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Prairie Island, Units 1 and 2 - Enclosure 1, Form 1576 Emergency Action Level (EAL) Matrix, Revision 10 and Emergency Action Level Technical Bases, Revision 13
	ML18113A054
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Site:	Prairie Island
Issue date:	04/17/2018
From:	; Northern States Power Co, Xcel Energy
To:	; Office of Nuclear Reactor Regulation
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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;
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	L-PI-18-016
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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 Facility Address Transmittal Date Vital Ack Req

US NRC C/0 PAM JOHNSON (P.I.) Pl PAMELA JOHNSON DOCUMENT CONTROL DESK US NRC 12/18/2017 Facility Doc Type Sub Type Document Number Pl PRO EP F3-2.1 Marked (*) documents require your acknowledgement.

From C-DOC CNTRL-PI Address 1717 WAKONADE DR WELCH, MN 55089 PARTIAL UPDATE Vital NO Transmittal Group ID 1018 Status Revision Status Date ISSUED 013 12/18/2017 Acknolwedgement Date: ________________

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PRAIRIE ISLAND NUCLEAR GENERATING PLANT EMERGENCY PLAN IMPLEMENTING PROCEDURE . EMERGENCY ACTION LEVEL TECHNICAL BASES ;;::~f*ff*ifl~*t:*****, NUMBER: F3-2.1

REV: 13

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:

12/1/17 PCR #: 602000001144 Page i PRAIRIE ISLAND NUCLEAR GENERATING PLANT EMERGENCY PLAN IMPLEMENTING PROCEDURE Title Page EMERGENCY ACTION LEVEL TECHNICAL BASES Record of Revision Record of Revision Emergency Action Level Technical Bases Document (22 pages) Table R-0 Category R -Abnorma_l Rad Levels/Radiological Effluent (19 pages) Table C-0 Category C -Cold Shutdown/Refueling System Malfunction (29 pages) Table E-0 Category E -Independent Spent Fuel Storage Installations (ISFSI) (4 pages) Table F-0 Category F -Fission Product Barrier Degradation (17 pages) Table H-0 Category H -Hazards (28 pages) Table S-0 Category S -System Malfunction (30 pages)

Page ii NUMBER: F3-2.1

REV: 13 Date of Revision Revision Number 2017 13 2017 13 2015 11 2014 10 2017 12 6 2017 13 I_

2014 10 2017 12 *

*

PRAIRIE ISLAND NUCLEAR GENERATING PLANT EMERGENCY PLAN IMPLEMENTING PROCEDURE Section Table F-0 Table 1 Change NUMBER: EMERGENCY ACTION LEVEL TECHNICAL BASES Significant Changes From the Previous Revision F3-2.1 REV:* 13 Incorporate changed values for subcooling and RVLIS for Table F-1 basis information for Fission Product Barriers . Page iii *

UE FU1 ANY Loss or ANY Potential Loss FA1 of Containment Op. Modes: Power Operation, Hot Standby, Startup, Hot Shutdown

Table F-0 Recognition Category F F.ission Product Barrier Degradation INITIATING CONDITION MATRIX ALERT ANY Loss or ANY Potential Loss FS1 of EITHER Fuel Clad OR RCS Op. Modes: Power Operation, Hot Standby, Startup, Hot Shutdown NOTES ' SITE AREA EMERGENCY Loss or Potential Lo~s of ANY Two Barriers Op. Modes: Power Operation, Hot Standby, Startup, Hot Shutdown 1. The logic used for these initiating conditions reflects the following considerations:

FG1

GENERAL EMERGENCY Loss of ANY Two Barriers AND Loss or Potential Loss of Third Barrier Op. Modes: Power Operation, Hot Standby, Startup, Hot Shutdown

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 I Cs 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 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months ) 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 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months ). 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 Fuel Clad Barrier EALS LOSS POTENTIAL LOSS 1. Criticai Safety Function Status Core-Cooling Red OR Core Cooling-Orange OR Heat Sink-Red 2. Primary Coolant Activity Level Coolant Activity GREATER THAN 300 µCi/gm 1-131 equivalent PINGP Not Applicable LOSS RCS Barrier EALS POTENTIAL LOSS . 1. Critical Safety Function Status Not Applicable OR 2. RCS Leak Rate GREATER THAN available makeup capacity as indicated by a loss of RCS subcooling LESS THAN 21 [40]* degree F "Adverse containment conditions are defined as a . containment pressure greater than 5 psig or containment radiation level greater than 1 E4 R/Hr. During adverse

containment conditions use iCCM to determine RCS subcoollng.

6-F-3 RCS Integrity-Red OR Heat Sink-Red Unisolable leak exceeding 60gpm Loss or Potential Loss of Third Barrier Containment Barrier EALS LOSS POTENTIAL LOSS 1. Critical Safety Function Status Not Applicable OR 2. Containment Pressure Rapid unexplained decrease following initial increase OR Containment pressure or sump level response not consistent with LOCA conditions Containment-Red 46 PSIG and increasing OR Containment hydrogen concentration GREATER THAN OR EQUAL TO 6% . OR Containment pressure GREATER THAN 23 psig with LESS THAN one full train of depressurization equipment operating F3-2. 1, Rev. 13 I I I*

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 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months ). 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 Fuel Clad Barrier EALS LOSS POTENTIAL LOSS OR 3. Core Exit Thermocouple Readings GREATER THAN 1200 degree F PINGP

GREATER THAN 700 degree F LOSS Barriers RCS Barrier EALS POTENTIAL LOSS 6-F-4

Loss or Potential Loss of Third Barrier Containment Barrier EALS LOSS

POTENTIAL LOSS 3. Core Exit Thermocouple Readings Not applicable Core exit thermocouples in 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 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 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months ). In this imminent loss situation use judgment and classify as if the thresholds are exceeded.

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

OR 4. Reactor Vessel Water Level Not Applicable OR Level LESS THAN:

40% RVLIS Full Range (no RCPs)

30% RVLIS Dynamic Head Range (1 RCP)

60% RVLIS Dynamic Head Range (2 RCPs) 5. Containment Radiation Monitoring Containment rad monitor 1 (2) R-48 or 49 reading GREATER THAN 200 R/hr PINGP Not Applicable OR 3. SG Tube Rupture SGTR that results in an ECCS (SI) Actuat!on OR Not Applicable

4. Containment Radiation Monitoring Containment rad monitor 1 (2) R-48 or 49 reading GREATER THAN 7 R/hr 6-F-5 Not Applicable LOSS POTENTIAL LOSS OR 4. SG Secondary Side Release with P-to-S Leakage RUPTURED S/G is also FAUL TED outside of containment OR . Primary-to-Secondary leak rate GREATER THAN 10 gpm with nonisolable steam release from affected S/G to the environment OR Not applicable

5. CNMT Isolation Valves Status After CNMT Isolation Containment Isolation Valve(s) not closed AND Direct pathway to the -environment exists after Containment Isolation signal OR Not Applicable

6. Significant Radioactive Inventory in Containment Not Applicable Containment rad monitor reading.1 (2) R-48 or 49 GREATER THAN 800 R/hr' 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 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months ). 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 Fuel Clad Barrier EALS LOSS OR 6. Other Indications Not Applicable OR POTENTIAL LOSS Not Applicable

7. Emergency Director Judgment Any condition*

in the opinion of the Emergency Director that indicates Loss or Potential Loss of the Fuel Clad Barrier PINGP

LOSS OR RCS Barrier EALS POTENTIAL LOSS 5. Other) Indications

Not Applicable OR Not Applicable

6. Emergency Director Judgment Any condition in the opinion of the Emergency Director that indicates Loss or Potential Loss of the RCS Barrier 6-F-6 *

Loss or Potential Loss of Third Barrier Containment Barrier EALS LOSS POTENTIAL LOSS OR 7. Other Indications Not Applicable OR Not Applicable

8. Emergency Director Judgment Any condition in the opinion of the Emergency Director that indicates Loss or Potential Loss of the Containment Barrier 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 1 E+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 A T_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 (A TWS) 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 (SI). ECCS (SI) actuation is caused by: 1

PRZR pressure less than 1830 psig

Either SG pressure less than 530 psig

Containment pressure greater than 3.5 psig PINGP 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 (A TWS) 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

*

L ---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, 1 O] 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 from

I the RWST upon an ESF actuation signal. [Ref. 11, 12] . PINGP 6-F-13 F3-2.1, Rev. 13

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 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 1 O

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 I Cs. [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 Facility Address \ Transmittal Date Vital Ack Req

US NRC C/0 PAM JOHNSON (P.I.) Pl PAMELA JOHNSON DOCUMENT CONTROL DESK US NRC 12/18/2017 Facility Doc Type Sul?_ Type Document Number Pl FRM PINGP 1576 £-PVttJ 8o.01< P} ~"? i~~.ls)******

Marked (*) documents require your acknowledgement.

From C-DOC CNTRL-PI Address 1717 WAKONADE DR WELCH, MN 55089 Vital NO Transmittal Group ID 1020 Status Revision ISSUED 010 ISSUED 010 <. Status Date 12/18/2017 1211a,2017 Acknolwedgement Date: / Signature:

. /'i If documents no longer required'for this copyholder, complete QF2122 Request for Service, and submit to Document Control. Copy Holder 515 515 Media Copies 1

*

Prairie Island Nuclear Generating Plant EMERGENCY AC T ION LEVEL MATRIX -GENERAL EMERGENCY SITE AREA EMERGENCY ALERT UNUSUAL EVENT Abnormal Rad Release Rad Effluent Offsite Rad Cond itions Onsite Rad Conditions PIN GP 1576 , Rev. 10 Doc Type/S ub Type: EP/EVT Retent io n: L i fetime + RG 1 Offsne Dose Res ult i ng from an Actua l or Imm i nent Release of Gaseous Rad i oact ivi ty Exceeds 1000 mR em TEDE or 5000 m Rem Thyro i d COE for the A c tual o r Projected Durat i on of the Release Using Actual Meteorology. RG1.1 1 ! 2 3 4 5 6 ! DEF ! NOTE: If dose assessment results are available at the t i me of declaration , the classification should be based on RG1 .2 i nstead of RG1 .1. While necessary declarations should not be delayed aw a iting resu l ts , the dose assessmen t should be i nitiated/ completed in order to determ ine if the classification should be subsequently escalated. VALID reading on one or more monitors listed in Tab le R-1 that e xc eeds or expected to exceed column " GE" for 15 minutes or longer: RG1 2 I 2 3 4 5 ! 6 I DEF Dose assessment us i ng actua l meteorology indicates doses GREATER THAN 1000 mRem TE D E or 5000 mRem thyroid CDE at or beyond the site b ou ndary. RG1.3 ! 2 3 4 5 6 DEF Fie l d survey results indicate c l osed window dose ra t es exceeding 1000 mR/hr expected to continue for more than one hour. at or beyond site boundary; OR Analyses of field survey samples indicate thyroid COE of 5000 mRem for one hour of inhalation. at or be yon d site boundary. RS1 Offsite Dose Resulting fr o m an Actua l or Imminent Release of Gaseous Rad i oact i vfy Exceeds 1 00 mRem TEDE or 500 mRem Thyro i d COE for the Actual or Projected Duration of the Release. RS1.1 2 3 4 5 6 ! DEF ! NOTE: If dose assessment results are ava il ab le at the t ime of declaration , the c l assification should be based on RS1 .2 instead of RS1 .1. While necessary dec l ara t ions s h ould not be delayed awaiting results , the dose assessment should be initiated/ completed in order to determine i f the classification should be subsequently escalated. VALID reading on one or more monitors listed in Table R-1 that exceeds or is expected to exceed column " SA E" for 15 minutes or longer: RS1 .2 I 2 3 4 5 6 ! DEF ! Dose assessment using actua l meteorology in d ic ates doses GREATER T H A N 100 mRem TED E or 500 mRem thyroid COE at or beyond the s i te boundary. RS1 .3 I 2 3 4 5 6 ! DEF F ie l d survey r esults indic a te closed window dose r a tes exceedi n g 100 mR/hr expected to continue for more than one hour. at or beyond the s ite boundary; OR Analyses of field survey samples indicate thy r oid COE of 500 mRem for one hour of inhalation. at or be yo nd the site boundary. Table R-1 Effluent Monitor Classification Thresholds Monitor Gaseous 1 (2) R-50 H i gh Range Stack Gas Monit or 1 R-22* Shield Build i ng Vent Rad Mon i tor 2R-22* Shield Bui l d i ng Vent Rad Mon itor 1 R-30* & 1 R-37" Un it 1 Aux. Bu i ld i ng Vent Rad Monit ors 2R-30* Unit 2 Aux. Building Ven t R ad Monitors 2R-37" Uni t 2 Aux. Bu il d ing Vent Ra d Monitors R-35* Radwaste Bu i lding Vent Rad Monitor R-25* & R-31* S ent F uel Pool Vent Rad Mon itors Liquid R-18* Waste Effluent L iquid Mon i tor 1 R-19* SG Slowd own Radiat ion Mon itor 2R-19* SG Slowdown Rad i at ion Mon itor R-21 Circ Water D i schar e Mon i tor GE SAE 4 3000 mR/hr 4300 mR/h r N/A N/A N/A N/A N/A N/A N/A N/A N/A N/A N/A N/A N/A N/A N/A N/A N/A N/A N/A N/A N/A N/A Alert CPM N/A 1so.000*1 1.s Es 100 , 000*t 1 ES 100 , 000*t 1 ES 100 , 000*t 1 E 5 120 , 000*1 1.2 E 5 100 , 000*t 1 ES 800 , 000*1 8 ES 900 , 000*/ g ES 100 , 000*t 1 ES 60 , 000*t 6 E4 800 , 000/ 8 E S UE CPM N/A 1 , 600*/ 1.6 E 3 1.000*11 E3 1.000*11 E3 1.000*1 1 E3 1 , 200*1 1.2 E3 1 , 000*11 E 3 8 , 000*15 E 3 30.000*1 3 E4 1.000*11 E 3 600*/6 E2 8 , 000/ 8 E3 N o t es: 1) ERCS EAL Alarms indicate a n EA L t h r es h o ld M ay have be en e x ceede d. Furth e r ev a lu at ion of the rad iation m o nitor r e a d in g is r e qu i r ed to d e t e rm ine i f t he EA L th r es h old is excee d ed. 2)

Appl ies w hen Efflu e nt d isch a rge n ot iso l a t ed. RA1 Any UNPLANNED Release of Gaseous or Liquid Radioactivfy to the Env i ronment that Exceeds 200 T i mes the Offs i te Dose Calculation Manua l S pec i fication fo r 15 M i nutes or Longer. RA 1.1 2 ! 3 4 5 6 I DEF I VALID reading on any effluent mon i tor th a t exceeds 200 Times the a lar m setpoint establis h ed by a current radioact i vity discharge pe rm i t for 1 5 mi nu tes or lo n g e r. OR VALID reading on effluent mon i to r R-18 tha t exceeds 900 , 000 cpm for 15 m inutes or longe r. RA1.2 2 3 4 5 6 ! DEF VALID reading on one o r more of the following radiation mon ito rs (Ta ble R-1) that exceeds th e r eading s hown for 15 minutes or longer: RA 1.3 I 2 3 4 5 6 ! DEF ! Confirmed sample anal ysi s for gaseous or l iqui d release ind icat es concentrations or release rat e s , with a re le ase dura t io n of 1 5 minutes or longer , i n excess of 200 T imes ODCM specificat i on. RA2 Da m age to I rrad i ated Fue l or Loss of Wate r Leve l that Has o r Will Resu l t in the Uncover i ng of Irradiated Fue l Outs i de the Reactor Vesse l. RA2.1 2 3 4 5 6 ! DEF ! A VALID a l a rm on one or more of the following radiation monitors:

R-25 or R-31 SFP Air Mon i tor (HI Alarm)

R-5 Fuel H andling Area Monitor reading (HI Alarm)

R-28 New Fue l P ool Criticality Area Monitor (HI Alarm)

1 (2) R-1 1 CtmVSBV Air P art ic ulate Mon itor (HI Alarm)

1 (2) R-12 Ctm V SBV Radio Gas Monitor (HI Alarm)

1 (2) R-2 Containment V e ssel A r ea M onitor (H I A l arm) RA2.2 I 2 3 4 5 6 ! DEF Water l e ve l LESS THAN 10 feet abo ve an irr ad i ated fuel assembly fo r the reactor refuel i ng c avity , spent fuel pool and fuel t r ansfer canal that will result i n i rrad i ated fuel unc ov er i ng RA3 Release of Rad i oactive Materi a l or Increases i n Rad i at i on Leve l s With i n the Facility That Impedes O peration of Systems Requ i red to Ma i ntain Safe Operat i ons or to Establ i sh or Ma i nta i n Cold Shutdown. RA3.1 2 3 4 5 6 ! DEF ! VA LI D radiation mo n itor re a dings GR E A TER T HAN 15 mR/hr in areas requ i ring cont i nuous occupancy to maintain p l ant safet y functions: Control Room (Rad mon i tor R-1); OR Central Alarm Statio n (by portable radia t ion moni t oring ins t rumentation). RA3.2 I 2 3 4 5 6 ! DEF Any VALID rad i ation mon i tor reading GR E AT E R THAN 1 R/hr i n areas requiring infrequent acc e ss to ma i ntain plant safety functions (T a ble H-1). Area -Shie l d/Containment B uild i ng -Auxiliary Bu il d i ng -D5/D6 D i esel Generator Bu il ding -Plant Screenh o use -Control Ro om -Re l ay Room -Turb i ne Bu il d ing -Condensate Storage Tanks RU1 Any UNPLJ'.NNED Release of Gaseous or Liqu i d Radioact i vfy to the Env ir->nment that Exceeds Two T i mes the Offsne Dose Calcu l at i on Manua l Specificat i on for 60 Minutes or Longer. RU1.1 2 3 4 5 6 ! DEF ! VALID reading on any effluent mon it or that exceeds two times the alarm se t point established by a current r a dioactivity discharge permi t for 60 m inutes o r longer. RU1 .2 I 2 3 4 5 6 ! DEF VALID readi ng on one or m ore of the following radiation mon i tors (T able R-1) that exceeds the reading shown for 60 m i nutes or longer: RU1 .3 I 2 3 4 5 6 ! DEF ! Confirmed sample analysis for gaseous or liquid release i nd ic ates concentrations or re l ease rates , with a re l e a se duration of 60 minutes or longer , in excess of two t im es ODCM specification. RU 2 Une x pe c t ed I ncreas e i n Plant Rad i ation. RU2.1 2 3 4 5 6 DEF ! VALID i nd icatio n of uncontrolled water level d e crease in the reactor refuel in g cavity , spe nt fue l pool. or fuel transfer canal with all i rrad iated fuel assemblies rema i ning covered by water as indicated by level LESS THAN SFP low water le v el alarm , Refue l ing Canal Le vel , or vi s u al observation (752.5 feet ele v ation); AND Any UNP LA N NED V ALID Area Rad i a tion Mo n itor r eading increases as indic a ted by:

R-5 F uel Hand lin g Area Mon i tor reading

R-28 New Fue l Pool C rit ic al i t y Area Mon itor

1 (2) R-2 Containment Vessel Area Monitor

Other Portable A rea Radiat i on Mon itori ng Instrumentat ion RU2.2 I 2 3 4 5 6 ! DEF Any UN P LANNED V ALID Area Rad i at io n Mon i tor read i ng i ncreases b y a factor of 1000 over normal* levels. *N orm al levels c a n be considered a s the highest reading i n the past twenty-four hours excluding the curre nt pe a k v a l ue. ----Table H-1 Plant Areas HU16* HU 2.1* H A1.2 HA1.3 HA1.4 HA1.5 HA2.1 HA3_1* H A3.2* X X X X X X X X X X X X X X X X X X X X X X X X X X X X X X X X X X X X X X X X X X X X X X X X X X X X X X X X X X X X X X X X X X X

Also co ns i der areas contiauous to these. HOT & COLD Abnormal Rad Release Rad Effluent RA3.2 X X X X X Offslte Rad Conditions Onsite Rad Conditions Page 1 of 8

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Prairie Island Nuclear Generating Plant Fire o r E x plo sion Toxic and Flammabl e Gas PIN G P 1576 , Rev. 1 0 Doc. Type/Sub Type: EP/EVT Reten tion: Lifetime + None None N one None None None EMERGENCY ACTION LEVEL MATRIX Natural and Des t ructive Phenomena Affec t ing the Plant VITAL A R EA. ! ! 2 3 4 5 6 ! DEF ! Seismic Event G R EATER TH AN Opera t ing Basis E arthquake (O B E) as indicated by "O BE Exceedance" alarm on Seismic Mon ito ring Pane l. HA 1.2 ! 1 ! 2 ! 3 ! 4 ! 5 ! 6 ! DEF ! Tornado or high winds GR E A T ER THAN 95 mph within PROT E C TE D A RE A bou n dary a n d r esulting in VISIB LE DAMAG E to a n y of the following plant structures

/ equipment or Control Room i ndication of degraded performance of those systems (Table H-1 ). HA1.3 ! 2 3 4 5 6 I DEF Vehicle crash within PROTECTED AREA boundary and resulting in VISIBLE DAMAGE to any of the following plant structures

/ equipment therein or Control Room indicat i on of degraded performance of those systems (Table H-1). HA 1.4 ! 1 2 3 4 5 6 i DEF ! T urbine failure-generated missiles result in any V ISI B LE DAMAG E to or penetration of any o f the foll owing p lan t areas (Tab le H-1 ). HA 1.5 ! 1 ! 2 ! 3 ! 4 ! 5 ! 6 ! DEF ! Uncontrolled flooding in any Table H-1 ar ea of the plant that results in degraded safety system performan c e as indicated in the Control Room or that creates industr i al safet y hazards (e.g .. electric sho ck) that precludes acces s necessary to operate or monitor safety equipment.

2 3 4 5 6 ! DEF ! H i gh or low ri v er w ater level occurrences affe cting the PROTEC T ED AREA as indic ated by: R iver intake level GREAT E R THAN 698 ft MSL; OR R iv er intake leve l LESS THAN 666.5 ft MSL. HA2 FIR E or EXPLOSION Affecting the Operability of P lan t Safety Systems Required to Establish or Ma intain Safe Shutdown. HA2.1 ! 1 ! 2 ! 3 ! 4 I 5 ! 6 ! DEF ! FIRE or EXPLOSION i n an y of the following a reas (Ta ble H-1): AND Affected system parameter indications show degraded performance or plant personnel report V I S I BLE D AMAG E to permanent structures or equipment within the specified area. HA3 Re leas e of Toxic or F lammable Gases Within or Contiguous to a VITAL AREA Which Jeopardizes Operation of Systems Required to Maintain Safe Operations or Establish or Maintain Safe Shutdown. HA3.1 2 3 4 5 6 ! DEF ! Report or detection of toxic g a ses within or contiguous to Table H-1 areas in concentrations that may result in an atmosphere I MM E DIA T ELY DANGEROUS TO LIFE AND HEAL TH (IDLH). HA3.2 ! ! 2 ! 3 ! 4 ! 5 6 i DEF i Report or detection of gases in concentration GREA TE R T HAN the LOWER FLAMMABILITY LIMIT within or contiguous to T able H-1 areas. Tab l e H-1 Plant Areas Area HU1.6* HU2.1* HA 12 HA1.3 HA1.4 HA1.5 H A2.1 HA3.1' HA3.2* R A3. -Shi eld/Conta i nment Build i ng X X X X X X X X X -Auxiliary Building X X X X X X X X X X -D5/D 6 Diesel Generator Build i ng X X X X X X X X X X -Plant Screenhouse X X X X X X X X X X -Control Ro om X X X X X X X X X -Relay Ro om X X X X X X X X X X -Turb ine Bu i lding X X X X X X X X X X -Condensate Storage Tanks X X X X

Also cons i der areas conti uous to these . Natural and Destruct iv e Phenomena Affecting the PROTECTED AREA. ! I 2 ! 3 4 5 6 DEF Earthquake felt in plant as indicated by VALID "Event" alarm o n Se i sm ic Mon i toring Panel. HU1 .2 ! 1 ! 2 ! 3 4 5 6 i DEF i Report by plant personnel of t ornado or high winds GR E A T ER THAN 95 mph striking wi t hin PRO TE C TE D AR E A boundary. HU1 .3 ! 1 ! 2 ! 3 ! 4 ! 5 ! 6 i DEF ! Vehicl e crash in t o p l a nt structures or syste ms w ith in PROTECTED AREA boundary. HU1 .4 ! 1 ! 2 3 4 5 6 i DEF Report by plant per so nne l of an unanticipated EXPLOSION with i n PROT E CTED ARE A boundary resu l t i ng in VISIBLE DAMAGE to permanent structure or equipment.

HU1 .5 ! 2 3 4 5 6 i DEF Report of turbine fa il ure resulting in casi ng penetrat ion or damage to turbi ne or generator seals. HU1 .6 ! 1 ! 2 ! 3 4 5 6 i DEF Uncontrolled flooding in fo l lowing areas of the plant that has the potential to affect safet y related equ i pment needed for th e curren t operat i ng mode (T a ble H-1 ). HU1.7 ! 1 2 3 4 5 6 i DEF i High or l ow r iv er wa t er level occurrences affecting the PROTEC T ED AREA as indicated by: R iver intake leve l GREATER TH A N 692 ft MSL; OR Riv e r intake leve l L ESS THAN 669.5 ft MSL. HU2 FIRE Within PROTECTED AREA Boundary Not Ext inguished Within 15 Minutes of Detection. HU2.1 2 3 4 5 6 ! DEF ! FIRE i n build i ngs or areas con t ig uous (in actua l co nta ct with or imm ed i ate ly ad j ace n t) to any Table H-1 area not ext i ngu i shed withi n 15 minutes of control room not ific at ion or v erificat ion of a control room alarm. HU 3 Release of Toxic or F lammab le Gases Deemed Detr imental to Normal Operation of the Plant. HU3.1 2 3 4 5 6 DEF ! Report or detection o f toxic or flammable gases t hat has or could enter the site a rea boundary in amounts that can affect NORMAL PLANT OPERATIONS. HU3.2 .. I ---,,---2--,...--3

--.-4--.--5--.- ... I -D_E_F....,I Report by Lo c al , County or State Officials for e vacua t io n or sheltering of site personnel based on an offsite event. D es tr ucti v e Phenomenon Tox ic and Flammable Gas Page 2 of 8

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Prairie Island Nuclear Generating Plant Hazards Continued Security Emergenc y D irector Judgment PINGP 1576 , Rev. 10 Doc. Type/Sub Type: EP/EVT Retention: Lifetime + HOSTILE ACTION R esulting in Loss of Physical Control of the Facility_ ! 1 ! 2 3 4 5 6 i DEF A HOSTILE ACTION has occurred such that plant personnel are unable to operate equipment required to maintain safety funct io ns. 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. None HG 2 Other Conditions Exist i ng Wh i ch i n the Judgment of the Emergency D irector Warrant Declaration of General E mergency. HG2.1 ! 2 3 4 5 6 i DEF i Other conditions exist which in the judgment of the Emergency Director indicate that events are in process or have occurred which involve actual or imminent substantial core degradation or melting with potential for loss of containment int egrity. Releases can be reasonably expected to exceed EPA Prote ctiv e Action Gu ideline exposure levels offsite for more than the immediate site area. EMERGENCY ACTION LEVEL MATRIX 2 3 4 5 ! 6 DEF A HOSTILE ACTION is occurring or has occurred within the PROTECTED AREA as reported by Shift Security Supervision. HS2 Control R oo m Eva cu at ion Has Been Initiated and Plant Control Cannot Be Establ i shed. HS2.1 2 3 4 ! 5 6 DEF Control room evacuation has been i n i tiated; AND Control of the plant cannot be establ i shed per 1(2)C1.3 AOP-1 , Shutdown from Outside the Control Room or F-5 Appendix B , Control Room Evacuation (F ire) within 15 minutes. HS3 Other Conditions E xi sting Which in the Judgment of the Emergency D irector Warrant D eclaration of Site Area Emergency . HS3.1 2 3 4 5 6 i DEF ! Other conditions exist which in the judgment of the Emergency D irecto r indic ate that events are in process or have occurred which involve actual or likely major fa i l ures of plant funct ions needed for protection of the public. Any releases are not expected to result in exposure levels which exceed EPA Protecti ve Action Guideline exposure levels beyond the site boundary. or Airborne Attack T hreat. 2 ! 3 4 5 6 ! DEF A HOSTILE ACTION is occurring or has occurred within the OWNER CONTROLLED AREA as reported by Security Shift Supervision. HA4.2 .. , _,-,---2--,,---3--,,---4--,.---5---,,---6---,I-D-EF---,I A validated no t ification from NRC of an airliner attack threat within 30 minutes of the site. HA5 Control Room Evacuat ion Has Been In itia ted. HAS.1 2 3 4 5 6 DEF ! Entry into 1 (2)C1.3 AOP-1 Shutdown from Outside the Control Room or F-5 Appendix B Control Room Evacuat ion (F ire) for control room evacuation. HA6 Other Conditions Existing Which in the Judgment of the Emergency D irector Warrant Declaration of an Alert. HA6.1 2 3 4 5 6 i DEF ! Other conditions exist which in the ju dgment of the Emergency Dire ctor indicate that events are in process or have occurred which involve actual or likely potential substantial degradation of the level of safety of the plant. Any releases are expected to be limit ed to small fractions of the EPA Protective Action Guideline exposure l e vels. Table H-1 Plant Areas Area HU1.6" HU2.1" HA1.2 HA 1.3 HA1.4 HA1.5 HA2.1 HA3.1" HA3.2" RA3. -Shield/Containment Building X X X X X X X X X -Auxiliary Bu i lding X X X X X X X X X X -05/06 Diesel Generator Build ing 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 -Rela y Room X X X X X X X X X X -Turbine Bu ilding X X X X X X X X X X -Condensate Storage Tanks X X X X

Also consider areas con t i uous to these . Confirmed SECURITY CONDITION or Threat Wh ic h I nd ic ates a Potent i al Degradation in the Level of Safety of the Plant. ! 1 ! 2 ! 3 ! 4 ! 5 ! 6 ! DEF ! A SECURITY CONDITION that does NOT i nvolve a HOSTILE ACTION as re r orted b l Securit r Shift Su p ervision. HU4.2 ! 1 ! 2 _ 3 _ 4 _ 5 ! 6 ! DEF A credible PINGP security threat notification. 4 5 6 A v al idated notificat i on from NRC prov idin g information of an aircraft threat. None DEF HU5 Other Cond i tions Existing Wh i ch i n t he Judgment of the Emergenc ,y D i rector Warrant D eclaration of a UE. HUS.1 i 2 3 4 5 6 DEF ! Other conditions exist which in the judgment of the Emergency Directo r indicate that events are in process or have occurred which i ndicate a potential degradation of the level of safety of the plant. No releases of radioactive material requiring offsite response or monitoring are expected unless further degradation of safety systems occurs. Emergency Director Judgment Page 3 of 8

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Prairie Island Nuclear Generating Plant S y stem Malfunct.

Los s o f Pow e r RPS Fa il ure Inability to Reach o r Maintain Shutdown Conditions Ins t./ Comm. PINGP 1576 , Rev. 10 Doc. Type/Sub Type: EP/EVT Retention: Lifet i me + 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 15 and 16 (25 and 26); AND Either of the following: a. Restorat i on of Safeguards Bus 15 or 16 (25 or 26) with i n 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months i s not l i kely; OR b. Continuing degradation of core cooling based on F i ssion Product Barrier monitoring as indicated by Core Cooling-RED or ORANGE path. SG2 Failure of the Reactor Protection Sys t em to Complete an Automatic Trip and Manual Trip was NOT Successful and There is Indication of an Extreme Challenge to the Ability to Cool the Core. SG2.1 I 2 ! lndication (s) exist that automatic and manual tr i p were NOT successful i n reducing power to LESS THAN 5%; AND Either of the following: a. Core cooling is extremely challenged as i nd i cated by Core Cool i ng -RED path; OR b. Heat removal is extremely challenged as indicated by Heat Sink -RED path. None None EMERGENCY ACTION LEVEL MATRIX 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 power to Safeguards Buses 15 and 16 (25 and 26); AND Failure to restore power to Safeguards Bus 15 or 16 (25 or 26) with i n 15 minutes from the time of loss of both offsite and ons i te AC power. SS3 Loss of All Vital DC Power. SS3.1 i 2 3 4 ! Loss of all Safeguards DC power based on LESS THAN 112 VDC on 125VDC Pane l s 11 and 12 (21 and 22) for GREATER THAN 15 minutes. SS2 Failure of Reactor P rotection System I nstrumentation to Complete or Initiate an Automatic Reactor Trip Once a Reactor Protection System Setpoint Has Been Exceeded and Manual Trip Was NOT Successful.

SS2.1 i 2 ! lnd i cation (s) exist that automatic and manual trip were NOT suc c essful in reducing power to LESS THAN 5%. SS4 Complete L oss of Heat Removal Capability SS4.1 I 2 3 4 ! Loss of core cooling and heat s i nk as indicated by: a. Core Cooling -RED path; AND b. Heat Sink -RED path. SS6 Inability to Monitor a SIGNIFICANT TRANSIENT in Progress. SS6.1 i 2 3 4 ! Loss of most (approxima t ely >75%) or all annunciators associated with safety systems:

Ma i n Control Boards A , B-1 (2), C-1 (2), D-1 (2), E-1 (2), F-1 (2), G-1 (2) NIS Racks I , II , Ill , IV , and ERCS Alarms; AND A SIGNIFICANT TRANSIENT in progress; AND Compensatory non-alarming i nd i cat i ons are una v ailable; AND Indications needed to monitor the ability to shut down the reactor , maintain the core cooled , ma i ntain the reactor coolant system i ntact , and maintain conta i nment intact are unava i lable. AC power capability t o Safeguards Buses reduced t o a single power source for GREATER THAN 15 minutes such that any additional single failure would result in station blackout.

SA5.1 ! 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:

T r ansformer CT-11;

Transformer CT-12;

Transformer 1 RY;

Transformer 2R Y;

Diesel Generator D1 (D5);

Diesel Generator D2 (DB); AND Any add i t i onal s i ngle fa i lure w i ll resu l t i n stat i on blackout.

SA2 F ailure of Reactor Protection System I nstrumentation to Complete or Initiate an Automatic Reactor Trip Once a Reactor Protection System Setpoint Has Been Exceeded and Manual Tr i p Was Successful.

SA2.1 ! 2 3 NOTE: A failed manual trip followed by a successful manual trip redu c ing reactor powe r to less than 5% meets th i s EAL. lndication (s) exist that a Reactor Protection System setpoint was exceeded; AND RPS automatic trip did not reduce power to LESS THAN 5%; AND Any of the follow i ng operator act i ons are successful i n reduc i ng power to LESS THAN 5%, Manual Control Board:

Reactor Trip

AMSAC/DSS Actuation

Turbine Trip None SA4 UNPLANNED Loss of Most or All Safety System Annunciation or Indication in Control Room With Either (1) a SIGNIFICANT TRANSIE N T in Progress , or (2) Compensatory Non-Alarming Indicators are Unavailable. SA41 i 2 3 4 I UNPLANNED loss of most (approximately

>75%) or all annunciators or indicators associated with safety systems for GREATER THAN 15 minutes:

Ma i n Control Boards A , B-1 (2), C-1 (2), D-1 (2), E-1 (2), F-1 (2), G-1 (2) NIS Racks I , II , Ill , IV , and E RCS Alarms; AND Either of the following: a. A SIGNIFICANT TRANSIENT i n progress; OR b. Compensatory non-alarming indications are unavailable. Table C-1 Onsite Communicat i ons Systems Sound Powered Phones Plant Pag i ng System Plant Telephone Network Plant Rad i o System 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 GREAT E R THAN 15 minutes; AND Two Diesel Generators (D1 , D2 , D5 , DB) are supplying power to Safeguards Buses 1 5 and 16 (25 and 26). None SU2 Inability to Re a ch Required Shutdown Within Technical Specification Limits. SU2.1 ! 1 ! 2 ! 3 4 Plant is not brought t o required operat i ng mode with i n Technical Specificat i ons LCO Act i on Statement Time. SU3 UNPLANNED Loss of Most or All Safety System Annunciation or Indication in the Control Room for Greater Than 15 minutes. SU3.1 I 2 3 4 ! UNPLANNED loss of most (approximately

>75%) or all annunciators or indicators associated with safety systems for GREATER THAN 15 minutes:

Main Contro l Board A , B-1 (2), C-1 (2), D-1 (2), E-1 (2), F-1 (2), G-1 (2) NIS Racks I , II , Ill , IV , and ERCS Alarms. SUB UNPLANNED Loss of All Onsite or Offsite Communications Capabilities. SU6.1 ! 1 ! 2 3 4 Loss of all Table C-1 onsite communications capability affecting the ability to perform routine operations. sus.2 I 1 I 2 I 3 I 4 I Loss of all Table C-2 off site communications ca abilit. Table C-2 Offsite Communications System Plant Telephone Network Plant Radio System (dedicated offsite channels)

ENS Network HOT Loss of Pow e r RPS Fa i lure Inability to Reach or Ma i ntain Shutdown Conditions In s t./ Comm. S y stem Malfunct.

Page 4 o f 8

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Prairie Island Nuclear Generating Plant System Malfunct. ISFSI Events Fuel Clad Degradat i on RCS Leakage Inadvertent Critical i ty Cask Confine. Boundary PINGP 1576 , Rev. 10 Doc. Type/Sub Type: EP/EVT Retention: Lifetime + N one None Non e None None None None None EMERGENCY ACTION LEVEL MATRIX None None None Table C-1 Onsite Communications System s Sound Powered Phones Plant Paging System Plant Telephone Network Plant Radio System None SU4.1 ! 2 3 ! 4 ! Radiation M onitor 1(2)R-9 GREA TER TH A N 1.2 R/hr indicating fuel clad degradatio

n. SU4.2 ! 1 ! 2 3 4 Coolant sample acti vi ty GREA T E R TH A N Techn ic a l Specific at ion 3.4.17 Condition C allowable limits indicating fuel clad degradati on. SU5 R CS Leaka g e. 2 3 4 I Un i dentified or pressure boundary leakage GREATER T H AN 10 gpm. sus.2 I 2 3 4 I Identified leakage GREATER THAN 25 gpm. SUB Inadvertent Criticality. 3 4 I An UNPLANNED sustained positive startup r a te observed on nuclear i nstrumenta tion. Table C-2 Offsite Communications System Plant Te lephone Network P l ant Radio System (dedic ated offsite ch anne ls) ENS Network Nat ural phenomena events affec t i ng a loaded cask CO NFINEMENT BOUNDARY as indic ated by VISIBLE DAMAGE to the cask:

earthquake

tornado (and tornado missile)

fiood

lightning

snow/ ice EU1.2 Accident conditions affecting a loaded cask CONFINEMENT BOUNDARY as indicated by VISI BLE DAMAGE t o the cask:

dropped cask

tipped over cask

c as k burial

explosi on

fire EU 1.3 Any condition in the opinion of the Emergency D i rector that indicates loss of loaded fuel storage cask CONFINEMENT BOUNDARY . HOT Fuel Clad Degradation RCS Leakage Inadvertent Criticality System Malfunct.

MODE-NA Cask Confine. Boundary ISFSI Events Page 5 o f 8

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Prairie Island Nuclear Generating Plant EMERGENCY ACTION LEVEL MATRIX Fission Product Barriers PINGP 1576 , Rev. 10 Doc. Type/Sub Type: EP/EVT Retention: L ifetime + 2 3 4 I 2 3 4 I 2 3 4 I Loss of ANY two Barriers AND Loss or Potential Loss of Th i rd Barrier (Table F-1). Loss or Potent i al Loss of ANY two Barriers (Table F-1 ). ANY Loss or ANY Potential Loss of EITHER Fuel Clad OR RCS (Table F-1). ANY Loss or ANY Potential Loss of Containment (Table F-1). Table F-1 FISSION PRODUCT BARRIER REFERENCE TABLE NOTE Determine which combination of the three barriers are lost or have a potentia l 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 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months ). In this imminent loss situation use judgment and classify as if the thresholds are exceeded. Fuel Cladding Barrier RCS Bar r ier Containment Ba r rier 0 Loss D 1. Crit i cal Safety Function Status Core-Cooling Red. 2. Pr i mary Coo l ant Acti vi ty Le v e l Coolant A c ti v ity GREATER THAN 300 µC i/gm 1-131 equ iv alent. 3. Core Ex i t Thermocouple Read i ngs GREATER THAN 1200 degree F. 4. Reactor Vessel Water Level Not Applicable. 5. Containment R ad i ation Monitoring Conta i nment rad mon i t o r 1 (2)R-48 or 49 reading GREATER THAN 200 R/h r. 6. Other Indications Not Applicable

7. Emergency D i rector Judgment Any condition in the op i n i on of the Emergency Director that indicates Loss of the Fuel Clad Barrier. D Potential Loss 1. Critical Safety Function Status Core Cooling-Orange

OR Heat Sink-Red. D 2. Primary Coo l ant Acti v ity Le v el Not Appl ic ab l e. D 3. Core Ex i t Thermocouple Readings GREATER THAN 700 degree F. D 4. Reactor Vessel Water Level Level LESS THAN: 40% RVLIS Full Range (no RCPs); 30% RVLIS Dynamic Head Range (1 RCP); 60% RVLIS Dynamic Head Range (2 RCPs). 5. Containment Radiation Monitoring Not Appl i cable. D 6. Other I ndications Not Applicable. D 7. Emergen c y D i re c tor Judgment Any condition in the opin i on of the E mergency Director that indicates Potential Loss of the F uel Clad Barrier. D Loss 1. Critical Safety Function Status Not Applicable. 2. RCS Leak Rate GREATER THAN a v ailable makeup capac i t y as ind i cated by a loss of RCS subcool i ng LESS THAN 21 [40)" degree F.

Adverse containment conditions are defined as a containment pressure greater than 5 p s ig or conta i nment rad i ation le v el greater than 1E4 R/Hr. 3. SG Tube Rupture SGTR that results in an ECCS (SI) Actuation. 4. Containment Radiation Monitoring Conta i nment rad monitor 1 (2)R-4 8 or 49 reading GREATER THAN 7 R/h r. 5. Other I ndications Not Applicable. 6. Emergen c y Director Judgment Any cond i t i on in the opinion of the Emergency Director that indicates L oss of the RCS Barrier. D Potential Loss 1. Critical Safety Function Status RCS In t egrity-Red

OR Heat Sink-Red. 2. RCS Leak Rate Un i solab l e leak exceed i ng 60 gpm. 3. SG Tube Rupture Not Applicable. 4. Containment Radiation Monitoring Not Applicable. 5. Other Indications Not Appl i cable. 6. Emerg e n c y Director Judgment An y c on d i t i on i n the op i n i on of the E mergenc y Director that i ndicates Potential Loss of the RCS Barrier. D Loss 1. Cr i tical Safety Function Status Not Applicable. 2. Conta i nment Pressure Rap i d unexp l a i ned decrease fo llow ing i n i t i a l i ncrease; OR Containment pressure or sump level response not consistent with LOCA conditions. 3. Core Exit Thermocouple Read i ngs Not Applicable. 4. SG Secondary Side Release with P-to-S Leakage RUPTURED S/G is also FAUL TED outside of conta i nment; OR P ri mary-to-Secondary leak rate GREAT E R T H AN 10 gpm wi th non i solable steam release from affected S I G to the en v ironment.

5. CNMT Isolation Valves Status After CNMT Isolation Containment isolation Val v e (s) not closed; AND D i rect pathway to the env i ronment exists after Containment Isolat i on signa l. 6. Significant Rad i oactive In v entory in Containment Not Applicable. 7. Other Ind i cat i ons Not Applicable. 8. Emergency Director Judgment Any condition in the opinion of the Emergency Director that indicates Loss of the Containment Barrier. D Potential Loss 1. Cri t ical Safety Function Status Containment

-Red. 2. Conta i nment Pressure 46 PSIG and increas i ng; O R Containment hydrogen concentration GREATER THAN OR EQUAL TO 6%; O R Containment pressure GR E ATER THAN 23 psig with LESS THAN one fu ll train of depressurization eq u ipment operat i ng. 3. Co r e E xit Thermocouple Readings Core exit thermocouples in excess of 1200 degrees F and restoration procedures not effective within 15 minutes; O R Co r e ex i t thermocoup l es i n excess of 70 0 degrees F w i t h reactor v esse l l eve l belo w 40% RVLIS Full Range and restorat i on procedures not effecti v e w i thin 15 minutes. 4. SG Secondary Side Release with P-to-S Leakage No t Applicable

5. CNMT Isolation Valves Status After CNMT Isolation No t Applicable. 6. Significant Radioactive Inventory to Containment Cont a inment rad monitor 1(2)R-48 or 49 reading G R EATER THAN 800 R/hr. 7. Other I nd i cat i ons Not Applicable. 8. Emergency D i rector Judgment Any condition in the opinion of the Emergency Director that indicates Poten t ial Loss of the Containment Barrier. HOT Fission Product Barriers Page 6 of 8

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Prairie Island Nuclear Generating Plant Cold SD I Refuel System Malfun ct. PINGP 1576 , Rev. 10 Loss o f P ower Reacto r Vessel Level Doc. Type/Sub Type: EP I EVT Retention: Lifet i me + None CG1 Loss of RPV Inventory Affecting Fuel Clad Integrity with Containment Challenged with Irradiated Fuel in the RPV. CG 1.1 I i i I I 5 I 6 I 1. Loss of RPV inventory as indicated by unexplained level incre a se in Containment Sumps A or C , or Waste H oldup Tank as indicated by sump pump run times , leve l s , or a l a r ms; AND 2. RPV Le vel: a. LESS THAN 63% RVUS Full Range for GREATER THAN 30 minutes; OR b. cannot be monitored , with i ndication or core uncovery for GREATER THAN 30 minutes as evidenced by one or more of the following:

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

Erratic Source Range Mon itor Indication

AND 3. Indicat ion of CONTAINMENT challenged as i ndicated by one or mor e of the follow i ng

Containment hydrogen concentration GREATER THAN OR EQUA L T06%

CONTAINMENT CLOSURE not established

Containment pressure GREATER THAN 1.0 psig with CONTAINMENT CLOSURE established. EMERGENCY ACTION LEVEL MATRIX None CS 1 Loss of RPV Inventory Affecting Core Decay Heat Removal Capability. CS 1.1 i i i 5 With CONTA I NMENT CLOSU RE not established

a. RPV inventory as indicated by R PV level LESS THAN 73% RV LI S Full Range; OR b. RPV level cannot be monitored for GREATER THAN 30 minutes with a loss of RPV inventory as i ndicated by unexplained level incre ase in Containment Sumps A or C , or Waste Ho l dup Tank as indicated by sump pump run times , levels , or alarms. C S1.2 I i 5 With CONTAIN ME N T CLOSUR E established

a. RPV inventory as indicated by R P V level LESS THAN 63% RVLIS F u ll Range; OR b. RPV level cannot be mon i tored for GREATER THAN 30 minutes with a loss of RPV inventory as indic a ted by either:

Unexplained level increase in Containment Sumps A or C , or Waste H o l dup Tank as indicated by sump pump run times , levels , or alarms

Erratic Source Range Monitor Indication CS2 Loss of RPV Inventory Affect i ng Core Decay Heat Removal Capability with Irrad i ated Fue l in the RPV. N O T E: CS2.1 and CS2.2 should not be used for classification unless RP V 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 i n the Refueling mode. CS2.1 6 i With CONTAI N MENT CLOS UR E not e s tablished , and RPV l ev e l 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 N MENT CLOSU RE e s t ab lished , and RPV level cannot be monitored , with indication of core uncovery as evidenced by one or more of the following:

Conta i nment Vessel Area Mon itor R-2 read ing GREATER THAN 1000 mR/hr

Errat ic Source Range Monitor Indication Loss of All Offsite P ower and Loss of All Onsite AC Power to Safeguards Buses. I I i 5 6 i DEF 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 F ail ur e of Diese l Generators 01 a n d 0 2 (05 and 0 6) to supply power to Sa f eguards Buses 15 and 16 (25 and 26); AND Failure to restore power to Safeguards Bus 15 or 16 (25 or 26) within 15 minutes from the lime of loss of both offsite and onsite AC power. CA1 Loss of RCS Inventory. CA 1.1 i 5 Lo s s of R CS inventory a s indic a ted by R P V level at O inches Refue l ing Canal I RCS Na rr ow Range I Ultrasonic (a t or L E SS TH A N 75% R V LI S Fu l l R ange). CA1.2! i i i i 5 Loss of RCS inventory as indicated by unexplained level i ncrease in Containment Sumps A or C , or Waste Holdup Tank as in d ic ated by sump pump run times , levels , or alarms; AND RCS level cannot be monitored for GREATER THAN 15 minutes. CA2 Loss of RPV Inventory with Irradiated Fuel in the RPV. CA2.1 i 6 Loss of RPV inventory as indicated by RPV level at O inches Refueling Canal I RCS Narrow R ange I Ultrason ic. CA2.2 i i i i i i 6 Loss of RCS i nventory as indicated by unexpla i ned level increase in Containment Sumps A or C , or Waste Holdup Tank as indicated by sump pump run times , levels , or alarms; AND RPV level cannot be monitored for GREATER THAN 15 minutes. Loss of All Offsite Power to S a fegu a rds B uses for GREATER THAN 15 Minutes. s s 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 At least o n e Diese l Ge n era t or (0 1 , 0 2 , 0 5 , 0 6) is supplying power t o one of the affected safeguards buses. CU7 UNPLANNED Loss of Required DC Power for GREATER THAN 15 Minutes. CU 7.1 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 F ailure to restore power to at least one required DC panel within 15 m i nutes from the time of loss. CU2 UNPLANNED Loss of RCS Inventory with Irrad i ated Fuel i n the RPV. CU2.1 i i 6 UNPLA NN ED R CS l evel decrease below t h e RP V fiange for GREATER TH AN OR EQ U AL TO 15 mi n utes. CU2.2 i i 6 Loss of RPV inventory as indicated by unexplained level increase in Containment Sumps A or C , or Waste Holdup Tank as indicated by sump pump run times , levels , or alarms; AND RPV level cannot be monitored. COLD L o s s of Po we r Reactor Vessel Level Cold SD I R e fu e l Sy s t e m M a lfunct. Pag e 7 of 8

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P r airie Island Nuclea r Ge n erat i ng Plan t Co ld S D/ Refuel System Malfunct.

RCS Temp. C omm. Fuel Clad D egradation RC S Leakage Ina dv ertent C ritic a li t y PINGP 1576 , Rev. 10 Do c. Type/Sub Type: EP/EVT Retent i on: Lifetime + None None None N o n e None None None None None None EMERGENC Y AC T I ON LEVEL MATRIX I s s W i th CONTAINMENT CLOSURE and RCS integrity not establ i shed an UNPLANNED event results in RCS temperature exceed i ng 200'F. NOT E S 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. 2 1f the Pr es sur i zer is solid then only the R CS temperature thre s hold is appl i cable to C A4.3. CA4.2 ! 5 6 W i th CONTAINMENT CLOSURE establ i shed and RCS i ntegr i ty not estab li shed Q[ RCS inventory reduced an UNPLANNED e v ent results in RCS temperature exceeding 200°F for GREATER THAN 20 minutes 1. CA 4.3 r-1---r----.,---.,--"T"-

5-.--6--r----, An UNPLANNED e v ent results i n RCS temperature exceed i ng 20 0°F for GREATER THAN 60 minutes 1 or results i n an RCS pressu re f T None None None None T a b le C-1 O nsite Communications Systems Sound Powered Phones Plant Paging System Plant Telephone Network Plant Rad i o System U NP LA NN ED L oss of Decay Hea t R emoval Capa b ility with Irradiated Fuel in the RPV. s s I An UNPLANNED event results i n RCS temperature exceeding 200°F. s s I Loss of all RCS tempe r ature and RPV leve l indication for GREATER T H AN 15 minutes. CU6 UNPLA NN ED L oss of All Onsite or Offsi t e Communications Ca p abilities. C U6.1 ! l 5 6 Loss of all Table C-1 onsite communications capability affecting the ability to perform routine operations. cus.2 j I I I s s Loss of all Table C-2 offs i te commun i cat i ons capab i l i ty. CU5 Fuel Clad D egradation. C US.1 j 5 6 l RCS Letdown Rad Monitor 1 (2)R-9 or portable radiation monitoring instrumentation GREATER THAN 1.2 R/hr indicating fuel clad degradation. C U S.2 '"! ---,.---...----,--,....-5-..--

6-..---, Coolant sample acti v ity GREATER THAN Techn i cal Specification 3.4.17 Cond i tion C allowable lim i ts indicating fuel clad degradation. CU1 RCS L eakage. cu 1.1 i I s Un i dentified or pressure boundary leakage GREATER THAN 10 gpm. cu 1.2 i I s Ident i fied l eakage GREATER THAN 25 gpm. CUB Inadvertent Criticality. c ua.11 s s I An UNPLANNED sustained positive startup r ate observed on nuclear instrumentation. Table C-2 O ffsite Comm u nicat i ons System Plant Telephone Network Plant Radio System (dedicated offsite channels) ENS Network COLD RC S Temp. C o mm. Fu el Cl ad Degrada t i o n R C S Leakage I nad v ert en t Cr itic al ity Cold SD/ Refuel System Malfunct.

Pag e 8 o f 8

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Prairie Island Nuclear Generating Plant EMERGENCY ACTION LEVEL MATRIX -GENERAL EMERGENCY SITE AREA EMERGENCY ALERT UNUSUAL EVENT Abnormal Rad Release Rad Effluent Offsite Rad Conditions Onsite Rad Conditions PINGP 1576 , Rev. 10 Doc. Type/Sub Type: EP/EVT Retention: Lifetime + RG 1 Offsite Dose Resulting from an Actual or Imm inent Release of Gaseous Radioact i vity Exceeds 1000 mRem TEDE or 5000 mRem Thyroid COE for the Actual or Projected Durat ion of the Release Using Actual Meteorology. RG1.1 2 3 4 5 ! 6 i DEF i NOTE: If dose assessment results are available at the time of declaration , the classification should be based on RG1 .2 instead of RG 1 .1. While necessary declarations shou ld not be delayed awaiting results , the dose assessment should be initiated I completed in order to determine if the classification should be subsequently escalated. VALID reading on one or more monitors listed in Table R-1 that exceeds or expected to exceed column " GE" for 15 minutes or longer: RG1 2 i 2 3 4 5 6 i DEF ! Dose assessment using actual meteorology in dicates doses GREAT ER THAN 1000 mRem T ED E or 5000 mRem thyroid COE at or beyond the site boundary. RG1~ i 2 3 4 5 6 ! DEF Field survey resul ts indicate closed w i ndow dose rates exceeding 1000 mR/hr expected to continue for more than one hour , at or beyond site boundary; OR Analyses of field survey samples indicate thyroid COE of 5000 mRem for one hour of inhalation , at or beyond site boundary. RS1 Offsrte Dose Result i ng from an Actual or Imminent Release o f Gaseous Radioactiv ity Exceeds 100 mRem TEDE or 500 mRem Thy roi d COE for the Actual or Projected Duration of the Release. RS1 .1 2 3 4 5 6 ! DEF ! NOTE: 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 t o determine if the classification should be subsequently escalated. VALID reading on one or more monitors listed i n Table R-1 that exceeds or i s expected to exceed column " SA E" for 15 minutes or longer: RS1 .2 i 2 3 4 5 6 ! DEF Dose assessment using actual meteorology indicates doses GR E ATER THAN 100 mRem TEDE or 500 mRem thyroid COE at or beyond the site boundary. RS1 .3 i 2 3 4 5 6 ! DEF F i eld survey results ind i cate closed window dose rates exceeding 100 mR/hr expected to continue for more than one hour , at or beyond the site boundary; OR Analyses of field survey samples indicate thyroid COE of 500 mRem for one hour of inhalation , at or beyond the site boundary. Table R-1 Effluent Monitor Classification Thresholds Monitor GE SAE Alert UE Gaseous CPM CPM 1 (2) R-50 High Range Stack Gas Mon i tor 43000 mR/hr 4300 mR/hr N/A N/A 1 R-22' Sh i eld Building Vent Rad Mon i tor N/A N/A 160 , 000'/ 1.6 E5 1 , 600'/ 1.6 E3 2R-22' Shield Building Vent Rad Monitor N/A N/A 100 , 000'/ 1 E5 1,000'/ 1 E3 1 R-30' & 1R-3r Unit 1 Aux. Building Vent Rad Mon i tors N/A N/A 100 , 000'/ 1 E5 1 , 000'/ 1 E 3 2R-30' Unit 2 Aux. Building Ven t Rad Monitors N/A N/A 100 , 000'/ 1 E 5 1 , 000'/ 1 E3 2R-37" Unit 2 Aux. Building Vent Rad Monitors N IA N/A 120 , 000'/ 1.2 E5 1 , 200'/ 1.2 E3 R-35' Radwaste Building Vent Rad Monitor N/A NIA 100 , 000'/ 1 E5 1 , 000'/ 1 E3 R-25' & R-31' Spent Fuel Pool Vent Rad Monitors N/A N/A 800 , 000'/ 8 E5 8 , 000'/ 8 E3 Liquid R-18' Waste Effiuent Liqu i d Monitor N I A N/A 900 , 000'/ 9 E5 30 , 000'/ 3 E4 1R-19' SG Slowdown Ra diation Monitor N/A N/A 100 , 000'/ 1 E5 1 , 000'/ 1 E 3 2R-19' SG Slowdown Radiat ion Monitor N/A N/A 60 , 000'/ 6 E4 600'/ 6 E2 R-21 Circ Water Dischar e Monitor N/A N/A 800 , 000/ 8 E 5 8 , 000/ 8 E3 N otes: 1) ERCS EAL Alarm s i ndi ca t e an EAL threshold M ay hav e been exceeded.

Furth e r eva lu ation of th e radi ation monitor r eading is r e qu i r ed t o de t erm in e if the EAL th r eshold is exceeded.

2)" Appli es when Effluent discharge not isola t ed. RA1 Any UNPLANNED Release of Gaseous o r Liquid Rad i oactivity to the Environment that E xceeds 2 00 Times the Offsrte Dose Calculation Manua l Spec ifi cat i on for 15 M i nutes or Longer. RA 1.1 2 3 ! 4 5 6 ! DEF ! VALI D reading on any effluent monitor that exceeds 200 Times th e alarm setpoint established by a current radioactivity discharge permit for 15 minutes or longer. OR VALID reading on effluent monitor R-18 that exceeds 900 , 000 cpm for 15 minutes or longer. RA1.2 2 3 4 5 6 ! DEF VA LID reading on one or more of the following radi a tion monitors (Table R-1) that exceeds the reading shown for 15 minutes or longer: RA1.3 I 2 3 4 5 6 ! DEF ! Confirmed sample analysis for gaseous or liqu i d release indicates concentrations or release rates , w i th a release duration of 15 minutes or longer , in excess of 200 T imes ODCM specification. RA2 Da m ag e to Irradiated Fue l o r Loss of Water Le v e l that Has or Will Result in the Uncove ring of I rrad i ated Fue l Outs i de the Reactor Vessel. RA2.1 2 3 ! 4 5 6 ! DEF ! A VALi D alarm on one or more of the following radiation monitors:

R-25 or R-31 SFP Air Monitor (HI Alarm)

R-5 Fuel Handling Area Monit or reading (HI Alarm)

R-28 New Fuel Pool Criticality Area Monitor (HI Alarm)

1 (2) R-11 Ctmt/SBV Air Particulate Monitor (HI Alarm)

1 (2) R-12 Ctmt/SBV Rad io Gas Mon i tor (HI Alarm)

1 (2) R-2 Containment Vessel Area Mon i to r (HI Al ar m) RA2.2 i 2 3 4 5 6 ! DEF Water level LESS THAN 1 O feet above an irradiated fuel assembly for the reactor refueling cavity , spent fuel pool and fuel transfer canal that will result in irradiated fuel uncovering RA3 Release of Rad i oact iv e Material or Increases i n Rad i at ion Levels W i thin the Fac i l ity That Impedes Operation of S ystems Requ ir ed to Ma i nta i n Safe Operations or to Establ i sh or Ma i nta i n Cold Shutdown. RA3.1 2 3 4 5 6 ! DEF ! 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 i ation mon i toring instrumentat i on). RA3.2 i 2 3 4 5 6 ! DEF Any VALID radiation monitor rea ding GREATER THAN 1 R/hr in are as requiring infrequent access to maintain plant safet y functions (Tabl e H-1). Area -Shield/Containment Building -Auxiliary Bui lding -05/06 Diesel Generator Bu i lding -Plant Screenhouse

-Control Room -Relay Room -Turbine Bu ilding -Condensate Storage Tanks RU1 Any UNPLANNED Release of Gaseous or Liquid Rad1oact1vity to the En vironment that Exceeds Two Times the Offsite Dose Ca l culation Manual Specificat i on f or 6 0 Minutes or Longer. RU1.1 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. RU1.2 i 2 3 4 5 6 ! DEF VALID reading on one or more of the following radiation monitors (Table R-1) that exceeds the read i ng shown for 60 m i nutes or longer: RU 1.3 l 2 3 4 5 6 ! DEF ! Confirmed sample analysis for gaseous or liquid release indicates concentrations or re l ease rates , with a rele as e duration of 60 minutes or longer , in excess of two times ODCM specification. RU 2 Une x pecte d I ncreas e in Plant Rad i at i on. RU2.1 2 3 4 5 ! 6 DEF VALID indication of uncontrolled water level decrease in the reactor re f ue l ing cavity , spent fuel pool , or fuel transfer canal with all irradiated fuel assemblies remaining covered by water as indicated by level LESS THAN SFP low water level alarm , Refueling Canal Level , or visual observation (752.5 feet elevation); AND Any UNPLANNED V ALID Area Radiation Mon i tor reading increases as ind i cated by:

R-5 Fuel Handling Area Monitor reading

R-28 New Fuel Pool Criticality Area Monitor

1 (2) R-2 Containment Vessel Area Monitor

Other Portable Area Radiation Monitoring Instrumentation RU2.2 i 2 3 4 5 6 i DEF ! Any UNPLANNED VALID Area Radiation Monitor read i ng increases by a factor of 1000 ove r normal' le v e l s. 'Normal le v els can be considered as the highest reading in the past twenty-four hours excluding the current peak value Table H-1 Plant Areas HU1.6' HU2.1' HA1.2 HA13 HA1.4 HA1.5 HA2.1 HA3.1' HA3.2' X X X X X X X X X X X X X X X X X X X X X X X X X X X X X X X X X X X X X X X X X X X X X X X X X X X X X X X X X X X X X X X X X X X

Also consider areas contiguous to these. HOT & COLD Abnormal Rad Release Rad Effluent RA3.2 X X X X X Offsite Rad Conditions Onsite Rad Conditions Page 1 of 8

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Prairie Island Nuclear Generating Plant Hazards Na t ur a l & Destru c t ive Phenomenon Fir e or Explosion Toxic and Flammable Gas PINGP 1576 , Rev. 10 Doc. Type/Sub Type: EP/EVT Retention: Lifetime + None None None None None None EMERGENCY AC T ION LEVEL MATRIX N atural and De structi v e P henome na Aff e cting the Pla nt V IT AL A R EA ! 1 ! 2 3 4 5 6 j DEF Seismic Event G R EA TE R THA N Operati n g Basis Earthquake (QBE) as indicated by "OBE Exceedance" alarm on Seismic Monitoring Panel. HA 1.2 ! 1 ! 2 ! 3 ! 4 ! 5 ! 6 ! DEF ! T ornado or high winds G RE A TE R T HAN 95 mph wit h in PRO TE CT E D AR E A boundary and resulting in VISIBLE D AMAG E to any of the following plant structures

/ equipment or Control Room indication of degraded performance of those systems (T able H-1 ). HA 1.3 ! 2 3 4 5 6 j DEF Vehicle crash wi t hin PROTECTED AREA boundary and resulting in VISIBLE DAMAGE to any of the following plant structures

/ equipment the r ein or Control R oom indication of degraded performance of those systems (T able H-1 ). HA1.4 ! 1 2 3 4 5 6 j DEF ! Turbine failure-generated missiles result in any V I S IB L E DAMAG E to or penetration of any of the following plant areas (Table H-1). HA1.5 ! 1 ! 2 ! 3 ! 4 ! 5 ! 6 ! DEF ! Uncontrolled flooding in any Table H-1 area of the plant that resu l ts in degraded safety system performance as indicated in the Control Room or that creates industrial safety hazards (e.g., electric shock) that precludes access necessary to operate or monitor safety equipment.

2 3 4 5 ! 6 j DEF j High or lo w river water level occurrences affecting the PROTECTED AREA as indicated by: River intake level GREA TE R THAN 698 ft MS L; OR River intake level LESS T HAN 666. 5 ft MSL. H A2 F I RE or E X PL OS IO N Aff e cting t h e Op e rabi l ity of Plant Saf e ty Systems R e quired t o E stablish or Maint a in Safe Shut d own. HA 2.1 ! 1 ! 2 ! 3 ! 4 ! 5 ! 6 ! DEF ! FIRE or EXPLOSION in any of the following areas (Table H-1): AND Affected system parameter indications show degraded performance or plant personnel report V IS I B LE DAMAG E to permanent structures or equipment within the specified area H A3 R e l ease of T oxic or F l a mmab l e Gases Wi t hin or Con t iguous to a VI T AL AR E A Which Jeopardizes Operation of Systems Re quir e d to Mai n tain Safe Operations or E stablish or Main t ain Safe Shutdown. HA3.1 2 3 4 5 6 ! DEF ! Report or detection of toxic gases within or contiguous t o Table H-1 areas in concentrations that may result in an atmosphere IMMEDIA T ELY DANG E ROUS TO LIFE AND HEA L T H (ID L H) HA3.2 l ! 2 ! 3 ! 4 ! 5 6 j DEF j Report or detection of gases in concentration GR E ATER T H AN the L OW E R FLAMMA B ILITY LIMI T within or contiguous to Table H-1 areas. Table H-1 Plant Areas Ar e a HU1_6* HU2.1* H A1.2 HA1.3 HA14 HA1.5 H A2.1 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 Build i ng X X X X X X X X X X -P lant 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 N a tur a l and D es t ru c tive P h enomen a Aff ectin g t he PR O TE CTE D A RE A l ! 2 ! 3 4 5 6 DEF Earthquake fel t i n plant as indicated by VALID "Event" alarm on Seismic Moni t oring Pane l. HU1~ l l 2 3 4 5 6 j DEF R eport by plant personnel of torn a do or high winds G RE AT ER TH AN 95 mph striking with i n P R O TE C TE D A R EA boundary.

HU1 .3 l 1 ! 2 ! 3 ! 4 ! 5 ! 6 j DEF ! Vehicle crash in t o p l ant structures or systems within PROTEC T ED AREA boundary. HU 1 .4 ! 1 ! 2 3 4 5 6 j DEF Report by plant personnel of an unanticipated E XPLOSION within PRO TE CT E D AR E A boundary r esulting in V I SIBLE D AMAG E to permanent s t ructure or equipment.

2 3 4 5 6 l DEF ! Report of turbine fai l ure resulting in casi n g penetration or damage to turbine or generator seals. HU1.6 l ! 2 ! 3 4 5 6 ! DEF ! Uncontrolled flooding i n follow i ng areas of the plant that has the potential to affect safe t y related equipment needed for the current operating mode (Table H-1) HU1.7 ! 1 ! 2 3 4 5 6 j DEF j H igh or low river wa t er level occurrences affecting the P RO TE C TE D AREA as indicated by: River intake leve l GREATER THAN 692 ft MSL; OR R i ver intake leve l L ESS TH AN 669.5 ft MSL. H U2 F I RE Within PR O TE C TED A RE A B ound ar y Not E xtinguish e d Wi t hin 15 Minutes of Det ection. HU2.1 2 3 4 5 6 ! DEF ! FIRE in buildings o r areas contiguous (in actual contact w ith or immediately adjacent) to any Table H-1 area not extinguished within 15 minutes of con t rol room notification or verification of a control room alarm. HU 3 Release o f Toxic or Fl a mmab l e Gases D eemed Detrimental to Normal Operation of the Plant. HU3.1 2 3 4 5 6 DEF ! Report or detec t ion of toxic or fl a mm a ble gases that has or could enter the site area boundary in amounts that can affect NORMA L PLANT OPERA T IONS. HU3.2 ! 2 3 4 5 6 j DEF j Report by Local , County or State Officials for evacuation or sheltering of site personnel based on an offsite event. Natural & Destructive Phenomenon Fire o r Explosion Toxic and Flammable Gas Hazards Page 2 of 8

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Prairie Island Nuclear Generating Plant Hazards Continued S ec ur i ty Control Room Ev acuation Em erge nc y Directo r Judgment PINGP 1576 , Rev. 10 Doc. Type/S ub Type: EP/EVT Retention: Lifetime + H OS T I LE A C TI ON Res u lti n g in Los s of P hysical Control of the Facil i t y. ! 1 ! 2 3 4 5 6 i DEF ! A HOSTILE ACTION has occurred such that plant personnel are unable to operate equipment required to maintain safety functions. HG1.2 ! 1 ! 2 ! 3 ! 4 ! 5 ! 6 ! DEF A H OST IL E AC T IO N has caused failure of Spent F ue l Cooling Systems and I MM I NENT fuel d a mage is likely for a freshly off-loaded reactor core in pool. None H G2 Other Conditions E xisting Which in the Judgment of the Emergency Director Warrant Declaration of General Emergency. HG2.1 2 3 4 5 6 f DEF f Other conditions exist which in the judgment of the Emergency Director indicate that events are in process or have occurred which involve actual or imminent substantial core degradation or m e lting with potential for loss of containment integrity. Releases can be reasonably expected to exceed EPA Protective Action Guideline exposure levels offsite for more than the immediate site area. EMERGENCY ACTION LEVEL MATRIX 2 3 4 5 6 ! DEF A HOSTILE ACT I ON is occurring or has occurred within the PROTECTED AREA as reported by Shift Security Supervision. HS2 Control Room E v acuation H as Been In itiate d and Plant Control Cannot Be Establ ish ed. HS2.1 f 2 3 4 5 6 DEF Control room e v acuation has been initiated; AND Control of the plant cannot be est a blished per 1(2)C 1.3 AOP-1 , Shutdown from Outside the Control R oom or F-5 Appendix B , Control Room E vacuation (Fire) within 15 minutes. HS3 Other Conditions Existing Which in the Judgment of the E mergency D ir ector Warrant Declaration of Site Area Emergency . HS3.1 2 3 4 5 6 i DEF ! Other conditions exist which in the judgment of the Emergency D irector indicate that events are in process or have occurred which involve actual or likely major failures of plant functions needed for protection of the pub l ic. Any releases are not expected to result in exposure levels which exceed EPA Protective Action Guideline expos u re levels beyond the site boundary. or Airborne Attack Threat. 2 f 3 4 5 6 i DEF A HOSTILE ACTION is occurring or has occurred within the OWNER CONTROLLED AREA as reported by Security Sh ift Supervision. HA4.2 r! ---..--2-..--3--r--4--r--5-"T"-6--,ir-D-E-F--, A va l idated notification from N RG of an air l i n er attack t hreat within 30 minutes of the site. HA5 Control Room Evacuation Has Been Initiated. HA5.1 2 3 4 5 6 DE F f Entry into 1(2)C1.3 AOP-1 Shutdown from Outside the Control Room or F-5 Appendix B Cont r ol Room E vacuation (Fire) for control room evacuation. HA6 Other Conditions E xisting Which i n the Judgment of the E mergency D ir ector Warrant Declaration of an Alert. HA6.1 2 3 4 5 6 i DEF f Other conditions exist which in the judgment of the E mergency Director indicate that events are in process or have occurred which involve actual or likely potential substantial degradation of the level of safety of t he plant. Any r e leases are expected to be l imited to small fractions of the EPA Protective Action Guideline exposure levels. Tabl e H-1 Plant A reas Are a H U1 .6* H U2 1" H A1.2 HAU HA14 H A1.5 HA2.1 HA3.1* HA3.2* R A3. -Shield/Cont ainment Building X X X X X X X X X -Auxiliary Building X X X X X X X X X X -D5/D6 D i esel Generator Bu i ld i ng 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 -Condensa t e Stor a ge T anks X X X X

Also consider areas conti uous to these a Poten t ial Degradation in the L evel of Safety of the Plant. ! 1 f 2 f 3 ! 4 f 5 ! 6 ! DEF f A SECURITY CONDITION that does NOT involve a HOSTILE ACT I ON as re[°rted b J Securi} Shift Supervision. HU4.2 ! 1 ! 2 _ 3 _ 4 _ 5 ! 6 f DEF A credible PI N GP secu r ity threat notifica t ion. 2 3 4 5 6 A va lidated notificat i on from NRG providing information of an aircraft threat. None DEF HU5 Other Conditions Existing Which in the Judgment of the Emergency D ir ector Warrant Declarat io n of a UE. HU5.1 i 2 3 4 5 6 DEF ! Other conditions exist which in the judgment of the Emergency Director indicate that events are in process or have occurred which indicate a potential degradation of the level of safety of t he plant. No releases of radioactive material requiring offsite response or monitoring are expected unless further degradation of safety systems occurs. Em e rg ency Dir e ctor Judgmen t Page 3 of 8

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Prairie Island Nuclear Generating Plant System Malfunct.

Loss of Po wer RPS Failure Inability to Reach or Mainta i n Shutdown Cond i t i ons Inst./ Comm. PINGP 1576 , Rev. 10 Doc. Type/Sub Type: EP/EVT Retent io n: Lifetime + Prolonged Loss of All Offsite Power and Pro longed 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 15 and 16 (25 and 26); AND Either of the following: a. Restoration of Safeguards Bus 15 or 16 (25 or 26) within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months is not likely; OR b. Continuing degradation of core cooling based on Fission Product Barrier monitoring as indicated by Core Cooling-RED or ORANGE path. SG2 Failure of the Reactor Protection System to Complete an Automatic Trip and Manual Trip was NOT Successful and There is Indic a tion of an Extreme Challenge to the Ability to Cool the Core. SG2.1 I 2 ! lndication(s) exist that automatic and manual trip were NOT successful in reducing power to LESS THAN 5%; AND E ither of the following: a. Core cooling is extremely challenged as in d ic ated by Core Cooling -R E D path; OR b. Heat removal is extremely challenged as indicated by Heat Sink -RED path. None None EMERGENCY ACTION LEVEL MATRIX 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 power to Safeguards Buses 15 and 16 (25 and 26); AND Failure to restore power to Safeguards Bus 15 or 16 (25 or 26) within 15 minutes from the time of loss of both offsite and onsite AC power. SS3 Loss of All Vital DC Power. SS3.1 ! 2 3 4 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. SS2 Failure of Reactor Protection System Instrumentat ion to Complete or In itiate an Automatic Reactor Trip Once a Reactor Protection System Setpo int Has Been Exceeded and Manual Trip Was NOT Successfu l. ss2.1 I 2 I lndication (s) exist that automatic and manual trip were NOT successful in reducing power to LESS THAN 5%. SS4 Complete Loss of Heat Removal Capability. SS4.1 I ! 2 3 4 Loss of core cooling and heat sink as indicated by: a. Core Cooling -RED path; AND b. Heat S ink -RED path. SS6 Inability to Monitor a SIGNIFICANT TRANSIENT in Progress. SS6.1 i 2 3 4 ! Loss of most (approximately

>75%) or all annunciators assoc i ated with safety systems:

Main Control Boards A , 8-1 (2), C-1(2), D-1 (2), E-1(2), F-1 (2), G-1(2) NIS Racks I , II , Ill , IV , and ERGS Alarms; AND A SIGNIFICANT TRANSIENT in progress; AND Compensatory non-alarm in g i nd ic ations are unavailable

AND Indications needed to monitor the abi li ty to shut down the reactor , maintain the core cooled , maintain the reactor coolant system i ntact , and maintain containment intact are unavailable. AC power capability to Safeguards Buses reduced to a single power source for GR E ATER THAN 15 minutes such that any additional single failure would result in station blackout.

SA5.1 ! 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;

Transformer 1RY;

Transformer 2RY;

Diesel Generator D1 (D5);

Diesel Generator D2 (D6); AND Any additional single failure will resu lt in stat ion blackout.

SA2 Failure of Reactor Protection System Instrumentation to Complete or Initiate an Automatic Reactor Trip Once a Reactor Protection System Setpoint Has Been Exceeded and Manual Trip Was Successful.

SA2.1 I 2 3 ! NO T E: A failed manual trip followed by a successful manual trip reducing reactor power to less than 5% meets th is EAL. lndication(s) exist that a Reactor Protect ion System setpo int was exceeded; AND RPS automati c trip did not reduce power to LESS THAN 5%; AND 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 None SA4 UNPLANNED Loss of Most or All Safety System Annunciation or Indication in Control Room With Either (1) a SIGNIFICANT T RA N S I ENT in Progress , or (2) Compensatory Non-Alarming Indicators are Unavailable. SA4 1 i 2 3 4 ! UNPLANNED loss of most (approximately

>7 5%) or all annunciators or indicators assoc i ated with safet y systems for GREATER THAN 15 minutes:

Main Control Boards A , 8-1 (2), C-1 (2), D-1 (2), E-1 (2), F-1 (2), G-1(2) NIS Racks I , II , Ill , IV , and ERGS Alarms; AND Either of the following: a. A SIGNIFICANT TRANSIENT in progress; OR b. Compensatory non-alarming indica t ions are unavailable. Table C-1 Onsite Commun i cations Systems Sound Powered Phones Plant Pag ing System Plant Te lep hone Network P lan t Rad io System 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 i n a loss of all offsite power to both Safeguards Buses 15 and 16 (25 and 26) for GREATER THA N 15 minutes; AND Two Diesel Generators (D1 , D2 , D5 , D6) are supplying power to Safeguards Buses 15 and 16 (25 and 26). None SU2 Inability to Reach Required Shutdown Within Technical Specification Limits. SU2.1 ! 1 ! 2 ! 3 4 Plant i s not brought to requ ire d operating mode within Technical Spec ific at ions LCO Action Statement Time. SU3 UNPLANNED Loss of Most or All Safety System Annunciation or Indi cation in the Control Room for Greater Than 15 minutes. SU3 1 I 2 3 4 ! UNPLANNED loss of most (approximately

>75%) or all annunciators or indicators assoc i ated with safety systems for GREATER THAN 15 minutes:

Main Control Board A , 8-1(2), C-1(2), D-1 (2), E-1 (2), F-1(2), G-1 (2) NIS Racks I , II , Ill , IV , and ERGS Alarms. SU6 UNPLANNED Loss of All Onsite or Offsite Communications Capabilities SU6.1 ! 1 ! 2 3 4 Loss of all Table C-1 onsite communications capability affecting the abil ity to perform routine operations. SU6.2 ! 1 ! 2 ! 3 ! 4 L oss of all Table C-2 off site communications ca abilit . Table C-2 Offsite Communications System Plant Telephone Network Plant Rad io System (de d ic ated offsite channels)

ENS Network HOT Loss of Powe r RPS Fa i lu r e Inability to Reach or Maintain Shutdown Conditions Ins t./ Comm. System Malfunct.

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Prairie Island Nuclear Generating Plant System Malfunct. ISFSI Events Fuel Clad Degradation RCS Leakage Inadvertent Criticality Cask Confine. Boundary PINGP 1576 , Rev. 10 Doc. Type/Sub Type: EP/EVT Retention: Lifetime + None None N o ne None None None None None EMERGENCY AC T ION LEVEL MATRIX None None None Table C-1 Onsite Communications Systems Sound Powered Phones Plant Paging System P lant Telephone Network Plant Radio System None SU4.1 ! 2 3 4 ! Radialion Monitor 1(2)R-9 GREAT ER THAN 1.2 R/hr indicaling fuel clad degradation. SU4.2 ! 1 ! 2 3 4 Coolant sample activity GREA T ER THAN T echn ica l Specification 3.4.17 Condition C allowable limits i nd icatin g fuel clad degradation.

SU5 RCS Lea kage. 2 3 4 I Un i dentified or pressure boundary leakage GREATER THAN 10 gpm. sus.2 I 2 3 4 I Ident i fied leakage GREATER T H AN 25 gpm. SUB Inad v ertent Crit i cality. 3 4 I An UNPLANN ED sustained positive startup rate observed on nuclear instrumenta ti on. Table C-2 Offsite Communica tio ns System Plant Telephone Network Plant Rad i o System (dedicated offsite channels) ENS Network 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 EU 1.2 Accident cond i t i ons affecting a loaded cask CONFINEMENT BOUNDARY as indi c ated by VISIBLE DAMAGE to the cask:

dropped cask

tipped over cask

cask burial

explos i on

fire EU1.3 An y cond i t io n in the op i n i on of the Emergenc y Director that indicates loss of loaded fuel storage cask CONFINEMENT BOUNDARY . HOT Fuel Clad Degradation RCS Leakage Inadvertent Criticality System Malfunct.

MODE-NA Cask Confine. Boundary ISFSI Events Page 5 o f 8

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Prairie Island Nuclear Generating Plant EMERGENCY ACTION LEVEL MATRIX Fission Produ c t Barrie r s PINGP 1576 , Re v. 10 Doc. T y pe/Sub T ype: EP/EVT Retention: Lifet ime + 2 I 3 4 2 3 4 I 2 3 4 1 Loss of ANY two Barr i ers AND Loss or Potential Loss of Third Barrier (Table F-1). L oss or Potent i a l Loss of ANY two Barr iers (Table F-1 ). ANY Loss or ANY Potential Loss of EITHER Fuel Clad OR RCS (Tab le F-1). ANY L oss or ANY Potent ial Loss of Containment (Tab le F-1). Table F-1 FISSION PRODUCT BARRIER REFERENCE TABLE NOTE Determ ine which combi nation of t he 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 L oss or Potential Loss thresholds is i mm ine nt (i.e., within 1 to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months ). I n this imminent loss situation use ju dgment and classify as if the thresholds are exceeded. Fuel Cladd i ng Barrie r RCS Barrier Containment Ba r rier D Loss D 1. Critical Saf ety Function Status Core-Cooling Red 2. Primary Coolant Activity Level Coolant Activity GREATER TH AN 300 µCi/gm 1-131 equ iv alent. 3. Core E xit T hermocouple Readings GREATER THAN 1200 degree F. 4. Reactor Vessel Water L evel N ot Applic ab le. 5. Containment Radiat ion Monitoring Containment rad monitor 1(2)R-48 or 49 reading GREATER TH A N 200 R/hr. 6. Other Indications Not Applicable

7. Emergency D i rector Judgment Any condition in the opinion of the Emergency D i rect or that indicates Loss of the Fuel Clad Barrier. D Po t en t ia l Los s 1. Critical Safety Function Status Core Cooling-Orange

OR Heat Sink-Red. D 2. Primary Coolant Activi t y Le v e l N ot Applicable. D 3. Core E xit Thermocouple Read ings GREATER THAN 700 degree F. D 4. Reactor Vessel Water Level Le vel LE SS THAN: 40% RVLIS Full Range (no RCPs); 30% RVLIS Dynamic Head Range (1 RCP); 60% RVLIS Dynamic Head Range (2 RCPs). 5. Containment Rad i at ion Monitoring Not Applicable. D 6. Other Indications Not Applicable. D 7. Em ergency D i rector Judgment Any condition in the opin io n of the Emergency D i rector that indicates Potent ial Loss of the Fuel Clad Barrier. D Loss 1. Critical Safety Funct ion Status Not Applicable. 2. RCS Leak Rate GREATER THAN available makeup capaci t y as i ndicated by a loss of RCS subcooling LESS TH AN 21 [40]' degree F.

Adverse contai nm ent con ditions are defined as a co nt ainment pressure greater than 5 psig or containment radiation level greater than 1E4 R/Hr. 3. SG Tube Rupture SGTR that resu lts in an EC CS (S I) Actuation. 4. Containment Radiation Monitoring Containment rad monitor 1 (2)R-48 or 49 reading GREATER THAN 7 R/h r. 5 Other Indications Not Applicable. 6. Em er gen cy Director Judgment Any condi ti on in the opin io n of the Emergen cy Director that indicates Loss of the RCS Barrier. D P o t entia l Loss 1. Critical Safety Funct ion Status RCS In tegri t y-Red; OR Heat Sink-Red. 2. RCS L eak Rate Unisolable leak exceeding 60gpm 3. SG Tube Rupture Not Applicable. 4. Containment Radiation Monitoring N ot Applicable. 5. Other Ind ic at io ns Not Applicable. 6. Emergency D i rector Judgment Any condition in the opinion of the Emerg ency Director that indicates Potent ial Loss of the RCS Barrier. D Los s 1. Critical Safety Functi on Status N ot Applicable. 2. Containment Pressure Rap i d unexplained decrease following i n i tial increase; OR Containment pressure or sump level response not consistent with LOCA conditions. 3. Core E xit Thermocouple Readi ngs N ot Applicable. 4. SG Secondary Side Release wi th P-to-S Leakage RUPTURED S/G is also FAUL TED outside of containment

OR Pr i mary-to-Secondary leak rate GREAT E R T H AN 10 gpm with noniso l able steam release from affected S I G to the environment.

5. CNMT I so l ation Valves Status After CNM T I solat io n Containment isol ation Valve (s) not closed; AND D irect pathway to th e environment ex i sts after Co ntain ment Iso l ation s i gnal. 6. Sig nifi cant Rad io act ive In v entory in Con ta inment Not Applicable. 7. Other Ind icatio ns N ot Applicable. 8. Emergen cy D irector J udgment Any condition i n the opinion of the Emergency D i rector that indicates Loss of the Containment Barrier . D Potential Loss 1. Cri t ic al Safety Funct io n Status Conta inme nt-Red. 2. Containment Pressure 46 PSIG and i ncreasing; O R Containment h ydrogen concentrat ion GREATER THAN OR EQUAL TO 6%; OR Co n tainment pressure GREATER THAN 23 psig with LE SS THAN one full train of depressurization equipment operating. 3. Core E xit Therm ocouple Readings Co r e exit thermo cou ples in excess of 1200 degrees F and restoration procedures not effective within 15 minutes; O R Co r e exit thermocouples in e xcess of 700 degrees F with reactor vessel level bel ow 40% RVLIS F ull Range an d restoration procedures not effecti v e within 15 minutes. D 4. SG Secondary Side Release with P-to-S Leakage No t Applicable

5. CNMT I so l ation Valves Status After CNMT Isolation No t Applicable. 6. Significant Radioactive In ventory to Containment Containment rad monitor 1 (2)R-48 or 49 reading GR E ATER THAN 800 R/hr. 7. Ot h er Ind ications No t Applicable. 8 E mer gen cy Director Judgment An y condition in the opinion of the Emergency Dir ector that ind ica tes Potentia l Loss of the Co n tainment Barrier. HOT Fission Product Barrier s Page 6 of 8

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Prairie Island Nuclear Generating Plant Cold SD/ Re f uel System Malfunc t. PINGP 1576 , Rev. 10 Loss of Powe r Reactor Vessel Leve l Doc. Type/Sub Type: EP/EVT Retention: Lifetime + None CG1 Loss of RPV Inventory Affecting Fuel Clad Integrity with Containment Challenged with I rradiated Fuel i n the RPV. CG 1.1 ! ! ! ! ! 5 ! 6 ! 1. L oss of RPV inventory as indicated by unexplained level increase in Containment Sumps A or C , or Waste Holdup Tank as indicated by sump pump run times , levels , or alarms; A ND 2. RPV Le vel: a. LESS THAN 63% RVLIS Full Range for GREAT ER THAN 30 minutes; OR b. cannot be monitored , with indication or core uncovery for GREATER THAN 30 minutes as evidenced by one or more of the following:

Containment Vessel Area Monitor R-2 reading GREA T ER THAN 1000 mR/hr

Erratic Source Range Monitor Indication

AND 3. Indication of CONTAINM E NT challenged as indicated by one or more of the following

Containment hydrogen concentration GREATER THAN OR EQUAL T06%

CONTAINMENT CLOSURE not established

Containment pressure GREATER THAN 1.0 psig with CONTAINMENT CLOSURE established. EMERGENCY ACTION LEVEL MATRIX None CS1 Loss of RPV Inventory Affecting Core Decay Heat Removal Capability. CS 1.1 ! ! ! 5 With CONTA I NMENT CLOSUR E not established

a. RPV inventory as indicated by RPV l evel LESS THAN 73% RVLIS Full Range; OR b. RPV level cannot be monitored for GREATER THAN 30 minutes with a loss of RPV inventory as indicated by unexplained level increase in Containment Sumps A or C , or Waste Holdup Tank as indicated by sump pump run times , levels , or alarms. CS1.2 j ! 5 With CONTAINMENT CLOSURE established

a. RPV inventory as indicated by RPV level LESS THAN 63% RVLIS Full Range; OR b. RPV leve l cannot be monitored for GREATER THAN 30 minutes with a loss of RPV inventory as ind icated by either:

Unexplained level incr ease in Containment Sumps A or C , or Waste Holdup Tank as indicated by sump pump run times , levels , or alarms

Errat ic Source Range Monitor Indication CS2 Loss of RPV Inventory Affec t ing Core Decay Heat Removal Capability with Irradiated Fuel i n the RPV. N O T E: 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 a t or above the Bottom ID , CU2 or CA2 should be used for event classification in the Refueling mode. CS2.1 6 ! With CONTAINMEN T 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 Mon itor R-2 reading GREATER THAN 1000 mR/hr

Erratic Source Range Monitor Indication CS2.2 j ! 6 With CONTAINM E NT 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 Ind ication to Safeguards Buses. ! ! ! 5 6 j DEF j Loss of power to or from Transformers CT-11 , CT-12 , 1 RY , 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 01 and 02 (05 and 06) to supply power to Safeguards Buses 15 and 16 (25 and 26); AND Failure to restore power to Safeguards Bus 15 or 16 (2 5 or 26) within 15 minutes from the time of loss of both offsite and onsite AC power. CA1 Loss of RCS Inventory. CA1.1 ! 5 Loss of RCS inventory as i ndicated by RPV le vel at O inches Refuel ing Canal I RCS Narrow Range I Ultrason ic (at or LESS THAN 75% R V LIS F ull Range). CA1.2 ! I I I ! 5 Loss of RCS inventory as indicated by unexplained le v el increase i n Containment Sumps A or C , or Waste Holdup Tank as i ndicated by sump pump run times , levels , or alarms; AND RCS level cannot be monitored for GR E AT E R TH AN 15 m i nutes. CA2 Loss of RPV Inventory with Irradiated Fuel in the RPV. CA2.1 6 ! Loss of RPV invento r y as ind icated by RP V level at O inches Refueling Canal I RCS Narrow Range I Ultrasoni c. CA2.2 ! ! ! ! ! ! 6 Loss of RCS inventory as indicated by unexplained level increase in Containment Sumps A or C , or Waste Holdup Tank as indicated by sump pump run times , levels , or alarms; AND RPV level cannot be monitored for GREATER THAN 15 minutes. Loss of All Offsite Power to Safegu a rds B uses for GREAT E R T H AN 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 a l l 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. CU7 UNPLANNED Loss of Required DC Power for GREATER THAN 15 Minutes. CU 7.1 5 ! 6 UNPLANNED Loss o f requ i red vital DC power based on LESS THAN 112 voe on 125 voe Panels 11 and 12 (21 and 22); AND Failure to restore powe r to at least one required DC panel wi th in 15 minutes from the time of loss. CU2 UNPLANNED Loss of RCS Inventory with Irradiated Fuel in the RPV. CU2.1 ! ! 6 UNPLANNED RCS level decrease be l ow the RPV flange for GREATER TH AN OR EQUAL TO 15 minutes. CU2.2 ! 6 Loss of RPV i nventory as indicated by unexplained level increase in Containment Sumps A or C , or Waste Holdup Tank as indicated by sump pump run t i mes , levels , or alarms; AND RPV level cannot be monitored. COLD Loss of Power Reactor V e ssel Le v el Cold SD/ Refuel System Malfunc t. Page 7 o f 8

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Prairie Island Nuclear Generating Plant Cold SD/ Refuel System Malfunct.

RCS Te mp. Comm. Fuel Cl a d Degradatio n RCS Leakage Inadvertent Criticality PINGP 1576. Re v. 10 Do c. T ype/S ub T y p e: EP/EVT Retention: L ifetime + None None None None None None None Non e None None EMERGENCY AC T ION LEVEL MATRIX 5 1 6 With CONTAINMENT CLOSURE and RCS integrity not established an UNPLANNED event results in RCS temperature exceed ing 200°F. NOTES 1 1f an RCS heat removal system is in operation within thi s t i me frame and RCS temperature is being reduced then this E AL is not applicable. 2 1f the Pressurizer is solid then only the RCS temperature threshold is applicable to CA4.3. CA4.2 5 6 ! With CONTAINMENT CLOSURE established and RCS in tegr ity not established Q!: RCS i nventory reduced an UNPLANNED event results in RCS temperature exceeding 200°F for GREATE R T HAN 20 minutes 1. CA43 ~,--~-~--~---,.----5-~-6-~-~, An UNPLANNED event results i n RCS temperature exceed i ng 200°F for GREATER THAN 60 m i nutes 1 or results i n an RCS pressure f R AT R TH A N i 2. None None None None Tabl e C-1 Onsite Communications Systems Sound Powered Phones Plant Paging System P lant Telephone Ne twork P l ant Rad io System UNPLANNE D L oss of Decay Heat Removal Capability with Irradiated Fuel in the RP V. 5 6 I An UNPLANNED e v ent resu lts i n RCS temperature exceeding 200°F. I 5 6 Loss of all RCS temperature and RP V level indication for GREATER THAN 15 minutes. CU6 UNPLANNED Loss of All Onsite or Offsite Communications Capab i lities. CU6.1 ! ! 5 6 Loss of all Tab le C-1 onsite comm un ications capability affecting the ability to perform rou tine operations. cu6.2 ! I I I 5 6 Loss of all Table C-2 offsite communications capability. CU5 Fuel Clad D egradation. CU5.1 I 5 6 ! RCS Letdown Rad Monitor 1 (2)R-9 or portable radiation monitoring instrumentation GREATER THAN 1.2 R/hr ind ic ating fuel clad degradatio

n. CU5.2 .-, --.---..... ------,-,-5--,.---..,.6-.----, Coolant samp l e act ivity GREATER THAN Techn ic a l Specification 3.4.17 Condition C allowable limits indica t i ng fuel clad degradation. CU1 RCS Leakage. cu1.1 I 5 I Un i dentified or pres s ure boundary leakage GR E A TER THAN 10 gpm. CU 1 2 I I 5 Ident ified leakage GREATER THAN 25 gpm. CUB Inad vertent Criticality. cua.11 5 6 I An UNPLANNED sustained pos i ti v e startup rate obse rv ed on nuclear i nstrumentat io n. Table C-2 Off s ite Communications System Plant Telephone Network Plant Radio System (dedicated offsite channels) EN S Network COLD RCS Te mp. Comm. Fuel Clad Degradation RCS Leakage Inadvert e nt Criticality Cold SD/ Refuel System Malfunct.

Page 8 o f 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 EP-114 and SPC-EP-121 by use of the engineering change process.

ML18113A054: Difference between revisions

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