minutes or longer.

OR 6  ! DEF VALI D reading on any effluent monitor that exceeds 200 Times the alarm RU1 RU1 .1 Any UNPLANNED Release of Gaseous or Liquid Rad1oact1vity to the Environment that Exceeds Two Times the Offsite Dose Calculation Manual Specification for 60 Minutes or Longer.

2 3 4 5 6  ! DEF VALID reading on any effiuent monitor that exceeds two times the alarm setpoint established by a current radioactivity discharge permit for 60 minutes or longer.

determine if the classification should be subsequently escalated . determine if the classification should be subsequently escalated .

VALID reading on effluent monitor R-18 that exceeds 900,000 cpm for 15 minutes or longer.

RU1 .2 i 2 3 4 5 6  ! DEF VALID reading on one or more monitors listed in Table R-1 that exceeds VALID reading on one or more monitors listed in Table R-1 that exceeds VALID reading on one or more of the following radiation monitors or expected to exceed column "GE" for 15 minutes or longer: or is expected to exceed column "SAE" for 15 minutes or longer: (Table R-1 ) that exceeds the reading shown for 60 m inutes or longer:

Offsite Rad Offsite Rad Conditions Conditions RG1 2 i 2 3 4 5 6 i DEF  ! RS1 .2 i 2 3 4 5 6  ! DEF RA1.2 2 3 4 5 6  ! DEF RU 1.3 l 2 3 4 5 6  ! DEF  !

Dose assessment using actual meteorology indicates doses GREAT ER Dose assessment using actual meteorology indicates doses GR EATER VALID reading on one or more of the following radi ation monitors Confirmed sample analysis for gaseous or liquid release indicates THAN 1000 mRem TEDE or 5000 mRem thyroid COE at or beyond the THAN 100 mRem TEDE or 500 mRem thyroid COE at or beyond the (Table R-1 ) that exceeds the reading shown for 15 minutes or longer: concentrations or release rates , with a release duration of 60 minutes site boundary. site boundary. or longer, in excess of two times ODCM specification.

RA1.3 I 2 3 4 5 6  ! DEF !

RG1~ i 2 3 4 5 6  ! DEF RS1 .3 i 2 3 4 5 6  ! DEF Confirmed sample analysis for gaseous or liquid release indicates Field survey results indicate closed window dose rates exceeding Field survey results indicate closed window dose rates exceeding concentrations or release rates , with a release duration of 15 minutes 1000 mR/hr expected to continue for more than one hour, at or beyond 100 mR/hr expected to continue for more than one hour, at or beyond or longer, in excess of 200 Times ODCM specification.

site boundary; the site boundary; OR OR Analyses of field survey samples indicate thyroid COE of 5000 mRem for Analyses of field survey samples indicate thyroid COE of 500 mRem for one hour of inhalation, at or beyond site boundary. one hour of inhalation, at or beyond the site boundary.

RA2 Damage to Irradiated Fuel or Loss of Water Level that Has or RU2 Unexpected Increase in Plant Radiation.

Abnormal Will Result in the Uncovering of Irradiated Fuel Outside the Abnormal Rad Reactor Vessel. Rad Release RA2.1 2 3  ! 4 5 6  ! DEF ! RU2.1 2 3 4 5  ! 6 DEF Release A VALi D alarm on one or more of the following radiation monitors: VALID indication of uncontrolled water level decrease in the reactor Rad Rad

• R-25 or R-31 SFP Air Monitor (HI Alarm) refueling cavity, spent fuel pool , or fuel transfer canal with all irradiated Effluent Effluent

• R-5 Fuel Handling Area Monitor reading (HI Alarm) fuel assemblies remaining covered by water as indicated by level

• R-28 New Fuel Pool Criticality Area Monitor (HI Alarm ) LESS THAN SFP low water level alarm , Refueling Canal Level , or

• 1(2) R-11 Ctmt/SBV Air Particulate Monitor (HI Alarm ) visual observation (752.5 feet elevation);

• 1(2) R-12 Ctmt/SBV Radio Gas Monitor (HI Alarm ) AND

• 1(2) R-2 Containment Vessel Area Monitor (HI Alarm ) Any UNPLANNED VALID Area Radiation Monitor reading increases as indicated by:

• R-5 Fuel Handling Area Monitor reading RA2 .2 i 2 3 4 5 6  ! DEF

• R-28 New Fuel Pool Criticality Area Monitor Water level LESS THAN 1O feet above an irradiated fuel assembly

• 1(2) R-2 Containment Vessel Area Monitor for the reactor refueling cavity, spent fuel pool and fuel transfer canal

• Other Portable Area Radiation Monitoring Instrumentation Onsite Rad that will result in irradiated fuel uncovering Onsite Rad Conditions RU2.2 i 2 3 4 5 6 i DEF  ! Conditions RA3 Release of Radioactive Material or Increases in Radiation Levels Any UNPLANNED VALID Area Radiation Monitor read ing increases by Within the Facility That Impedes Operation of Systems Required a factor of 1000 over normal' levels.

to Maintain Safe Operations or to Establish or Maintain Cold Shutdown. 'Normal levels can be considered as the highest reading in the past RA3.1 2 3 4 5 6  ! DEF ! twenty-four hours excluding the current peak value VALID radiation monitor readings GREATER THAN 15 mR/hr in areas requiring continuous occupancy to maintain plant safety functions :

Control Room (Rad monitor R-1 );

OR Central Alarm Station (by portable rad iation monitoring instrumentation).

RA3.2 i 2 3 4 5 6  ! DEF Any VALID radiation monitor rea ding GREATER THAN 1 R/hr in Table R-1 Effluent Monitor Classification Thresholds areas requiring infrequent access to maintain plant safety functions Monitor GE SAE Alert UE (Table H-1).

Gaseous CPM CPM 1(2) R-50 High Range Stack Gas Monitor 43000 mR/hr 4300 mR/hr N/A N/A 1R-22' Shield Building Vent Rad Monitor N/A N/A 160,000'/ 1.6 E5 1,600'/ 1.6 E3 Table H-1 Plant Areas 2R-22' Shield Building Vent Rad Monitor N/A N/A 100,000'/ 1 E5 1,000'/ 1 E3 1R-30' & 1R-3r Unit 1 Aux. Building Vent Rad Monitors N/A N/A 100,000'/ 1 E5 1,000'/ 1 E3 Area HU1 .6' HU2.1' HA1.2 HA13 HA1.4 HA1 .5 HA2 .1 HA3.1' HA3.2' RA3 .2 2R-30' Unit 2 Aux. Building Vent Rad Monitors N/A N/A 100,000'/ 1 E5 1,000'/ 1 E3 2R-37" Unit 2 Aux. Building Vent Rad Monitors NIA N/A 120,000'/ 1.2 E5 1,200'/ 1.2 E3 - Shield/Containment Building X X X X X X X X X R-35' Radwaste Building Vent Rad Monitor N/A NIA 100,000'/ 1 E5 1,000'/ 1 E3 - Auxiliary Building X X X X X X X X X X R-25' & R-31' Spent Fuel Pool Vent Rad Monitors N/A N/A 800,000'/ 8 E5 8,000'/ 8 E3 - 05/ 06 Diesel Generator Building X X X X X X X X X X Liquid - Plant Screenhouse X X X X X X X X X X R-18' Waste Effiuent Liquid Monitor NIA N/A 900,000'/ 9 E5 30,000'/ 3 E4 - Control Room X X X X X X X X X 1R-19' SG Slowdown Radiation Monitor N/A N/A 100,000'/ 1 E5 1,000'/ 1 E3 - Relay Room X X X X X X X X X X 2R-19' SG Slowdown Radiation Monitor N/A N/A 60,000'/ 6 E4 600'/ 6 E2 - Turbine Building X X X X X X X X X X R-21 Circ Water Dischar e Monitor N/A N/A 800,000/ 8 E5 8,000/ 8 E3 - Condensate Storage Tanks X X X X Notes: 1) ERCS EAL Alarms indicate an EAL threshold May have been exceeded. Further eva luation of the radiation monitor reading is required to determ ine if the EAL th reshold is exceeded. 2)" Applies when Effluent discharge not isola ted.

• Also consider areas contiguous to these.

PINGP 1576, Rev. 10 Doc. Type/Sub Type: EP/EVT Retention: Lifetime + Page 1 of 8

Prairie Island Nuclear Generating Plant EMERGENCY ACTION LEVEL MATRIX

• HA 1.2 Natural and Destructive Phenomena Affecting the Plant VITAL AREA 1

1 2

2  !

3 3  ! 4 4

5 Seismic Event GREATE R THAN Operating Basis Earthquake (QBE) 5  !

Tornado or high winds GREATE R THAN 95 mph within PROTECT ED 6

6 AR EA boundary and resulting in VISIBLE DAMAG E to any of the j DEF as indicated by "OBE Exceedance" alarm on Seismic Monitoring Panel.

DEF ! HU1 ~

l Natural and Destructive Phenomena Affecting the PROTECTED AREA l

2 2

! 3 3

4 4

5 5

6 Earthquake felt in plant as indicated by VALID "Event" alarm on Seismic Monitoring Panel.

l 6 Report by plant personnel of tornado or high winds GREATER THAN 95 mph striking within PROTECTED AREA boundary.

DEF j DEF following plant structures / equipment or Control Room indication of degraded performance of those systems (Table H-1).

HU1 .3 l 1  ! 2  ! 3  ! 4  ! 5  ! 6 j DEF  !

Vehicle crash into plant structures or systems within PROTECTED HA 1.3  ! 2 3 4 5 6 j DEF AREA boundary.

Natural & Vehicle crash within PROTECTED AREA boundary and resulting in HU 1.4  ! 1  ! 2 3 4 5 6 j DEF Natural &

Destructive VISIBLE DAMAGE to any of the following plant structures / equipment Report by plant personnel of an unanticipated EXPLOSION within Destructive Phenomenon therein or Control Room indication of degraded performance of those PROTECT ED AR EA boundary resulting in VISIBLE DAMAG E to Phenomenon None None systems (Table H-1 ). permanent structure or equipment.

HA1 .4  ! 1 2 3 4 5 6 j DEF  ! 2 3 4 5 6 l DEF  !

Turbine failure-generated missiles result in any VISIBLE DAMAG E Report of turbine fai lure resulting in casing penetration or damage to to or penetration of any of the following plant areas (Table H-1 ). turbine or generator seals .

HA1.5  ! 1  ! 2  ! 3  ! 4  ! 5  ! 6  ! DEF ! HU1.6 l  ! 2  ! 3 4 5 6  ! DEF  !

Uncontrolled flooding in any Table H-1 area of the plant that results in Uncontrolled flooding in following areas of the plant that has the degraded safety system performance as indicated in the Control Room potential to affect safety related equipment needed for the current Hazards or that creates industrial safety hazards (e.g ., electric shock) that operating mode (Table H- 1) Hazards precludes access necessary to operate or monitor safety equipment. HU1.7  ! 1  ! 2 3 4 5 6 j DEF j 2 3 4 5  ! 6 j DEF j High or low river water level occurrences affecting the PROTECTED High or low river water level occurrences affecting the PROTECTED AREA as indicated by:

AREA as indicated by: River intake level GREATER THAN 692 ft MSL; River intake level GREATE R THAN 698 ft MS L; OR OR River intake level LESS THAN 669.5 ft MSL.

River intake level LESS THAN 666.5 ft MSL.

HA2 FIRE or EXPLOS IO N Affecting the Operabi lity of Plant Safety HU2 FIRE Within PROTECTED AREA Boundary Not Extinguished Systems Required to Establish or Maintain Safe Shutdown. Within 15 Minutes of Detection.

Fire or HA2. 1  ! 1  ! 2  ! 3  ! 4  ! 5  ! 6  ! DEF  ! HU2.1 2 3 4 5 6  ! DEF  ! Fire or Explosion None None FIRE or EXPLOSION in any of the following areas (Table H-1): FIRE in buildings or areas contiguous (in actual contact with or Explosion AND immediately adjacent) to any Table H-1 area not extinguished within Affected system parameter indications show degraded performance 15 minutes of control room notification or verification of a control or plant personnel report VISIBLE DAMAG E to permanent structures room alarm .

or equipment within the specified area HA3 Release of Toxic or Flammable Gases Within or Contiguous HU3 Release of Toxic or Flammable Gases Deemed Detrimental to to a VI TAL AREA Which Jeopardizes Operation of Systems Normal Operation of the Plant.

Required to Maintain Safe Operations or Establish or Maintain Safe Shutdown.

Toxic and None None HA3.1 2 3 4 5 6  ! DEF  ! HU3.1 2 3 4 5 6 DEF  ! Toxic and Flammable Report or detection of toxic gases within or contiguous to Table H-1 Report or detection of toxic or flammable gases that has or could enter Flammable Gas areas in concentrations that may result in an atmosphere IMMEDIATELY the site area boundary in amounts that can affect NORMAL PLANT Gas DANG EROUS TO LIFE AND HEALTH (IDLH) OPERATIONS.

HA3.2 l  ! 2  ! 3  ! 4  ! 5 6 j DEF j HU3.2  ! 2 3 4 5 6 j DEF j Report or detection of gases in concentration GR EATER THAN the Report by Local , County or State Officials for evacuation or sheltering LOWER FLAMMABILITY LIMI T within or contiguous to Table H-1 areas. of site personnel based on an offsite event.

Table H-1 Plant Areas Area HU1 _6* HU2.1* HA1 .2 HA1.3 HA14 HA1 .5 HA2 .1 HA3_1* HA3_2* RA3.

- Shield/Containment Building X X X X X X X X X

- Auxiliary Building X X X X X X X X X X

- 05/06 Diesel Generator Building X X X X X X X X X X

- Plant Screenhouse X X X X X X X X X X

- Control Room X X X X X X X X X

- Relay Room X X X X X X X X X X

- Turbine Building X X X X X X X X X X

- Condensate Storage Tanks X X X X

• Also consider areas conti uous to these PINGP 1576, Rev . 10 Doc. Type/Sub Type: EP/EVT Retention: Lifetime + Page 2 of 8

Prairie Island Nuclear Generating Plant EMERGENCY ACTION LEVEL MATRIX HOSTILE ACTION Resulting in Loss of Physical Control of the Facility.

1  ! 2 3 4 5 6 A HOSTILE ACTION has occurred such that plant personnel are i DEF unable to operate equipment required to maintain safety functions .

HG1 .2  ! 1  ! 2  ! 3  ! 4  ! 5  ! 6  ! DEF A HOSTILE ACTION has caused failure of Spent Fuel Cooling Systems and IMM INENT fuel damage is likely for a freshly off-loaded reactor core in pool.

2 3 4 5 A HOSTILE ACTION is occurring or has occurred within the PROTECTED AREA as reported by Shift Security Supervision.

6  ! DEF or Airborne Attack Threat.

2 f 3 4 5 A HOSTILE ACTION is occurring or has occurred within the OWNER CONTROLLED AREA as reported by Security Shift Supervision.

HA4.2 r!---..--2-..--3--r--4--r--5-"T"-6--,ir-D-E-F--,

6 i DEF A validated notification from NRG of an airliner attack threat within 30 minutes of the site .

a Potential Degradation in the Level of Safety of the Plant.

1 f 2 f 3  ! 4 f 5 A SECURITY CONDITION that does NOT involve a HOSTILE ACTION as re[°rted bJ Securi} Shift Supervision.

HU4.2  ! 1  ! 2 _ 3 _ 4 _

A credible PI NGP security threat notification.

5  !

6  !

6 f DEF DEF f DEF 2 3 4 5 6 Security A validated notification from NRG providing information of an aircraft threat.

Hazards Continued HS2 Control Room Evacuation Has Been Initiated and Plant Control HA5 Control Room Evacuation Has Been Initiated .

Cannot Be Established .

Control Room HS2 .1 f 2 3 4 5 6 DEF HA5.1 2 3 4 5 6 DEF f Evacuation None Control room evacuation has been initiated ; Entry into 1(2)C1 .3 AOP-1 Shutdown from Outside the Control Room None AND or F-5 Appendix B Control Room Evacuation (Fire) for control room Control of the plant cannot be established per 1(2)C 1.3 AOP-1 , evacuation.

Shutdown from Outside the Control Room or F-5 Appendix B, Control Room Evacuation (Fire) within 15 minutes .

HG2 Other Conditions Existing Which in the Judgment of the HS3 Other Conditions Existing Which in the Judgment of the HA6 Other Conditions Existing Which in the Judgment of the HU5 Other Conditions Existing Which in the Judgment of the Emergency Director Warrant Declaration of General Emergency. Emergency Director Warrant Declaration of Site Area Emergency Director Warrant Declaration of an Alert. Emergency Director Warrant Declaration of a UE.

Emergency .

Emergency HG2.1 2 3 4 5 6 f DEF f HS3.1 2 3 4 5 6 i DEF  ! HA6 .1 2 3 4 5 6 i DEF f HU5.1 i 2 3 4 5 6 DEF  ! Emerg ency Director Other conditions exist which in the judgment of the Emergency Director Other conditions exist which in the judgment of the Emergency Director Other conditions exist which in the judgment of the Emergency Director Other conditions exist which in the judgment of the Emergency Director Director Judgment indicate that events are in process or have occurred which involve actual indicate that events are in process or have occurred which involve actual indicate that events are in process or have occurred which involve actual indicate that events are in process or have occurred which indicate a Judgment or imminent substantial core degradation or melting with potential for or likely major failures of plant functions needed for protection of the public. or likely potential substantial degradation of the level of safety of the plant. potential degradation of the level of safety of the plant. No releases of loss of containment integrity. Releases can be reasonably expected to Any releases are not expected to result in exposure levels which exceed Any releases are expected to be limited to small fractions of the EPA radioactive material requiring offsite response or monitoring are expected exceed EPA Protective Action Guideline exposure levels offsite for more EPA Protective Action Guideline exposure levels beyond the site boundary. Protective Action Guideline exposure levels. unless further degradation of safety systems occurs .

than the immediate site area .

Tabl e H-1 Plant Areas Area HU1 .6* HU2 1" HA1.2 HAU HA14 HA1 .5 HA2.1 HA3 .1* HA3.2* RA3.

- Shield/Containment Building X X X X X X X X X

- Auxiliary Building X X X X X X X X X X

- D5/D6 Diesel Generator Building X X X X X X X X X X

- Plant Screenhouse X X X X X X X X X X

- Control Room X X X X X X X X X

- Relay Room X X X X X X X X X X

- Turbine Building X X X X X X X X X X

- Condensate Storage Tanks X X X X

• Also consider areas conti uous to these PINGP 1576, Rev. 10 Doc. Type/Sub Type: EP/EVT Retention: Lifetime + Page 3 of 8

Prairie Island Nuclear Generating Plant EMERGENCY ACTION LEVEL MATRIX HOT

• Prolonged Loss of All Offsite Power and Prolonged Loss of All Onsite AC Power to Safeguards Buses.

2 I 3 4 Loss of power to or from Transformers CT-11 , CT-12 , 1RY, and 2RY that results in a loss of all offsite power to both Safeguards Buses 15 and 16 (25 and 26);

AND Failure of Diesel Generators D1 and D2 (D5 and D6) to supply power to Safeguards Buses.

2 3 4 I Loss of power to or from Transformers CT-11 , CT-12, 1RY, and 2RY that results in a loss of all offsite power to both Safeguards Buses 15 and 16 (25 and 26) ;

AND Failure of both Diesel Generators D1 and D2 (D5 and D6) to supply SA5.1  !

AC power capability to Safeguards Buses reduced to a single power source for GREATER THAN 15 minutes such that any additional single failure would result in station blackout.

2 3 4 AC power capability to Safeguards Buses 15 and 16 (25 and 26) reduced to only one of the following sources for GREATER THAN 15 minutes:

• Transformer CT-11 ;

• Transformer CT-12; Loss of All Offsite Power to Safeguards Buses for GREATER THAN 15 Minutes.

I 2 3 4 Loss of power to or from Transformers CT-11 , CT-12, 1RY, and 2RY that results in a loss of all offsite power to both Safeguards Buses 15 and 16 (25 and 26) for GREATER THAN 15 minutes ;

AND Two Diesel Generators (D1 , D2, D5, D6) are supplying power to Loss of to Safeguards Buses 15 and 16 (25 and 26); power to Safeguards Buses 15 and 16 (25 and 26);

• Transformer 1RY; Safeguards Buses 15 and 16 (25 and 26). Loss of Power AND AND

• Transformer 2RY; Power Either of the following : Failure to restore power to Safeguards Bus 15 or 16 (25 or 26) within

• Diesel Generator D1 (D5);

a. Restoration of Safeguards Bus 15 or 16 (25 or 26) within 15 minutes from the time of loss of both offsite and onsite AC power.

• Diesel Generator D2 (D6);

4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> is not likely; AND OR SS3 Loss of All Vital DC Power. Any additional single failure will result in station blackout.

b. Continuing degradation of core cooling based on Fission Product Barrier monitoring as indicated by Core Cooling-RED SS3.1  ! 2 3 4 or ORANGE path. Loss of all Safeguards DC power based on LESS THAN 112 VDC on 125VDC Panels 11 and 12 (21 and 22) for GREATER THAN 15 minutes .

SG2 Failure of the Reactor Protection System to Complete an SS2 Failure of Reactor Protection System Instrumentation to SA2 Failure of Reactor Protection System Instrumentation Automatic Trip and Manual Trip was NOT Successful and Complete or Initiate an Automatic Reactor Trip Once a to Complete or Initiate an Automatic Reactor Trip Once a There is Indication of an Extreme Challenge to the Ability Reactor Protection System Setpoint Has Been Exceeded Reactor Protection System Setpoint Has Been Exceeded to Cool the Core. and Manual Trip Was NOT Successful. and Manual Trip Was Successful.

SG2.1 I 2  ! ss2 .1 I 2 I SA2.1 I 2 3  !

lndication(s) exist that automatic and manual trip were NOT lndication(s) exist that automatic and manual trip were NOT NOTE: A failed manual trip followed by a successful manual trip successful in reducing power to LESS THAN 5%; successful in reducing power to LESS THAN 5%. reducing reactor power to less than 5% meets this EAL.

AND RPS Either of the following :

lndication(s) exist that a Reactor Protection System setpoint was RPS Failure a. Core cooling is extremely challenged as indicated by exceeded; Failu re Core Cooling - RED path; AND None OR System RPS automatic trip did not reduce power to LESS THAN 5%; System

b. Heat removal is extremely challenged as indicated by Malfunct. AND Malfunct.

Heat Sink - RED path.

Any of the following operator actions are successful in reducing power to LESS THAN 5% , Manual Control Board :

• Reactor Trip

• AMSAC/DSS Actuation

• Turbine Trip SS4 Complete Loss of Heat Removal Capability. SU2 Inability to Reach Required Shutdown Within Technical Specification Limits .

Inability to Reach or SS4.1 I  ! 2 3 4 SU2.1  ! 1  ! 2  ! 3 4 Inability to Reach or None Loss of core cooling and heat sink as indicated by: None Plant is not brought to required operating mode within Technical Maintain Maintain

a. Core Cooling - RED path; Specifications LCO Action Statement Time.

Shutdown Shutdown AND Conditions Conditions

b. Heat Sink - RED path.

SS6 Inability to Monitor a SIGNIFICANT TRANSIENT in Progress . SA4 UNPLANNED Loss of Most or All Safety System Annunciation SU3 UNPLANNED Loss of Most or All Safety System Annunciation or Indication in Control Room With Either (1) a SIGNIFICANT or Indication in the Control Room for Greater Than 15 minutes .

T RANSIENT in Progress, or (2) Compensatory Non-Alarming Indicators are Unavailable.

SS6.1 i 2 3 4  ! SA4 1 i 2 3 4  ! SU3 1 I 2 3 4  !

Loss of most (approximately >75%) or all annunciators associated UNPLANNED loss of most (approximately >75%) or all annunciators UNPLANNED loss of most (approximately >75%) or all annunciators or with safety systems: or indicators associated with safety systems for GREATER THAN indicators associated with safety systems for GREATER THAN None

• Main Control Boards A, 8-1 (2), C-1(2), D-1 (2), E-1(2), F-1 (2), 15 minutes : 15 minutes:

Inst. / G-1(2) NIS Racks I, II , Ill , IV, and ERGS Alarms;

• Main Control Boards A , 8-1 (2), C-1 (2), D-1 (2), E-1 (2), F-1 (2),

• Main Control Board A, 8 -1(2), C-1(2), D-1 (2), E- 1(2), F-1(2) , G-1 (2) Inst./

Comm . AND G-1(2) NIS Racks I, II , Ill , IV, and ERGS Alarms; NIS Racks I, II, Ill , IV, and ERGS Alarms . Comm.

A SIGNIFICANT TRANSIENT in progress ; AND SU6 UNPLANNED Loss of All Onsite or Offsite Communications AND Either of the following :

Capabilities

a. A SIGNIFICANT TRANSIENT in progress; Compensatory non-alarming indications are unavailable; AND OR SU6.1  ! 1  ! 2 3 4 Indications needed to monitor the ability to shut down the reactor, b. Compensatory non-alarming indications are unavailable. Loss of all Table C-1 onsite communications capability affecting the maintain the core cooled, maintain the reactor coolant system ability to perform routine operations.

intact, and maintain containment intact are unavailable. SU6.2  ! 1  ! 2  ! 3  ! 4 Loss of all Table C-2 offsite communications ca abilit .

Table C-1 Onsite Communications Systems Table C-2 Offsite Communications System Sound Powered Phones Plant Telephone Network Plant Pag ing System Plant Radio System (dedicated offsite channels)

Plant Telephone Network ENS Network Plant Radio System PINGP 1576, Rev. 10 Doc. Type/Sub Type: EP/EVT Retention: Lifetime + Page 4 of 8

Prairie Island Nuclear Generating Plant EMERGENCY ACTION LEVEL MATRIX HOT

• Fuel Clad Degradation None None None SU4.1  !

fuel clad degradation.

SU4.2  ! 1  !

2 2

3 3

4 4

Radialion Monitor 1(2)R-9 GREATER THAN 1.2 R/hr indicaling Coolant sample activity GREAT ER THAN Technical Specification 3.4.17 Condition C allowable limits indicating fuel clad degradation.

Fuel Clad Degradation SU5 RCS Leakage.

System Malfunct. 2 3 4 I System Malfunct.

RCS None None None Unidentified or pressure boundary leakage GREATER THAN 10 gpm. RCS Leakage Leakage sus.2 I 2 3 4 I Identified leakage GREATER THAN 25 gpm .

SUB Inadvertent Criticality.

Inadvertent Criticality None None None 3 4 I Inadvertent Criticality An UNPLANN ED sustained positive startup rate observed on nuclear instrumentation.

Table C-1 Onsite Communications Systems Table C-2 Offsite Communications System Sound Powered Phones Plant Telephone Network Plant Paging System Plant Radio System (dedicated offsite channels )

Plant Telephone Network ENS Network Plant Radio System MODE-NA Natural phenomena events affecting a loaded cask CONFINEM ENT BOUNDARY as indicated by VISIBLE DAMAGE to the cask:

• earthquake

• tornado (and tornado missile)

• flood

• lightning

• snow / ice None None None ISFSI Cask EU1.2 Cask Events Confine. Accident conditions affecting a loaded cask CONFINEMENT Confine. ISFSI Boundary BOUNDARY as indicated by VISIBLE DAMAGE to the cask : Boundary Events

• dropped cask

• tipped over cask

• cask burial

• explosion

• fire EU1. 3 Any condition in the opinion of the Emergency Director that indicates loss of loaded fuel storage cask CONFINEMENT BOUNDARY .

PINGP 1576, Rev . 10 Doc. Type/Sub Type: EP/EVT Retention: Lifetime + Page 5 of 8

Prairie Island Nuclear Generating Plant EMERGENCY ACTION LEVEL MATRIX HOT

• 2 I 3 4 Loss of ANY two Barriers AND Loss or Potential Loss of Third Barrier (Table F-1 ).

2 3 4 I Loss or Potential Loss of ANY two Barriers (Table F-1 ).

Table F-1 FISSION PRODUCT BARRIER REFERENCE TABLE NOTE ANY Loss or ANY Potential Loss of EITHER Fuel Clad OR RCS (Table F-1).

2 3 4 1 ANY Loss or ANY Potential Loss of Containment (Table F-1).

Determine which combination of the three barriers are lost or have a potential loss and use the following key to classify the event. Also an event for multiple events could occur which result in the conclusion that exceeding the Loss or Potential Loss thresholds is imminent (i.e., within 1 to 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />) . In this imminent loss situation use judgment and classify as if the thresholds are exceeded .

Fuel Cladd ing Barrie r RCS Barrier Containment Ba rrier D Loss D Potential Loss D Loss D Potential Loss D Loss D Potential Loss D 1. Critical Safety Function Status 1. Critical Safety Function Status 1. Critical Safety Function Status 1. Critical Safety Function Status 1. Critical Safety Function Status 1. Critical Safety Function Status Core-Cooling Red Core Cooling-Orange; Not Applicable. RCS Integri ty-Red; Not Applicable. Containment-Red .

OR OR Heat Sink-Red. Heat Sink-Red .

Fission Product Fission Product Barriers

2. Primary Coolant Activity Level D 2. Primary Coolant Activity Level 2. RCS Leak Rate 2. RCS Leak Rate 2. Containment Pressure 2. Containment Pressure Barriers Coolant Activity GREATER TH AN Not Applicable. GREATER THAN available makeup Unisolable leak exceeding Rapid unexplained decrease following initial 46 PSIG and increasing ;

300 µCi/gm 1-131 equivalent. capacity as indicated by a loss of 60gpm increase; OR RCS subcooling LESS THAN 21 [40]' OR Containment hydrogen concentration GREATER THAN degree F. Containment pressure or sump level response OR EQUAL TO 6%;

not consistent with LOCA conditions. OR

• Adverse containment con ditions are defined Containment pressure GREATER THAN 23 psig with as a containment pressure greater than 5 psig LESS THAN one full train of depressurization or containment radiation level greater than 1E4 R/Hr. equipment operating .

3. Core Exit Thermocouple Readings D 3. Core Exit Thermocouple Readings 3. SG Tube Rupture 3. SG Tube Rupture 3. Core Exit Thermocouple Readings 3. Core Exit Thermocouple Readings GREATER THAN 1200 degree F. GREATER THAN 700 degree F. SGTR that results in an ECCS (S I) Not Applicable. Not Applicable. Core exit thermocouples in excess of 1200 degrees F Actuation. and restoration procedures not effective within 15 minutes ;

OR Core exit thermocouples in excess of 700 degrees F with reactor vessel level below 40% RVLIS Full Range and restoration procedures not effective within 15 minutes.

4. Reactor Vessel Water Level D 4. Reactor Vessel Water Level 4. Containment Radiation Monitoring 4. Containment Radiation Monitoring 4. SG Secondary Side Release with P-to-S Leakage D 4. SG Secondary Side Release with P-to-S Leakage Not Applicable. Level LESS THAN : Containment rad monitor 1(2)R-48 Not Applicable. RUPTURED S/G is also FAULTED outside of Not Applicable 40% RVLIS Full Range (no RCPs); or 49 reading GREATER THAN 7 R/hr. containment; 30% RVLIS Dynamic Head Range OR (1 RCP); Primary-to-Secondary leak rate GREATER THAN 60% RVLIS Dynamic Head Range 10 gpm with nonisolable steam release from affected (2 RCPs ). SIG to the environment.

5. CNMT Isolation Valves Status After CNMT Isolation 5. CNMT Isolation Valves Status After CNMT Isolation

5. Containment Radiation 5. Containment Radiation Monitoring Containment isolation Valve(s) not closed ; Not Applicable.

Monitoring Not Applicable. AND Containment rad monitor Direct pathway to the environment exists 1(2)R-48 or 49 reading after Containment Isolation signal.

GREATER THAN 200 R/hr.

6. Sig nificant Radioactive Inventory in Containment 6. Significant Radioactive Inventory to Containment

6. Other Indications Not Applicable . Containment rad monitor 1(2)R-48 or 49 reading Not Applicable D 6. Other Indications 5 Other Indications 5. Other Indications GREATER THAN 800 R/hr.

Not Applicable. Not Applicable. Not Applicable.

7. Other Indications 7. Other Indications Not Applicable. Not Applicable.

7. Emergency Director Judgment D 7. Emergency Director Judgment 6. Emergency Director Judgment 6. Emergency Director Judgment Any condition in the opinion of Any condition in the opinion of the Any condition in the opinion of the Any condition in the opinion of the 8. Emergency Director Judgment 8 Emergency Director Judgment the Emergency Director that Emergency Director that indicates Emergency Director that indicates Emergency Director that indicates Any condition in the opinion of the Emergency Any condition in the opinion of the Emergency indicates Loss of the Fuel Potential Loss of the Fuel Loss of the RCS Barrier. Potential Loss of the RCS Director that indicates Loss of the Containment Director that indicates Potential Loss of the Clad Barrier. Clad Barrier. Barrier. Barrier . Containment Barrier.

PINGP 1576, Rev. 10 Doc. Type/Sub Type: EP/EVT Retention: Lifetime + Page 6 of 8

Prairie Island Nuclear Generating Plant EMERGENCY ACTION LEVEL MATRIX COLD

• AND to Safeguards Buses.

!  ! 5 6 j DEF j Loss of power to or from Transformers CT-11 , CT-12, 1RY, and 2RY that results in a loss of all offsite power to both Safeguards Buses 15 and 16 (25 and 26);

Failure of Diesel Generators 01 and 02 (05 and 06) to supply power to Safeguards Buses 15 and 16 (25 and 26);

Loss of All Offsite Power to Safeguards Buses for GREATER THAN 15 Minutes.

s s I Loss of power to or from Transformers CT-1 1, CT-12 , 1RY, and 2RY that results in a loss of all offsite power to both Safeguards Buses 15 and 16 (25 and 26) for GREATER THAN 15 minutes; AND At least one Diesel Generator (01 , 02 , 05, 06) is supplying power to one of the affected safeguards buses.

Loss of None None Loss of AND Power Power Failure to restore power to Safeguards Bus 15 or 16 (25 or 26) within CU7 UNPLANNED Loss of Required DC Power for GREATER THAN 15 minutes from the time of loss of both offsite and onsite AC power.

15 Minutes.

CU7.1 5  ! 6 UNPLANNED Loss of required vital DC power based on LESS THAN 112 voe on 125 voe Panels 11 and 12 (21 and 22);

AND Failure to restore power to at least one required DC panel within 15 minutes from the time of loss.

CG1 Loss of RPV Inventory Affecting Fuel Clad Integrity with CS1 Loss of RPV Inventory Affecting Core Decay Heat CA1 Loss of RCS Inventory. CU2 UNPLANNED Loss of RCS Inventory with Irradiated Fuel in Containment Challenged with Irradiated Fuel in the RPV. Removal Capability. the RPV.

CG 1.1  !  !  !  !  ! 5  ! 6  ! CS 1.1  !  !  ! 5 CA1.1  ! 5 CU2.1  !  ! 6

1. Loss of RPV inventory as indicated by unexplained level increase With CONTAINMENT CLOSUR E not established : Loss of RCS inventory as indicated by RPV level at O inches UNPLANNED RCS level decrease below the RPV flange for GREATER in Containment Sumps A or C, or Waste Holdup Tank as indicated a. RPV inventory as indicated by RPV level LESS THAN Refueling Canal I RCS Narrow Range I Ultrasonic THAN OR EQUAL TO 15 minutes.

by sump pump run times, levels, or alarms ; 73% RVLIS Full Range; (at or LESS THAN 75% RV LIS Full Range ). CU2.2  ! 6 AND

2. RPV Level:

OR

b. RPV level cannot be monitored for GREATER THAN 30 minutes CA1.2  ! I I I  ! 5 Loss of RPV inventory as indicated by unexplained level increase in Containment Sumps A or C , or Waste Holdup Tank as indicated Loss of RCS inventory as indicated by unexplained level

a. LESS THAN 63% RVLIS Full Range for GREATER THAN 30 with a loss of RPV inventory as indicated by unexplained level increase in Containment Sumps A or C, or Waste Holdup by sump pump run times, levels, or alarms ;

minutes; increase in Containment Sumps A or C, or Waste Holdup Tank Tank as indicated by sump pump run times, levels, or alarms; AND OR as indicated by sump pump run times , levels, or alarms. AND RPV level cannot be monitored .

b. cannot be monitored , with indication or core uncovery for Cold SD/

Refuel GREATER THAN 30 minutes as evidenced by one or more of the CS1 .2 j  ! 5 RCS level cannot be monitored for GR EATER THAN 15 minutes. Cold SD/

Refuel following : With CONTAINMENT CLOSURE established : CA2 Loss of RPV Inventory with Irradiated Fuel in the RPV.

System System

• Containment Vessel Area Monitor R-2 reading a. RPV inventory as indicated by RPV level LESS THAN 63%

Malfunct. Malfunct.

GREATER THAN 1000 mR/hr RVLIS Full Range; OR CA2.1 6  !

• Erratic Source Range Monitor Indication; Loss of RPV inventory as indicated by RPV level at O inches AND b. RPV level cannot be monitored for GREATER THAN 30 minutes with Refueling Canal I RCS Narrow Range I Ultrasonic. Reactor

3. Indication of CONTAINM ENT challenged as indicated by one or more of the following :

a loss of RPV inventory as indicated by either:

• Unexplained level increase in Containment Sumps A or C, or CA2.2  !  !  !  !  !  ! 6 Vessel Level Loss of RCS inventory as indicated by unexplained level increase in

• Containment hydrogen concentration GREATER THAN OR EQUAL Waste Holdup Tank as indicated by sump pump run times ,

Containment Sumps A or C, or Waste Holdup Tank as indicated by T06% levels , or alarms Reactor sump pump run times , levels , or alarms;

• CONTAINMENT CLOSURE not established

• Erratic Source Range Monitor Indication Vessel AND

• Containment pressure GREATER THAN 1.0 psig with RPV level cannot be monitored for GREATER THAN 15 minutes.

Level CS2 Loss of RPV Inventory Affecting Core Decay Heat Removal CONTAINMENT CLOSURE established.

Capability with Irradiated Fuel in the RPV.

NOTE: CS2 .1 and CS2.2 should not be used for classification unless RPV level is below the bottom inside diameter (ID) of the RCS hot leg penetration . If level is at or above the Bottom ID , CU2 or CA2 should be used for event classification in the Refueling mode.

CS2.1 6  !

With CONTAINMENT CLOSU RE not established , and RPV level cannot be monitored , with indication of core uncovery as evidenced by one or more of the following :

• Containment Vessel Area Monitor R-2 reading GREATER THAN 1000 mR/hr

• Erratic Source Range Monitor Indication CS2.2 j  ! 6 With CONTAINMENT CLOSUR E established, and RPV level cannot be monitored, with indication of core uncovery as evidenced by one or more of the following :

• Containment Vessel Area Monitor R-2 reading GREATER THAN 1000 mR/hr

• Erratic Source Range Monitor Indication PINGP 1576, Rev. 10 Doc. Type/Sub Type: EP/EVT Retention: Lifetime + Page 7 of 8

Prairie Island Nuclear Generating Plant EMERGENCY ACTION LEVEL MATRIX COLD

• 5 With CONTAINMENT CLOSURE and RCS integrity not established 1

an UNPLANNED event results in RCS temperature exceeding 200°F.

NOTES 1 6

1f an RCS heat removal system is in operation within this time frame and RCS temperature is being reduced then this EAL is not applicable.

UNPLANNED Loss of Decay Heat Removal Capability with Irradiated Fuel in the RPV.

I 5

5 6

An UNPLANNED event results in RCS temperature exceeding 200°F.

6 Loss of all RCS temperature and RPV level indication for GREATER THAN 15 minutes.

I RCS RCS 2

Temp. None 1f the Pressurizer is solid then only the RCS temperature Temp.

None threshold is applicable to CA4 .3.

CA4.2 5 6  !

With CONTAINMENT CLOSURE established and RCS integrity not established Q!: RCS inventory reduced an UNPLANNED event results in RCS temperature exceeding 200°F for GREATER T HAN 1

20 minutes .

CA43 ~ , - - ~ - ~ - - ~ - - - , . - - - ~ ~ - ~ ,

An UNPLANNED event results in RCS temperature exceeding 200°F for GREATER THAN 60 minutes 1 or results in an RCS pressure 2

f R AT R THAN i .

CU6 UNPLANNED Loss of All Onsite or Offsite Communications Capabilities.

CU6.1  !  ! 5 6 Cold SD/ Comm . None None None Loss of all Table C- 1 onsite communications capability affecting the Comm .

Refuel ability to perform routine operations.

System Malfunct.

cu6.2  ! I I I 5 6 Cold SD/

Loss of all Table C-2 offsite communications capability. Refuel System CU5 Fuel Clad Degradation. Malfunct.

CU5.1 I 5 6  !

RCS Letdown Rad Monitor 1(2)R-9 or portable radiation monitoring Fuel Clad None None None instrumentation GREATER THAN 1.2 R/hr indicating fuel clad Fuel Clad Degradation degradation. Degradation CU5.2 .-,- - . - - -.....------,-,-5--,.---..,.6-.----,

Coolant sample activity GREATER THAN Technical Specification 3.4 .17 Condition C allowable limits indicating fuel clad degradation.

CU1 RCS Leakage.

RCS cu1 .1 I 5 I RCS Leakage None None None Unidentified or pressure boundary leakage GR EATER THAN 10 gpm . Leakage CU 1 2 I I 5 Identified leakage GREATER THAN 25 gpm.

CUB Inadvertent Criticality.

Inadvertent Inadvertent Criticality None None None cua.11 5 6 I Criticality An UNPLANNED sustained positive startup rate observed on nuclear instrumentation.

Table C-1 Onsite Communications Systems Table C-2 Offsite Communications System Sound Powered Phones Plant Telephone Network Plant Paging System Plant Radio System (dedicated offsite channels )

Plant Telephone Network EN S Network Plant Radio System PINGP 1576. Rev. 10 Doc. Type/Sub Type: EP/EVT Retention: Lifetime + Page 8 of 8

ENCLOSURE 2 10 CFR 50.54(q) Procedure Change Summary Analysis I

1 page follows

ENCLOSURE 2 10 CFR 50.54(q) Procedure Change Summary Analysis Change(#) 1

==

Description:==

The change will be made in both the Emergency Plan. EAL Matrix (PINGP 1576) Table F-1 FISSION PRODUCT BARRIER TABLE under the RCS Barrier, loss Column, #2 and in FS-2.1 (Emergency Action Level Technical Basis) Revision 12 page 6-F-7

1. Critical Safety Function Status:

"Less than or equal to 20[30] degrees F" will be changed to "less than 21 [40] degree F" The change will be made in both the Emergency Plan, EAL Matrix (PINGP 1576) Table f-1 FISSION PRODUCT BARRIER TABLE under the Fuel Clad Barrier, Potential Loss Column, #4 and in F3-2.1 (emergency Action Level Technical Basis) Revision 12 page 6-F-7

1. Critical Safety Function Status:

"32% with 1 RCP running" to 30% with 1 RCP running" and "62% with 2 RCPs running" to "60% with 2 RCPs running" ,

The changes are to align the Emergency Plan EAL Matric (PINGP 1576) thresholds with changes made to the Critical Safety Function Status Tree (CSFST) set points identified in the EOPs and ERCS per EC 27440. The CSFST set points are the basis for the noted EAL threshold values. The changes do not change the meaning or intent of the EAL and only align them with the new set point value. The screening determined that the revision meets the definition of change per Regulatory Guide 1.219 and that further evaluation is required.

Doc IDs or (Procedure Numbers)/ Revision Numbers: Prairie Island Nuclear Generating Plant Form 1576 - Emergency Action Level (EAL) Matrix, Revision 9, and F3-2.1 Emergency Action Level Technical Bases, Revision 12 Document

Title:

PINGP 1576 and F3-2.1 PCR Number: 602000001164 and 602000001144 Editorial Basis (applies to E-Plan changes only) NONE Licensing/Basis Affected NEI 99-01 Revision 4 scheme of Emergency Action Level Actions was implemented BY Prairie Island Nuclear Generating Plant (PINGP) in accordance with the USN RC Safety Evaluation Report (SER), dated November 18, 2005. Changes to the PINGP EALS are required by the evaluation against the EALs approved for use at PINGP per that SER.

Evaluation Determination: The EALs continue to comply with the approved SER and NEI 99-01, basis guidance. Per NEI 99-01, revision 4, the basis for the affected set points is those from the CSFST monitoring and functional restoration procedures. For Prairie Island Nuclear Generating Plant (PINGP),

these procedures are the EOPs and associated set points established by the EOPs. The meaning and intent of the basis for EALs remains unchanged with the parameter change. The effectiveness of the PINGP E-Plan is maintained by updating the thresholds to align with those approved in calculations SPC-EP-114 and SPC-EP-121 by use of the engineering change process.