L-PI-04-022, Revised EPIP Table of Contents, Revision 19 to F3-9, Emergency Evacuation, and Revision 0 to F3-17.2, Long Term Core Cooling.

Revised EPIP Table of Contents, Revision 19 to F3-9, Emergency Evacuation, and Revision 0 to F3-17.2, Long Term Core Cooling.

ML040610121
Person / Time
Site:	Prairie Island
Issue date:	02/20/2004
From:	Solymossy J Nuclear Management Co
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
L-PI-04-022 F3-17.2, Rev 0, F3-9, Rev 19
Download: ML040610121 (58)

Text

Ct to e e Prairie Island Nuclear Generating Plant Committed to Nudear xelne Operated by Nuclear Management Company, LLC L-PI-04-022 10CFR50.4 FEB 2 0 2004 U S Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555 PRAIRIE ISLAND NUCLEAR GENERATING PLANT DOCKETS 50-282 AND 50-306 LICENSE NOS. DPR42 AND DPR-60 PRAIRIE ISLAND EMERGENCY PLAN IMPLEMENTING PROCEDURES Furnished with this letter are the recent changes to the Prairie Island Nuclear Generating Plant Emergency Plan Implementing Procedures F3. This submittal includes the following documents:

INDEX:

Emergency Plan Implementing Procedures Table of Contents REVISIONS F3-9 Emergency Evacuation Rev. 19 ADDITIONS:

F3-17.2 Long Term Core Cooling Rev. 0 DELETIONS:

None TEMPORARY CHANGE DELETIONS:

None INSTRUCTIONS:

Instructions for updating the manual are included.

This letter contains no new commitments and no revisions to existing commitments.

1717 Wakonade Drive East . Welch, Minnesota 55089-9642 V-Y5 Telephone: 651.388.1121

USNRC NUCLEAR MANAGEMENT COMPANY, LLC L-PI-04-022 Page 2 As per 10 CFR 50.4, two copies have also been provided to the NRC Region IlIl Office and one to the NRC Resident Inspector. If you have any questions, please contact Mel Agen at 651-388-1121 Extension 7210.

Joseph M. Solymossy Site Vice President, Prairie Island Nuclear Generating Plant CC Steve Orth, USNRC, Region III (2 copies)

NRC Resident Inspector- Prairie Island Nuclear Generating Plant (w/o attachment)

Attachment 1717 Wakonade Drive East a Welch, Minnesota 55089-9642 Telephone: 651.388.1121

Mfst Num: 2004 - 0054 Date  : 01/29/04 FROM  : Bruce Loesch/Mary Gadient Loc  : Prairie Island TO :UNDERWOOD, BETTY J Copy Num: 515 Holder : US NRC DOC CONTROL DESK SUBJECT : Revisions to CONTROLLED DOCUMENTS Procedure # Rev Title Additions:

F3-17.2 0 LONG TERM CORE COOLING Revisions:

F3-9 19 EMERGENCY EVACUATION UPDATING INSTRUCTIONS Place this material in your Prairie Island Controlled Manual or File. Remove revised or cancelled material and recycle it. Sign and date this letter in the space provided below within ten working days and return to Bruce Loesch or Mary Gadient, Prairie Island Nuclear Plant, 1717 Wakonade Drive E.,

Welch, MN 55089.

Contact Bruce Loesch (ext 4664) or Mary Gadient (ext 4478) if you have any questions.

Received the material stated above and complied with the updating instructions Date

l PRAIRIE ISLAND NUCLEAR iTitle::

l GENERATING PLANT l Emergency Plan Implementing Procedures TOC I I Effective Date  : 01/29/04 0 lNOTE: This set may contain a partial-distribution Iy:JU~t D > d I-of'this-Document Type. Please refer to the CHAMPS lApproved By:__ __ __ I Module for specific Copy Holder Contents.l I BPA Designee ' -l - -

__ _ _ _ __ _ _ _ _ _ _ _ _ _ _ .1 _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ __l Document # Title -Rev F3-1 - ONSITE EMERGENCY ORGANIZATION 20 F3-2 CLASSIFICATIONS OF EMERGENCIES 33 F3-3 RESPONSIBILITIES DURING A NOTIFICATION OF UNUSUAL 18 EVENT F3-4 RESPONSIBILITIES DURING AN ALERT, SITE AREA, 28 OR GENERAL EMERGENCY F3-5 EMERGENCY NOTIFICATIONS' 22 F3-5.1 SWITCHBOARD OPERATOR DUTIES - 8 F3-5.2 RESPONSE TO FALSE SIREN ACTIVATION 10 F3-5.3 RESPONSE TO RAILROAD GRADE CROSSING BLOCKAGE 8 F3-6 ACTIVATION & OPERATION OF TECHNICAL SUPPORT CENTER 17 F3-7 ACTIVATION & OPERATION OF OPERATIONAL SUPPORT 16 CENTER (OSC) .

F3-8 RECOMMENDATIONS FOR OFFSITE PROTECTIVE ACTIONS 20 F3-8.1 RECOMMENDATIONS FOR OFFSITE PROTECTIVE ACTIONS FOR 13 THE ON-SHIFT EMERGENCY DIRECTOR /SHIFT MANAGER F3-9 EMERGENCY EVACUATION 19 F3-10 PERSONNEL ACCOUNTABILITY 20 F3-11 SEARCH & RESCUE 8 F3-12 EMERGENCY EXPOSURE CONTROL 14 F3-13 OFFSITE DOSE CALCULATION is F3-13.3 MANUAL DOSE CALCULATIONS 11 F3-13.4 MIDAS METEOROLOGICAL DATA DISPLAY 7 F3-13.5 ALTERNATE METEOROLOGICAL DATA 5 Page 1 of 3

PRAIRIE ISLAND NUCLEAR Title : Emergency Plan Implementing Procedures TOC GENERATING PLANT Effective Date : 01/29/04 Document # Title Rev F3-13.6 WEATHER:FORECASTING INFORMATION 12 F3-14.1 ONSITE RADIOLOGICAL MONITORING 11 F3-14.2 OPERATIONS EMERGENCY SURVEYS 10 F3-15 RESPONSIBILITIES OF THE RADIATION SURVEY TEAMS 23 DURING A RADIOACTIVE AIRBORNE RELEASE F3-16 RESPONSIBILITIES OF THE RADIATION SURVEY TEAMS 17 DURING A RADIOACTIVE LIQUID RELEASE F3-17 CORE DAMAGE ASSESSMENT 11 F3-17.1 CORE DAMAGE DETERMINATION 0 F3-17.2 LONG TERM CORE COOLING 0 F3-18 THYROID IODINE BLOCKING AGENT (POTASSIUM IODIDE) 10 F3-19 PERSONNEL & EQUIPMENT MONITORING & DECONTAMINATION 8 F3-20 DETERMINATION OF RADIOACTIVE RELEASE 18 CONCENTRATIONS F3-20.1 DETERMINATION OF STEAM LINE DOSE RATES 9 F3-20.2 DETERMINATION OF SHIELD BUILDING VENT STACK 9 DOSE RATES F3-21 ESTABLISHMENT OF A SECONDARY ACCESS CONTROL POINT 10 F3-22 PRAIRIE ISLAND RADIATION PROTECTION GROUP RESPONSE 17 TO A MONTICELLO EMERGENCY F3-23 EMERGENCY SAMPLING 19 F3-23.1 EMERGENCY HOTCELL PROCEDURE 13 F3-24 RECORD KEEPING DURING AN EMERGENCY 7 F3-25 REENTRY 8 F3-26.1 OPERATION OF THE ERCS DISPLAY 7 F3-26.2 RADIATION MONITOR DATA ON ERCS 7 F3-26.3 ERDS - NRC DATA LINK 1 Page 2 of 3

PRAIRIE ISLAND NUCLEAR  ! Title . Emergency Plan Implementing Procedures TOC GENERATING PLANT Effective Date : 01/29/04 Document # Title Rev F3-29 EMERGENCY SECURITY PROCEDURES ' - 18 F3-30 TRANSITION TO RECOVERY' .& 6 F3-31 RESPONSE TO SECURITY 'RELATED THREATS 8 F3-32 REVIEW OF EMERGENCY PREPAREDNESS DURING OR AFTER 2 NATURAL DISASTER EVENTS Page 3 of 3

PRAIRIE ISLAND NUCLEAR POWER PLANT. EMERGENCY PLAN IMPLEMENTING PROCEDURE I

M NUMBER: I M.ER.ENC'.EVEVACUATION A'CAT F3-9 I UIE53 IRI

- EMERGENCY MMUN W-Al l

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• 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.

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O.C. REVIEW DATE:

OWNER:. . . I . . .- . EFFE 1VEVATE

~f-2I T 4 l> i/ 06 M. Werner Page 1 of 27

PRAIRIE ISLAND NUCLEAR GENERATI NG PLANT EMERGENCY PLAN IMPLEMENTING PROCEDURE

... ,- . I, . .... ....

1.0 PURPOSE This procedure provides instructions for implementing an emergency evacuation of affected areas within the plant buildings or areas within the site boundaries.

2.0 APPLICABILITY This instruction SHALL apply to all plant personnel who are involved with evacuations caused by radiological hazards, fire, toxic gas, security threat, etc. This procedure does not apply to the evacuation of the general public located in affected areas beyond the site boundary.

3.0 PRECAUTIONS 3.1 The Emergency Director (ED) should consider radiation shine from the containments as well as other hazards when determining the habitability requirements of the assembly areas and evacuation routes.

3.2 When the evacuation alarm is heard, evacuate your work area while listening to specific evacuation instructions on the plant's Public Address (PA) system. If you cannot hear or understand the instructions, continue to leave the immediate area until you learn of the evacuation instructions.

3.3 When personnel are working inside Containment or the Shield Bldg, the Access Lead Radiation Protection Specialist (RPS) should establish communications with Containment Lead RPS and keep them informed of PA announcements (cannot understand PA due to echo).

4.0 RESPONSIBILITIES 4.1 The Emergency Director/Shift Manager (ED/SM) is responsible for ensuring that an appropriate evacuation (local, plant, or site) or an Early Dismissal is implemented when radiological or other conditions warrant such action.

Page 2 of 27

I PRAIRIE ISLAND NUCLEAR GENERATING PLANT PLAN IMPLtMENnNG PROCEDURE NUMBER:

EMERGENCY EVACUATION F73-9 REV: 19 4.2 The Radiological Emergency Coordinator (REC) is responsible to advise the ED/SM on choosing an appropriate Assembly Point and need for evacuation.

4.3. The Control Room -is responsible to sound the evacuation alarm and make the appropriate announcement over the plant's.public address system.

4.4 The Assembly Point Coordinator is responsible to coordinate.the activities at the designated assembly area.

4.5 A.ll plant personnel are responsible to follow the directions given'over the plant's

. public address'system and by the Asseimbly. Point Coordinator.

4.6 The Security Team is responsible to assist in the personnel accountability process and plant access control.

4.7 - The.Technical Support Center,(TSC) Coordinator is responsible for assisting

-,work group.leaders in determining nonessential personnel and reporting to ED when determination essential personnel is complete.

4.8 TSC work group leaders are responsible in determining essential and nonessential personnel.. - .. -

5.0 DISCUSSION - See Attachment 1" 6.0 PREREQUISITES*,. .. ., ,;, .<  ; .Ti'

- ~

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i ,, ~ t . '4, 4,  ; '  : 4'; - '

6.1 It has been determined that a personnel hazard exists or may exist; such as; radiological contamination, high radiological dose rates, fire, toxic gas, security threat, etc.,

OR

-6,2  :'A'Sie'Xe~'r'G'nd '"'-ii. as ei .  !- .-

6.2 ;A'Site'A'reaor

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General'Emrergehcy' hias been declared-

' . .- '4 '  : . .: - :i-:

Page'3 of 27

PRAIRIE ISLAND NUCLEAR GENERATING PLANT EMERGENCY PLAN IMPLEMENTlNG PROCEDURE NUMBER:

EMERGENCY EVACUATION F3-9 R:

KW 19 7.0 PROCEDURE 7.1 Early Dismissal An Early Dismissal may be warranted during an Alert classification if it is determined that the emergency may escalate to a higher level or may involve a hazardous release.

An Early Dismissal of personnel should proceed as follows:

7.1.1 The REC should recommend to the ED the need for Early Dismissal of nonessential personnel based on whether the event has or may involve a hazardous release or escalate to a higher classification.

7.1.2 The ED should direct the TSC Coordinator to assist work group leaders in dismissing nonessential personnel.

7.1.3 The TSC group leaders are to determine which personnel are NOT needed for short term emergency support.

7.1.4 TSC- Coordinator reports to ED with the TSC Group Leaders recommendation.

Personnel accountability is NOT necessary for Early Dismissal of nonessential personnel.

7.1.5 Once the Early Dismissal personnel have been identified, the TSC Group Leaders should direct the selected work group supervisors to instruct their selected individuals to leave the plant site and go home.

Page 4 of 27

PRAIRIE ISLAND NUCLEAR GENERATING PLANT .... - ._EMERGENCY PLAN IMPLEMENTING PROCEDURE NUMBER:

EMERGENCY'EVACILATION F3-9 REV: 19I 7.2 Local Evacuation A Plant Evacuation should be conducted If a'large' -

employee/worker population is affected.

7.2.1 IF during normal work ho'urs OR'an'Outage AND the hazard is NOT in I... the RCA, THEN'initiate Plant Evacuation (Step 7.3).

7.2.2 A Local Evacuation of a specific area of the plant may be necessary because of local hazards. A Local Evacuation may proceed as follows:

FOR SEVERE THUNDERSTORM, HIGH WINDS, OR  ;

TORNADO WARNINGS, FOLLOW AB-2, TORNADO/SEVERE THUNDERSTORM,' PROCEDURE UNTIL RADIATION PROTECTION GROUP DETERMINES DIFFERENTLY. -

A. DETERMINE assembly points using the table below as general guidance.

NORMAL WORK HOURS OFF HOURS

-LOCAL EVACUATION, AND OUTAGE - NON-OUTAGE FROM THESE AREAS ASSEMBLY AREA FOR ASSEMBLY AREA FOR

. ACCOUNTABILITY , ACCOUNTABILITY

'Containment Coentainen Poo ' 735' Basketball Court 735' Basketball Court S pentF uelPool _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _

, .- ,* Aux Bldg ...... . ,Access Control HP.Office . Operational

-. __- - - r- - - __

Support Old Admin ! ' -. "-;I i,,,,, I Center (OSC)

Turb Bldg

" New Admin:'

NPD or SBO Initiate" Office Trailers' . -- Plant, -

i iEvacuation . Security Building Contractor Fab Shop ,., (Guardhouse)

- Warehouse #1-' .... 4.... . . . .... ..

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Contractor Trailers

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.I . 4* .. I Page 5 "of27

PRAIRIE ISLAND NUCLEAR GENERATING PLANT EMERGENCY PLAN IMPLEMENTING PROCEDURE NUMBER:

1 EMERGENCY EVACUATION F3-9 REV: 19 B.. SOUND theEVACUATION ALARM.

C. ANNOUNCE the following over the plant page.

"ATTENTION ALL PLANT PERSONNEL. THERE IS A

.(hazard)

OCCURING IN . ALL PERSONNEL SHOULD EVACUATE (specify affected area)

THE USING THE NEAREST SAFE EXIT AND (specify affected area)

ASSEMBLE AT THE ".

(assembly area)

D. REPEAT the announcement.

E. DIRECT security (4318) to conduct Personnel Accountability using F3-1 0 as guidance.

7.2.3 Employees evacuating a particular area should exit via nearest 'safe" exit and proceed to (designated area), as announced by PA, to aid in determining accountability.

7.2.4 The Radiation Protection Group (RPG) should:

A. IF tornado, severe thunderstorm OR high wind warning exists, THEN notify ED if any life threatening radiological conditions exist.

B. Assume control of entry into the area for exposure control purposes.

C. Complete surveys in the area and when conditions are returned to normal, complete surveys again.

D. Recommend to the ED/SM that the area be returned to normal use or relax access control to that area.

Page,6 of 27

PRAIRIE ISLAND NUCLEAR GENERATING PLANT - EMERGENCY PLAN IMPLEMENTING PROCEDURE

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EME: EN EA.Ci; AIO-N NUMBER:

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EMERGENCY:EVACUATION F3-9 REV: 19 7.2.5 WorkGroup Le'ders orSupervisors shioud complete Accountability for their personnel. ' . . . '- -. .

- 7.2.6 Security or Senior Work Group Leader should contact CAS when Accountability has' bee completed.

7.2.7 When Accountability is satisfied, security should inform ED/SM.

7.2.8 When the affected area has been released for normal use, the ED/SM should announce over the P.A., that the affected area is now returned to normal use.

7.3 Plant Evacuation A Plant Evacuation may occur anytime and SHALL occur whenever a Site Area, or General Emergency is declared unless other constraints or circumstances make it impractical.

• IF the emergency is already being closed Xout AND' there is no threat to personnel safety, THEN a Plant Evacuation is NOT necessary.

IF both onsite Assembly Points are uninhabitable, THEN a Site Evacuation should be initiated.

. FOR SEVERE THUNDERSTORM, HIGH WINDS, OR AU N TORNADO WARNINGS, FOLLOW AB-2, TORNADO/SEVERE A THUNDERSTORM, PROCEDURE UNTIL RADIATION..

s- PROTECTION GROUP DETERMINES bIFFEFENTLY.

All nonessential plant personnel SHALL evacuate-to a designated onsite assembly' area fo accfnuntabityi and mmon'itoring, wliile emergency response personnel proceed to their respiective emergency operating centers.

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Page 76f 27

PRAIRIE ISLAND NUCLEAR GENERATING PLANT EMERGENCY PLAN IMPLEMENTING PROCEDURE NUMBER:

§U EMERGENCY EVACUATION REV:

F3-9 19 7.3.1 The Emerge~ncy; Director should ensure the following is performed during a Plant Evacuation:

'>. 2- During off-normal working hours when few non-shift Et , <tpersonnel are present, the Shift Manager/Emergency Director may choose the Operations Lounge as an assembly area, if appropriate.

4 CONSIDER RADIATION SHINE FROM THE CONTAINMENTS AS WELL AS NATURAL HAZARDS WHEN DETERMINING H4ABITABILITY REQUIREMENTS OF THE ASSEMBLY AREAS ad AND EVACUATION ROUTES.

A. Determine the wind direction and possible habitability problems at the onsite assembly areas. Choose either the North Warehouse or the Receiving Warehouse.

• May use North Warehouse if wind is from:

2360 to 3600 or 00 to 1230

• May use Receiving Warehouse if wind is from:

1230 to 3600 or 00 to 340 B. IF conditions are acceptable, THEN inform the Control Room Operator of the designated Assembly Point and direct the Operator to sound the plant evacuation alarm per Step 7.3.2.

C. Implement F3-10, "Personnel Accountability." Personnel evacuation accountability SHALL be completed within 30 minutes after the evacuation alarm is sounded.

Page. 8 of 27

PRAIRIE ISLAND NUCLEAR GENERATING PLA~r'- EMERGENCY PLAN IMPLEMENTING PROCEDURE l l ,. .- NUMBER:

E M RGN f &;E VAd AT'I ON . F3-9

'Di- Evacuate thAiry pratorso the OSC if:

1. General area radiation levels exceed 100 mR/hr, OR 2.; Recommended by the Rad Protection Group or the REC.

E.. - Direct security to verify personnel that normally work outside the plant's Protected Area evacuate to the designated assembly area.

F. IF the completion of the accountability check results in missing

- persons,-THEN direct a search'of the plant buildings in

- accordance

; withF3-11, "Search and Rescue" and F3-12,

- ; -' "Emergency Exposure Control."

G. WHEN plant conditions have stabilized, THEN direct reentry into selected areas of the plant in accordance with F3-25, "Reentry."

H. Consider dismissing personnel from the assembly area when:

Accountability has been'completed.

. . I ~. It has been determined Which work groups or individuals may be dismissed from the site.

- ' 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> staffinrg plans have been d6termined.

- Employee` Hot LineIhas beenestablished and the employee's have been notified of the phone number.

• i' 'r ! .

Page 9 bof'27

PRAIRIE ISLAND NUCLEAR GENERATING PLANT EMERGENCY PLAN IMPLEMENTING PROCEDURE 7.3.2 ,The Controi'R1~obr Operator should perform the following during a Plant Evacuation:

A. Sound the-6vacuation alarm.

l During drills, the message should begin and end with "THIS IS A DRILL."

B. Announce the following over the plant page:

"ATTENTION ALL PLANT PERSONNEL. A PLANT EVACUATION HAS BEEN DECLARED. ALL EMERGENCY ORGANIZATION PERSONNEL REPORT TO AND REMAIN AT YOUR EMERGENCY DUTY STATIONS. ALL OTHER PERSONNEL SHALL EVACUATE TO THE (specify assembly point)."

C. Repeat announcement.

7.3.3 The REC should perform the following during a Plant Evacuation:

A. IF tornado, severe thunderstorm OR high wind warning exists, THEN notify ED if any life threatening radiological conditions exist.

B. Designate an Assembly Point Coordinator to control operations at the designated assembly area.

C. Assist the ED/SM in selecting an Assembly Point.

• May use North Warehouse if wind if from:

2360 to 3600 or 00 to 1230 M use Receiving Warehouse if wind is from:

May 1230 to 3600 or 00 to 340 Page 10 of 27

PRAIRIE ISLAND NUCLEAR GENERATING PLANT  :. ". .EMERGENCY PLAN IMPLEMENTING PROCEDURE NUMBER:

EMERGENCYVACAUATION T: 3-9

.19.. REV D. Monitor the habitability.of the Auxiliary.Building and recommend to the ED/SM an evacuation -offthe.Auxiliary Building Operators to

.the OSC if the general area rdadiation levels exceed 100 mR/hr.

E. Periodically u'pdate Assembly Point Coordinator with status of emergency.

7.3.4 The Assembly Point Coordinator should perform the following during a Plant Evacuation (may initiate at time of Emergency Center staffing):

A. Contact the REC and request th6 location of the designated Assembly Point.

B. WHEN directed by the REC, THEN proceed to the designated onsite assembly area with the appropriate keys (keys are in Security Building) and set up the Assembly Point as shown in the Assembly Point floor plans, Figure 2 or Figure 3.

C. Supervise any'required monitoring or decontamination at the Assembly Point in'accordance with F3-14.1, uOnsite Radiological Monitoring," and F3-19, "Personnel and Equipment Monitoring and Decontamination." Give priority to personnel who evacuated directly out of the Radiological Controlled Area.

D.: IF contamination is highly'likely AND personnel have been dismissed from the Assembly Point, THEN monitor all vehicles departing the site in accordance with F3-19. Monitoring and/or decontamination should be performed onsite or at the PI Academy of Learning (Training Center), whichever is most practical.

• '; .,j t "'f.'*.^- .. 1..

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E. Assist in identifying personnel missed during accountability.

.. .. .. A, .

F. WHEN directed by ED or REC to dismiss personnel, THEN notify

.- -  :; ' personnel: ' '.:

• Who is to return to OSC or TSC.

What the Employee Hot Line number is.

Pa'ge 116of 27

PRAIRIE ISLAND NUCLEAR GENERATING PLANT EMERGENCY PLAN IMPLEMENTING PROCEDURE XX2L ..... NUMBER:

W EMERGENCY EVACUATION j REV:

F3-9 19 7.3.5 OSC Coordinator-should:

A. Direct OSC supervisory staff to account for assigned essential personnel NOT evacuating.

B. Comply with EPIP F3-10, Personnel Accountability when the accountability report is delivered by security.

7.3.6 Emergency Organization Supervisors should:

A. Verify the physical location and status of their essential personnel NOT evacuating plant.

B. Comply with EPIP F3-10, Personnel Accountability when the OSC Coordinator commences the accountability process.

7.3.7 Emergency Organization Support Personnel should contact your emergency organization supervisor and provide your location and physical status.

7.3.8 Personnel evacuating the plant should perform the following during a Plant Evacuation.

A. WHEN the evacuation alarm is heard, THEN evacuate your work area while listening to specific evacuation instructions on the plant's public address system.

B. IF you cannot hear OR understand the instructions, THEN continue to leave the immediate area until you learn of the evacuation instructions.

C. IF working in a contaminated area, THEN remove as much protective clothing as time permits, especially gloves, booties or rubbers.

Page 12 of 27

PRAIRIE ISLAND NUCLEAR GENERATING PLANT EMERGENCY PLAN IMPLEMENTING PROCEDURE

'I NUMBER:

EMERGENCY EVACUATION' F3-9

__ -; .- ..... ..19'REV:

D. IF wearing a double s'uit,'i~EN removalof outside clothing would only be necessary.

.~.  : 4;. , - .... .  :. ...

.S ..... .... ,- .

E. IF unable to remrove all-protective clothing, THEN inform personnel in charge at the assembly area of your condition.

r NOTE~Dduring a DRILL, remove ALL protective clothing prior to evacuating.  ; '

F. Evacuate via Security Building or as designated by the Security Team.

G. Give your badge to the Security Team member as you exit the Protected Area..,-

H. WHEN the North Warehouse is the designated assembly area, THEN all personnel enter the East Door or as directed by the Assembly Point Coordinator.

I. WHEN the Receiving Warehouse is the designated assembly area, THEN al! personnel enter the North Door or as directed by the Assembly Point Coordinator..

- . 7.3.9  !.Security Officers should perform thejfollowing during a plant

• : evacuation:.* - . -  :.

A. Assist in personnel traffic control and perform accountability

. . . .. activities accordinglto,F3-p,* PersonnqI Accountability."

.4 B. Verify personnel'have Vacbat6d 'alo of the buildings outside the plant's Protected Area.

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• . : .. .: . v .: . - .. . . . -

Page;13 of 27 '

PRAIRIE ISLAND NUCLEAR GENERAnNG PLANT EMERGENCY PLAN IMPLEMEN717ING PROCEDURE NUMBER:

EMERGENCY EVACUATION 19 7.3.10 The TSC Codrdinator should perform the following:

A. Ensure EOF Coordinator has established the Employee Hot Line.

B. Coordinate with Group Leaders in establishing 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> coverage.

C. Coordinate when personnel should return to relieve the shift.

7.4 Site Evacuation A Site Evacuation of nonessential personnel should be required when a Plant Evacuation is justified, but the onsite assembly areas are NOT habitable.

Personnel should be directed to evacuate to the parking lot and then using personal cars or plant vehicles, proceed to the offsite assembly area.

7.4.1 The Emergency Director should perform the following during a Site Evacuation:

A. Designate the Prairie Island Academy of Learning (Training Center) as the offsite assembly area, if possible.

Page 14 of 27

PRAIRIE ISLAND NUCLEAR GENERATING PLANT - _EMERGENCY PLAN IMPLEMENTNGPROCEDURE JNUMBER: F-EMERGENCYVEVACUATION lF3-9l REV: 19 In t1o the extreme'Grare'cse thatathe PlAcdun y of Learning Worke(Training Cente s NOT usable as an offsite assembly X i area, reassess, With input from the Radiological JiJ~~mergd'en~'Coordinat'or, andl tvitivcuation

Instructions, Figure 1.

l l ~2. ' A cara'va'n'may be led to the Red Wing Servce Center or g i to either the Goodhue or Dakota County Emergency

]i Worker Monitoring & Decon Centers for monitoring and decon. Appropriate notifications -tothe county sheriffs' departments should be made.

3. The evacuees may be sent directly to their homes without monitoring-If appropriate.

4. If the general public has been evacuated, evacuees may t l be sent directly to the state's 'public reception center for w ...... 4 ;monitoring & decon; -

B. Notify the Emergency Manager (EM) of the Site Evacuation and request EOF support. The EOF will be responsible for assisting the Assembly: PointCoordinator.

C. Inform the Control Room Operator of the offsite assembly point and direct the.Operator to sound the'evacuation alarm.

D. Implement F3-10, "Personnel Accountability." Personnel accountability should be completed within 30 minutes.

E. Evacuate the Auxiliary Building Operators to the OSC if:

1.- General area radiation levels exceed 100 mR/hr OR

2. Recommended by the Rad Protection Group or the REC.

F. Ensure that the Security Force has warned all personnel within the Owner Controlled Area, including all trailers, warehouse and construction sites.

Page 15 of 27

PRAIRIE ISLAND NUCLEAR GENERATING PLANT EMERGENCY PLAN IMPLEMENTlNG PROCEDURE G. IF the completion of the accountability check results in missing persons,-THEN direct a search of the plant buildings in accordance with F3-1 1, "Search and Rescue" and F3-12, "Emergency Exposure Control."

H. WHEN plant conditions have stabilized, THEN direct a reentry into selected areas of the plant in accordance with F3-25, "Reentry."

I. Consider dismissing personnel at the assembly area when accountability has been completed.

• Key personnel may be requested to return into the plant to augment the TSC or OSC staff.

• Return of personnel to the plant will have to be coordinated with Security and Rad Protection Group.

7.4.2 The Control Room Operator should perform the following during a site evacuation:

A. Sound the evacuation alarm.

l l During drills, the message should begin and end with "THIS

.I'NOE~ ISA DRILL."

B. Announce the following over the plant page.

"ATTENTION ALL PLANT PERSONNEL. PERSONNEL WITHOUT EMERGENCY ASSIGNMENTS SHALL EVACUATE THE PLANT SITE IMMEDIATELY. GET YOUR CAR KEYS AND EVACUATE TO THE PARKING LOT. USE YOUR CAR OR PLANT VEHICLE AND PROCEED TO THE (specify assembly point): ALL EMERGENCY PERSONNEL SHOULD REMAIN AT YOUR EMERGENCY OPERATING CENTERS" C. Repeat the announcement.

Page 16 of 27

.1 PRAIRIE ISLAND NUCLEAR GENERATING PLANT' EMERGENCY PLAN IMPLEMENTING PROCEDURE NUMBER:

EMERGENCY EVACUATION F3-9 REV: 19 7.4.3 The Radiological Emergency C0riidiriiator should perform the following during a SiteEvacuation: :

A. Assist the .ED in selecting anoffsite assembly area. In most cases the Prairie lsland Academy of Learning (Training Center) should be used. ' See the note after 7.4.1 .A for possible alternmatives.

B.- Direct an Assembly Point Coordinator to report to the offsite assembly area.

C. Monitor the habitability of the Auxiliary Building.

D., IF Auxiliary Building general average dose rates exceed 100'mrem/hr, THEN recommend to ED and evacuation of the

- .-Auxiliary Building Operators to OSC.

E. Consider establishing a secondary access control point in

- accordance with F3-21, "Establishment of a Secondary Access Control Point." .

7.4.4 The Assembly Point Coordinator should perform the following during a Site Evacuation:

A. WHEN directed by the REC, THEN proceed to the offsite

,,- , :I :assembly point (Training Center)'and set up the assembly area.

- . ,; " ~ .. ,* i . -!X B . ,Supervise,any.required monitoring or decontamination at the

.Asse~mbly.P~ointin accor~latnce witt, F34.1, uOnsite Radiological Monitoring," and F3-19, 'Personnel and Equipment Monitoring and Decontamination." Give priority to..personnel who evacuated directly out of the' Radiological Controlled Area.

C. IF contamination is highly likely AND personnel have been released, THEN monitor all vehicles departing from the site in accordance with F3-19 or perform monitoring and decontamination at a location further from the Pi Academy of Learning (Training Center).

Page 17 of 27'

EMERGENCY PLAN IMPLEMENTING PROCEDURE 7.4.5 Personnel evacuating the plant site should perform the following during a Site Evacuation

A. WHEN the evacuation alarm is heard, THEN evacuate your work area while listening to specific evacuation instructions on the plant's public address system.

B. IF you cannot hear or understand the instructions, THEN continue to leave the immediate area until you learn of the evacuation instructions.

C. IF working in a contaminated area, THEN remove as much protective clothing as time permits, especially gloves, booties or rubbers.

D. IF wearing a double suit, THEN removal of outside clothing would only be necessary.

E. IF unable to remove all protective clothing, THEN inform personnel in charge at the assembly area of your condition.

When an evacuation alarm sounds during a DRILL, remove lN.TE.- ALL protective clothing prior to evacuating.

F. Retrieve your personal car keys, if appropriate.

G. Evacuate via Security Building (Guardhouse) or as designated by the Security Force.

H. Give your badge to the Security Force member as you exit the Protected Area.

I. Proceed to your car or to the Assembly Point via other means.

J. IF the Prairie Island Academy of Learning (Training Center) is the offsite assembly area, THEN all personnel proceed to the PI Academy of Learning (Training Center)'s parking lot and enter the Southwest Door or as directed by the Assembly Point Coordinator.

K. IF the PI Academy of Learning (Training Center) cannot be used, THEN assemble in an evacuation caravan on the plant access road and follow the Assembly Point Coordinator to an offsite assembly area.

Page 18 of 27

.... _ ._.. __ It

' REV:

7.4.6 ' Security Officers shoiuldpi'erforrl during a Site Evacuation:

A.:. Assist in personnel trhffio coirohiridperform accountability activities according to F3-1O, ",Psrsonel Accountability."

B. Direct ,all personnel within owner controlled area to proceed to offsite as'sb'erbly area.

i'. , . . . . . I

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Pa'de 19'of27 '

EMERGENCY PLAN IMPLEMENTING PROCEDURE Figure 1 Slta Evacuation Instructions TO HN RECEPTION CENTER Page 20 of 27

PRAIRIE ISLAND NUCLEAR GENERATING PLANT EMERGENCY PLAN IMPLEMENTING PROCEDURE.

NUMBER:

EMERGENCY EVACUATIONR F3-9 REV: 19 Figure 1 Site Evacuation Instructions:

Evacuation Routes To near-site EOF

1. Exit Plant Parking Lot

2. Turn left on Wakonade Dr.

3. Proceed to PI Academy of Learning (Training Center)

4. Use Southwest Entrance To Red Wing Service Center

1. Exit Plant Parking Lot

2. Take the Plant Road to Wakonade Dr.

3. Left on Wakonade Dr. to County 18

4. Left on Country 18 to Hwy 61

5. Left on Hwy 61 Cannon River Road in Red Wing

6. Turn left on Cannon River Road to Pepin Street

7. Turn left on Pepin Street to Red Wing Service Center on the right.

To Hastings Public Works Building using Ravenna Trail

1. Exit Plant'Parking Lot

2. Take'the Plant Road to Wakonade Drive

3. Left on Wakonade Dr. to County 18

4. Right on Country 18

5. Right on Ravenna Trail to 10th Street I Smead Blvd. [Road Changes Names]

6. 10. Street / Smead Blvd to Progress Drive

7. Left on Progress Drive to Hastings, Public Works Building on the right.

To Hastings Public Works Building using Hwy 316

1. Exit Plant Parking Lot

2. Take the Plant Road to Wakonade Dr.

3. Left on Wakonade Dr. to Country 18

4. Right on County 18 to 200th Street [Road Changes Names]

5. 200' Street to Hwy 316

6. Right on Hwy 316 to Hwy 61 -

7. Right on Hwy 61 to 10"' Street

8. 10 Street / Smead Blvd. To Progressive Drive

9. Right on Progress Drive to Hastings Public Works Building on the right.

Page 21 of 27

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-4

PRAIRIE ISLAND NUCLEAR GENERATING PLANT EMERGENCY PLAN IMPLEMENTING PROCEDURE hi . '~NUMBER:

EMERGENCY EVACUATION F3-9 REV: 19 Attachment 1 Discussion The Emergency Director has four (4) options in removing personnel from an area or the plant determined by the type of emergency condition and/or magnitude of a release. The options are Early Dismissal, Local Evacuation, Plant Evacuation, or Site Evacuation.

1. Early Dismissal This is directing designated groups of nonessential personnel to leave the plant site and return to their homes. This may be initiated during an Alert classification when it is determined that conditions may escalate such that a Plant Evacuation may occur.

When an Alert is declared, the REC should initiate early dismissal assessment and advise the Emergency Director. No Early Dismissal is necessary if it is known that the Alert will be closed out in a short time. Ifthe Alert termination time is unknown or the plant may escalate to a higher classification, then an Early Dismissal should be considered.

Examples of Early Dismissal personnel are:

Visitors Vendors & Consultants Student Interns Selected Admin Staff, Document Control, Information Systems personnel Selected QC Specialists Selected personnel not necessary for immediate emergency support Page 24 of 27

_nAl- - -ft kill- ._ -6 Ak m~amfnnani, l a..aano I n~i& m l- i--i-n.a NUMBER: F-

>FS~~ ~ EMERGENCY EVACUATIOjN . RV:1 Attachmentl~Discussion, '%

2. Local Evacuation .;. .. . 3/4*...

This is an evacuation"of a spbcdfic area 6ofa plabnitbuildiig to aa'safe area Within the plant. This may be done at any emergency level to protect plant personnel from a localized hazard.

'f a larg'e'employee/worker population is affected (i.e;Turbine Bldg or New Admin

-Bldg), a Plant Evacuation should be considered, to facilitate'accountability'in a timely manner.'

During OFF HOURS with the fewer personnel on site it is better to get everyone in one location that would make accountability easier.

The OSC is probably the best location. Personnel will show up on the list as being there and they are available to assist.

• The Security Building (Guardhouse) should be the next choice.

Personnel will be log off site and will NOT show up on the list. The drawback is they need to'get back on site to assist.

Page 25 of 27

PRAIRIE ISLAND NUCLEAR GENERATING PLAINT EMERGENCY PLAN IMPLEMENTING PROCEDURE NUMBER:

EMERGENCY EVACUATION 1_3-9

__REV: 19 Attachment 1 Discussion

3. Plant Evacuation This is an evacuation of plant buildings inside the Protected Area to a designated Assembly Point outside the. Protected Area (North Warehouse or Receiving Warehouse). Personnel already outside the Protected Area should be asked to evacuate to the designated Assembly Point.

A Plant Evacuation MAY be initiated at an Alert classification and SHALL be initiated at the Site Area or General Emergency classification unless there are other constraints or circumstances that make a Plant Evacuation impractical, such as:

A. If the Site Area or General Emergency is already being closed out, then no Plant Evacuation is necessary.

B. If both onsite assembly points are uninhabitable, then a Site Evacuation should be warranted.

C. If there are high winds or tornado, then an evacuation may not be feasible.

Once plant accountability is completed, personnel at the assembly point should be sorted out according to those who go home, go to the OSC, or go to the TSC.

If the event was a contaminating event, personnel and cars should be monitored and decontaminated onsite or at the Academy of Learning (Training Center), whichever is most practical. If contamination exists beyond the site boundary or at the Training Center, personnel may use the Xcel Red Wing Service Center or the county's Emergency Worker Monitoring and Decon Centers.

Page 26 of 27

PRAIRIE isLM6 WCLEAR GENERATING PLANT ' EAERGENCY PLAN IMPLEMENTING PROCE.DURE

. l ............. . . . . . . ........... . N UM B ER:

P I N GE EMERGENI PEVACUATION PI NROCED Attaciieni Disbusion

4. Site Evacuation

• . , . *-., Io , . *  ;

• This is an evacuation of all plaint'buildings onsite (inside & outside the Protected Area) to the Academy of Leainirig (Training Center). This.should be initiated when a Plant Evacuation is justified, but the onsitebas'sembly areas are not habitable.

-' Once'plant accountability is completed, personnel at the Assembly Point should be sorted out according to thosewho'go home', go to the OSC, or go to the TSC.

If the event was a contaminating event, personnel and cars should be monitored and

-'decontaminated, as necessary and'practical. 'If con'tamination exists beyond the site boundary and at the Academy of Learning (Training Center), personnel may use the Red Wing Service Center or the county's Emergency Worker Monitoring and Decon Centers.

If the emergency resulted in evacuation of the general public, plant evacuees may be directed to the public reception center for monitoring and decontamination, as

.appropriate.

...:.1 .  : .. ;

.~ ., .. . .. .. .

Page'27o'f27

PRAIRIE ISLAND NUCLEAR GENERATING PLANT PLANT SAFETY PROCEDURE NUMBER:,

LONG TERM CORE COOLING R V3-17.2 REV: 0

' A'REFfEE t .-

• Procedure segments may be performed from memory.

. Use the procedure toverify segments are complete.

. Mark off steps within 'segment before continuing.

• Procedureshould be available at the work location.

1. 1,-- I . . - I

,,. . , -, it , ,, "

S..---,,-. .....-...

W R

. UWNLK: EFFE TIVE DAT O.C. REVIEW DATE: EFFECTIVE DATE

> /-21/-OV M. Werner I -Z? -01/

Page 1 of 25

PRAIRIE ISLAND NUCLEAR GENERATING PLANT PLANT SAFETY PROCEDURE

-,- .*4sa~NUMBER:

. F 31 LONG TERM CORE COOLING -E 07.2

_______IREV: 0 TABLE OF CONTENTS Section Title Page 1.0 PURPOSE ................................ 3 2.0 APPLICABILITY ................................ 3 3.0 PRECAUTIONS ................................ 3 4.0 RESPONSIBILITIES ............................ 3 5.0 DISCUSSION ........ . . 4 6.0 PREREQUISITES ........ . . 4 7.0 PROCEDURE ........ . . 5 7.1 Monitoring Long Term Recirculation Operation . . 5 7.2 RHR Flow Adjustment For Decay Heat Reduction .6 7.3 Refilling The RWST .. 8 7.4 Adjusting Containment Sump B pH ............................. 9

8.0 REFERENCES

................................ 10 LIST OF TABLES Table 1 High Head Recirculation Flow Blockage .11 Table 2 Low Head Recirculation Flow Blockage .12 LIST OF ATTACHMENTS Attachment 1 Background Information for F3-17.2, Long Term Core Cooling .13 Page 2 of 25

J PRAIRIE ISLAND NUCLEAR GENERATING PLANT PLANT SAFETY PROCEDURE t ,-.,:: NUMBER:

_ LONG TERM CORE COOLING F3-17.2 REV: 0 1.0 PURPOSE The purpose of this procedure is*to provide guidance for monitoring, and enhancing long-term post-LOCA recirculation operation.

Core decay heat reduction following a LOCA may enable ECCS flow reductions similar to RHR flow adjustments made during operation of RHR in a shutdown cooling lineup.

It may also be possible at some point to align one or both trains of RHR for shutdown cooling and terminate recirculation.' This procedure monitors core decay heat removal and determines if and how ECCS flow or alignment changes are warranted.

N The possibility exists for containment debris to be ingested into the recirculation system during post-LOCA recirculation operation. This debris has the potential to adversely affect post-LOCA recirculation operation. This instruction provides guidance for detecting and mitigating these affects -

2.0 APPLICABILITY K> S  : .. .....

This procedure SHALL apply to TSC staff and Control Room operators.'

-Control Room or TSC personnel may"implement this procedure. Control Room SHALL implement this procedure until the TSC is staffed, at which tim e the TSC should assume responsibility for monitoring'long term core cooling.

3.0 PRECAUTIONS High radiation levels may' be present in the Auxiliary Building.

4.0 RESPONSIBILITIES .... *,.',

4.1 Control Room Operators are responsible for the following:

• Implementation of this procedure until it is tumed over tothe TSC.

• Directing or approving the manipulation of any plant equipment.

• Implementing contingency actions for sump recirculation blockage indications.

Page 3 of 25

4.2 When the TSC is staffed, the Engineering and Operations Group Leaders are responsible for the following:

• Monitoring post-LOCA recirculation indications.

• Making recommendations to control room staff on performance of actions within this procedure.

• Recommending and developing, as necessary, contingency actions in the event that adverse affects from debris ingestion in the recirculation system are detected.

5.0 DISCUSSION This procedure provides instructions and guidance for monitoring long-term core cooling flow requirements and mitigating core cooling issues during post-LOCA sump recirculation. Specifically, the following guidelines are provided:

• Instructions for monitoring and reducing ECCS flow based on reduced decay heat generation.

• Instructions for refilling the RWST following the initiation of long-term recirculation operation. After long-term recirculation has been established, the RWST inventory has been depleted. The purpose of the RWST refill is to provide inventory for charging system operation and re-establishment of the ECCS injection mode operation capability.

• Instructions on re-establishing containment spray flow if necessary to inject additional NaOH to ensure a basic pH (pH > 7) in the sump solution.

• Instructions for monitoring ECCS recirculation mode of operation.

• Instructions for mitigating potential blockage in the ECCS system during recirculation system operation.

6.0 PREREQUISITES

• ECCS has been aligned for recirculation.

Page 4 of 25

J i PRAIRIE iSLAND NUCLEAR GENERATING PLANT PLANT SAFETY PROCEDURE

___ ~~NUMBER: F- .

LONG TER CRE COOLING-REV 1'l 'to -aa! A- ;--';.  !;.,

7.0 PROCEDURE .

A g e;lProcedure sections may be performed concurrently and In any order.. -.

7.1 Monitoring Long Term Recirculation Operation 7.1.1 Monitor the following parameters using Control Board and ERCS (TOC ECCS, RH1,'TC):

• Containment Sump Level

• RHR Pump Suction Pressure.

• RHR Pump Discharge Pressure

• RHR Flow Rate

• SI Pump Discharge Pressure

.. Sl Flow Rate

• Core exit T/Cs

• Subcooling

• RVLIS

RHR Pump motor current

. RHR Pump vibration (ERCS 1Y0631A & 1Y0632A [2Y0631A &

2Y0632A]

The indications and mitigation strategies on Table I and W Table 2 address the' most susceptible flow blockage'

- locations:

~* .~,.. *RHR Heat Exchangers Sithrottleva S vhesnorifices

f,

• Fuel assembly top and bottom nozzles Indilcatiohs and strategy for Surmp B'creen blockage is ,

provided, but not expected due to ithe larg'e opening size and I significant plugging necessary to result in degraded RHR Pump suction. -

7.1.2 Compare actual parameter-changes to trends provided on Table 1 for high head recirculation or Table 2 for low head recirculation.

Page5 of 25

PRAIRIE ISLAND NUCLEAR GENERATING PLANT PLANT SAFETY PROCEDURE 7.1.3 IF any trends indicate flow blockage per Table 1 or Table 2, THEN perform the following:

A. Verify ERO staff, if available, consulted with mitigation strategy provided in Table 1 or Table 2 B. Initiate charging flow to RCS using either of the following while continuing with Step 7.1.3C.

• Suction aligned to VCT at available VCT makeup capability

• Suction from RWST if level greater than 8%

C. Implement mitigation strategy provided in Table 1 or 2.

D. Reduce setpoint on CC heat exchanger outlet temperature controllers as low as possible.

7.2 RHR Flow Adjustment For Decay Heat Reduction Long-term RHR flow should be maintained greater than

~NO1E ~750 gpm.

7.2.1 IF the following conditions are satisfied, THEN reduce one train RHR flow or stop one RHR pump:

• Both SI pumps stopped

• Both RHR pumps on recirculation

• Core exit T/Cs less than 140 degrees

• Core exit T/C stable or decreasing

• RCS inventory stable or increasing Page 6 of 25

PRAIRIE ISLAND NUCLEAR GENERATING PLANT PLANT SAFETY PROCEDURE 4

l t.Z  : . '-tNUMBER:

LONG TERM CORE COOLING, F3-17.2 i  ; . ' .. R:

REV- 0 7.2.2 IF the following conditions are satisfiid, 'TEN a'djust RHR flow to maintain RCS temperature stable, as necessary:

Both SI pumps stopped

• One RHR 'pump on recirculation

• Core exit T/Cs less than 140 degrees

-Core exit T/Cs stabl or decreasing

• RCS inventory stable or increasing 7.2.3 IF the following 'conditions are satisfied, THEN consider aligning one RHR train for shutdown cooling using guidance provided in ES-1.1, Post LOCA'Cdoldo'Wn And Depressurization:

• Both SI pumps stopped

.. One RHR pump stopped. .- -

• RVLIS full range greater than 82%

' RVLIS stable orincreasing -

. RCS subcooling greater than 350F

' RCS subcooling stable or increasing

. . . 7. 2.

, C,

• ~~Page 7of'25'

PRAIRIE ISLAND NUCLEAR GENERATING PLANT PLANT SAFET

Y. PROCEDURE

7.3 Refilling The RWST The purpose of this section is to refill the RWST following the alignment to long-term sump recirculation. The additional water in the RWST provides a source for charging or a source to reinitiate injection, if required.

ALL OF THESE TRANSFER PROCEDURES REQUIRE ENTRY TO AREAS WITH POTENTIALLY VERY HIGH RADIATION FIELDS. FOR THE OUT-PLANT OPERATOR ACTIONS, AN RP TECHNICIAN SHOULD BE PRESENT TO IDENTIFY RADIATION FIELDS AND RECOMMEND TRAVERSE ROUTES. TO THE EXTENT PRACTICAL, TIME SHOULD BE MINIMIZED FOR ANY OPERATIONS OF SI-17-1 [2S-17-1].

! - U1 W THE COMBINED CONCENTRATION OF WATER

• -. 11-e,4;jI ws1 TRANSFERRRED TO THE RWST SHALL BE SUCH THAT THE

-,;- -S,,iS#b RESULTANT RWST BORON CONCENTRATION IS BETWEEN 2000 AND 3500 PPM.

Fill the RWST to greater than or equal to 50% using one or more of the following:

• Transfer water from the other unit RWST per C16.

• Transfer water from the CVCS Hold-Up Tank(s) per Cl2.8.

• Transfer water from the CVCS Monitor Tank(s) per C12.8.

• Refill the RWST using the BA Blender per C12.5.

Page 8 of 25

. PLANT SAFETY PROCEDURE

.. :, I!, I 7.4 Adjusting Containment Sump B pH During performance ,of E-1 and ES 2, or 1.-3, sufficient NaOH may not have been Injected to ensure that the pH of the

~NO1E ,sump solution Is greater than 7. A pH greater than 7 is

'3 X necessary to retain iodine in the sump liquid solution and

iminimize the potential for stress corrosion cracking of

stainless st el. .

-IFthe level indication In the Caustic Standpipe Is greater

,'JOTEI* c~isthan45%

charging'system AND a second'RWST or ECCS, THEN Is ittomay be injected using to be necessary adjust sump pH by establishing containment spray.

IF Containment Sump B pH adjustment is necessary, THEN perform the following:

7.4.1 Check caustic standpipe level greater than 45%

7.4.2 Verify RWST level is' greater than 50%

7.4.3 Manually actuate containment spray 7.4.4 IFonly one train CS is desired, THEN perform the following:

A. .,Reset containment spray,... -

B. Stop one containment spray pump 7.4.5 WHEN caustic standpipe'level is less than 45% OR RWST level is less than 8%, THEN perform the following:

A. Reset containment spray B. Stop containment spray pumps Page 9 of 25

PRAIRIE ISLAND NUCLEAR GENERATING PLANT PLANT SAFETY PROCEDURE

8.0 REFERENCES

8.1 NRC Bulletin 2003-01, Potential Impact of Debris Blockage on Emergency Sump Recirculation at Pressurized Water Reactors.

8.2 COM027522, Commitment from response to NRC Bulletin 2003-01. Implement a procedure revision to address potential debris ingestion. (N')

8.3 COM027520, Commitment from response to NRC Bulletin 2003-01. Implement a procedure to enhance operator guidance for refilling the RWST from alternative water sources. (N2) 8.4 OTH026878, Engineering Evaluations for NRC Bulletin 2003-01.

8.5 OTH026879, Identify RHR System Potential Blockage Locations.

8.6 OTH026881, Identify Si System Potential Blockage Locations.

8.7 OTH026882, Identify Core Passages Potential Blockage Locations.

8.8 EOP Setpoint Calculation J.5, Maximum Time Period for Interrupting Si Flow During Switchover to Recirculation.

8.9 CALC ENG-ME-005, Revision 3, Analysis of Available NPSH to the RHR Pumps from the Containment Sump.

8.10 CALC ENG-ME-557, Rev. 0, Containment Sump Liquid pH.

Page 10 of 25

PRAIRIE ISLAND NUCLEAR GENERATING PLANT . .PR PLANT SAFETY PROCEDURE I

I I NUMBER: . .

, C". " j , i," .  : -, ,*.: -- . f;.

L6NG !TERM CORE COOLiNO . F3-17.2

._ REV: Io TabledI High Head Recirculation Flow Blockage'_',_.,.:..

Tem Indications and mitigation strategies address the most

.,.susceptible flow blockage locations: RHR Heat Exchangers,,.

,-Si throttle vale-sloriflces, and nfu'el hisemibij bbitomn'noizzes.

Indication - Trend Mitigation Strategy Si Pump Suction Blockage: -_-;;_i_-_-_-_-_:,

Si Pump Discharge Pressure - . IFone train idle, THEN start idle train RHR and SI pump on DECREASING OR FLUCTUATING recirculation and stop affected train SI and RHR pump.

Si Flow Rate - . - IF both trains operating, THEN stop affected train SI and DECREASING OR FLUCTUATING RHR pumps.

RHR Pump Discharge Pressure - IFonly affected train available, THEN stop RHR and Si INCREASING pumps for time period not to exceed six minutes.

. .. .. ;Align RHR HX bypass line on affected train.

• Aig~n

.. SI pump flow to vessel rather than cold legs.

Expedite post-LOCA C/D and depressurization.

• IF RWST level greater than 8%, THEN align RHR suction to RWST as follows (Limit time of total flow stoppage to less

.. than 6 minutes):

aa Stop SI pump

b. 'Stop RHR pump

. c CLOSE Sump B to RHR suction valves

d. OPEN RWST to RHR suction valves

e. Start RHR pump .

.-- -f. 'Start Si pump . .

-. Referto ECA-1.1g SI Pump' Discharge Blockage SI Pump Discharge Pressure -

• I'ign S flow to vessel rather thlan cold legs.

INCREASING ' - . . .; .! '

• Attempt cold leg flow re-initiation after settling period.

Si Flow Rate - DECREASING '-',.X. ..

I Expedite post-LOCA.C/D and-depressurization.

RHR Pump Discharge Pressure- :.. i.. . .

INCREASING ,,cS Fuel Assembly Blockage (bottom)

Core Exit Thermocouples - INCREASING -e 'Align Si flow to vessel rather than cold legs.

SI Pump Discharge Pressure - .- Attempt cold leg flow re-initiation after settling period.

STABLE

• Expedite post-LOCA C/D and depressurization.

SI Flow Rate - STABLE P a 11. 25.o Page' 11-of 25

PRAIRIE ISLAND NUCLEAR GENERATING PLANT PLANT SAFET

Y. PROCEDURE

Table 2 Low Head. Recirculation Flow Blockage The indications and mitigation strategies address the most susceptible flow blockage locations: RHR Heat Exchangers and fuel assembly top nozzles. Sump B screen blockage strategy is provided, but screen blockage is not likely.

Indication - Trend Mitigation Strategy RHR HX Blockage RHR Pump Discharge Pressure - . IF one train idle, THEN start idle train RHR pump on INCREASING recirculation and stop affected train RHR pump.

RHR Flow Rate - DECREASING . IF both trains operating, THEN stop affected train RHR pump.

. IF only affected train available, THEN stop RHR pump for time period not to exceed six minutes.

. Throttle RHR flow as low as possible while maintaining RCS temperature stable.

. Align RHR HX bypass line on affected train.

Fuel Assembly Blockage (Top)

Core Exit T/Cs - . Align for High Head Recirculation as follows:

INCREASING OR DIVERGING a. CLOSE SI Pump suction from RWST RHR flow/pressure - STABLE b. OPEN RHR Pump discharge to SI Pump suction

c. Start SI Pump

d. CLOSE RHR vessel injection valve

. OPEN RHR to Loop B cold leg return Sump B Screen Blockage RHR Pump Discharge Pressure - . IF both trains operating, THEN stop one train RHR pump.

DECREASING OR FLUCTUATING . Throttle RHR flow as low as possible while maintaining RCS RHR Pump Suction Pressure - temperature stable or core covered.

DECREASING OR FLUCTUATING . IF only affected train available, THEN stop RHR pump for RHR Pump Motor Current - time period not to exceed six minutes.

DECREASING OR FLUCTUATiNG . .. IF RWST level greater than 8%, THEN align suction to RWST as follows (Limit time of total flow stoppage to less than 6 minutes):

a. Stop RHR Pump

b. CLOSE Sump B to RHR suction valves

c. OPEN RWST to RHR suction valve

d. Start RHR pump

e. Refer to ECA-1.1.

Page 12 of 25

PRAIRIE ISLAND NUCLEAR GENERATING PLANT PLANT SAFETY PROCEDURE NUMBER:

) F 3LONG TERM CORE COOL NG F3-17.2 REV: 0 Attachment I Background lnformati6n fo6 F3-Z, L2;ong Term Core Cooling

SUMMARY

,' *b, Following a loss of coolant accident (LOCA), qcntinued c6r'cboling may be provided using post-LOCA recirculation from the containment.sump. This procedure provides guidance for the following:

• Monitoring long-term post-LOCA recirculation operation: Numerous parameters are

.available to monitor long-term core cooling performance. Selected parameters are compared to expected trends for degrading recirculation flow conditions. Mitigating strategies are provided for'recirculation flow blockage in the most susceptible

'locations:- - ' '

• RHR flow adjustments for decay heat reduction: This section provides for RHR flow reductions and possible alignireiit of RHR for shutdown cooling based on decay heat reduction and system performance.

Refilling the RWST: Refilling or raising RWST level following transfer to recirculation may be desirable to,ensure an adequate source of water is available for ECCS'pump operation if sump recirculation flow is challenged or if containment spray pump operation is needed for sump pH adjustment and provide another source of water to the charging-pumps. This section provides direction for raising RWST level. . --

. Adjusting containment sump pH: For some event sequences it may be possible to inject the contents of the RWST without actuation of containment spray. Additional RWST water could be injecteddue to, ECCS pump realignment to address.

recirculation flow issues. For each of these conditions it may.be necessary to adjust containm'ent su'mp pH by'operation of the containment spray system. This section provides directidn for oper'ation of containment spray to raise sump pH.

The monitoring aspects of this piocedurd-can be performed foron either the Control Room or the TSC.: The TSC is the preferred.Focation for monitoring and trending long-te'rm recirculation operation; "Operation' otplnt equipment is performed by or with direction from the Control Room.,;

NRC Generic Safety Issue (GSl)A91'ias identified several potential concerns regarding post-LOCA recir6ulatio6 ciperatiin.-ln June 2003 the NRC issued Bulletin 2003-01 to request that licensees inplerment compensatory measures to reduce the potential risk associated with post-LOCA sump recirculation operation. Prairie"lsland responded to Bulletin 2003-01 in a letter dated August 6, 2003 (T-Track Item OE026823). In the response to Bulletin 2003-01, several commitments were made

..regarding implementation of procedural actions for long-term post-LOCA recirculation operation. This procedure implements actions to close several of these commitments.

Page 13 of 25

PRAIRIE ISLAND NUCLEAR GENERATING PLANT PLANT SAFETY PROCEDURE Attachment I Background Information for F3-17.2, Long Term Core Cooling (Continued)

BASES FOR ACTIONS Step 7.1 Note:

The procedure steps address different issues and are written such that they can be performed in any order as the condition or need arises.

Step 7.1 This step provides guidance for monitoring long-term recirculation operation by observing parameters for any indication, or trends, of possible degradation, comparing selected parameters to expected trends for degrading recirculation capability, and implementing appropriate mitigating strategies for degrading recirculation flow.

The most susceptible form of degradation during recirculation operation would be due to the ingestion of debris from the containment sump into the recirculation system. Due to the relatively large opening size of the recirculation sump screen (314"x 3-11/16"),

blockage of the sump screen is not expected. Thus, most of these strategies focus on mitigation of flow reduction caused by debris ingestion. Mitigation strategies are provided for sump screen blockage during RHR recirculation, as this is the time period where significant debris buildup on the screens may have a noticeable affect.

Step 7.1.1 This step provides a listing of indications of recirculation operation that can be monitored using ERCS from either the Control Room or TSC. In order to detect possible degradation of recirculation operation it is necessary to monitor several parameters. The parameters are expected to trend to stable, equilibrium values.

Although selected ERCS display turn on codes are listed, the information can be obtained using other ERCS displays and trend functions or PI Process Book. Several of the parameters are also available on meters or panel. displays in the Control Room.

Page 14 of 25

PRAIRIE ISLAND NUCLEAR GENERATING PLANT PLANT SAFETY PROCEDURE .

NUMBER:.

. G . - C.R- _ - E . I OO LONG TERM CORE COO LING F3-17.2 I REV: 0 Attachrneriti iBackgroun'd in'foma tio'fiifF31 7.,i L'on'g Termii Core Cooling

[E'-;:(C'ontinued)

Note for Step 7.1.2 Based on the smallest size openings in the recirculation system, the most likely locations for flow blockage are the RHR heat exchanger (0.652 inch), SI throttle valves or orifices (0.54 inch), fuel asserribly top nozzle (0.472 inch), and fuel assembly bottom nozzles (0.188 inch). Due to the'la'rg'e surface area of the sump screens, the large opening size in the screens (3/4"1 X 3 11/16") and the concrete "lip' around the sump, large pieces of debris and metallic items are prevented from entering the sump and

• flow blockage of the screens is not expected. Significant screen blockage must occur with high flow rates to have' any'noticeabl aff6ct on RHR pump operation. Mitigating

'strategies are provided in Table 2for sump screen blockage when high recirculation flow conditions exist'(i.e., low head recirculation).

Step 7.1.2 -

Table 1 and Table 2 compare selected parameters with expected trends indicating ECCS recirculation flow degradation.: 'Several options for mitigating strategies are

• provided for each susceptible flow blockage'location.' Table I provides indications for

' high head recirculation and Table 2 provides indications for low head recirculation.

-Although only selected parameters and trends are listed on the tables,,other parameters are being monitored and can be"used when evaluating recirculation flow status. ' -

Personnel evaluating' recirculation flow status need to be aware of and take into account other actions that have an affect on these parameters. For example, securing ECCS pumps or adjusting flow rates'ba'sed on the 'post-LOCAcooldown and depressurization procedure or Step 7.2 of-this'procedurevwill affect many of the parameters and trends may be indicative of coritrol based 6nthose procedures steps and not.degradation ofrecirculation'flow.;

Mitigation strategies for susceptible flow blobckQge i6oation6slare provided on Table 1 and Table 2. Implementation of the strategies and bases for the strategies is described in the next section.

Page 15 of 25'

PRAIRIE ISLAND NUCLEAR GENERATING PLANT PLANT SAFETY PROCEDURE Attachment I Background,lnformation for F3-17.2, Long Term Core Cooling (Continued)

Step 7.1.3 If degradation of ECCS recirculation flow is identified, then mitigating strategies should be implemented to ensure continued long-term core cooling capability. Since several options are available for each flow blockage location, available plant staff should be consulted as to the most appropriate or best method to use. Based on parameter trends, system alignments and availabilities, local component access, and personnel availability, one of the options may be preferred over another. The strategies are not listed in order of priority on the tables. Consulting available ERO staff ensures that input from operations and engineering personnel is used to correct the situation or provide compensatory measures.

Charging flow is established to the RCS to provide a diverse flow path for decay heat removal. Within a relatively short time period (after the first initial hours), the decay heat rate is within the capacity of two charging pumps. Thus, provided that the flow path exists between the charging connection to the RCS and the reactor assembly, adequate core cooling can be provided with two charging pumps. Charging flow can also be aligned to the pressurizer for flow to the hot leg, through seal injection lines to each cold leg, and to the normal cold leg return for alternate flow paths based on RCS parameter trends.

Lowering the CC heat exchanger temperature controller setpoint will increase RHR heat exchanger heat removal capability. Any added heat removal capability will provide the same core decay heat removal with lower RHR flow rates.

Page 16 of 25

PRAIRIE ISLAND NUCLEAR GENERATING PLANT . . PLANT SAFETY PROCEDURE I

3

.,,.e..NUMBER:

LONG TERM CORE COOLING F3-17.2

. .'REV: 0 Attachment I -Bac goun Lo Teir Coere Cooling (Continued)

Table 1 and Table 2 provide'mitigation strategies for the identified flow blockage locations. - i--

'Table' Table 1 provides str'ategies fo'r use during high head recirculation operation. The most susceptible blockage'6cations' are the RHR heat exchangers,' Sl throttle valves or orifices an-d the fuel assembiy bottom nozzles. The' indications and trends provided are those primrary.indi6ations expected to shoW degrading conditions. Other parameters are also being moibnitored and may provide valuable trend information when making decisions about recirculation performance.

-SI Pump Suction Blockage (e.g., RHR Heat Exchanger) strategies can be broken

,down as follows:

1. Switch to the other train if both trains are available. Since one train of ECOS pumps is idle, 'that train may not have' sustained any significant debris blockage. Stopping the operating train may allow accumulated debris to settle.' Operation of the train following a period of time for settling could later be attempted if necessary or desired..

2. 'Stopping one train of pumps, if both trains are running, may allow debris to settle.: Operation of the secured train following a period of time for settling could later be attempted if necessary or desired.

-3.- If only one-train is available'then stopping the running pumps and restarting the pumps may clear the debris by allowing it to settle or pushing it through

'the opening with the flow/pressure surge when the pumps are started. To avoid excessive core heat-6p flow to the core should not be stopped for more than six minutes. (EOP Setpoint calculation J.5 provides the basis for the six minutes maximum flow interruption time). , ,;

4. Open the RHR heat exchanger-bypass line. 'Establishing this flow path will

provide sump water to the SI pump with limited or no cooling in the RHR heat exchanger. Flow is provided to maintain the core covered and sump liquid temperature would increase to saturation. Decay heat is then removed by

-the containment fan coil units.; Analysis has demonstrated that no unacceptable conditions-would result from this scenario. (NMC Intemal Correspondence, dated 1/17/03,- Containment Response to LOCA with loss of CC at 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />," also see NAD File#OC.PX.2003.06).

Page 17 of 25

PRAIRIE ISLAND NUCLEAR GENERATING PLANT PLANT SAFETY PROCEDURE Attachment I Background Information for F3-17.2, Long Term Core Cooling

-. - (Continued)

5. Align the SI pump to the reactor vessel. Changing the SI pump discharge flow path to the hot legs and closing the cold leg injection line reduces the recirculation flow rate. The reduction in flow rate may minimize the affects of debris blockage sufficiently to provide adequate SI pump NPSH and maintain core cooling.

6. Expedite post-LOCA cooldown and depressurization. Post-LOCA cooldown and depressurization cools down the RCS, which permits RCS depressurization by sequentially stopping the SI pumps. The procedure may also place one or both trains of RHR in a shutdown cooling alignment. IF RHR can be placed in shutdown cooling alignment, then the need for recirculation is eliminated. This eliminates the potential for debris ingestion to further degrade post-LOCA recirculation operation.

7. Stop any train RHR and SI pump with inadequate suction, align pump suction to the RWST, and refer to guidance of ECA-1-1: Damage to ECCS pumps can occur within a matter of minutes if operated with inadequate suction.

Although this action may stop recirculation flow, it is necessary to preserve the pumps for future use. Loss of emergency sump recirculation procedure provides guidance for operation of ECCS pumps to preserve RWST inventory while providing sufficient injection for core cooling. Providing injection from the RWST will only be possible when inventory has been added to raise RWST level.

SI Pump Discharge Blockage (e.g., SI Pump Throttle Valves or Orifices) - These strategies can be broken down as follows:

1. Align the SI pump to the reactor vessel in lieu of the cold legs provides an alternate flow path to the RCS. If flow blockage is occurring in the cold leg injection throttle valves, then changing the flow path may allow the debris to settle or be forced through the system if flow is re-established to the cold legs in the future.

2. Re-initiating cold leg injection may clear the debris by allowing it to settle or pushing it through with the flow/pressure surge when the flow is re-initiated.

3. Expedite post-LOCA cooldown and depressurization. Post-LOCA cooldown and depressurization cools down the RCS, which permits RCS depressurization by sequentially stopping the SI pumps. The procedure may also place one or both trains of RHR in a shutdown cooling alignment. IF RHR can be placed in shutdown cooling alignment, then the need for recirculation is eliminated. This eliminates the potential for debris ingestion to further degrade post-LOCA recirculation operation.

Page 18 of 25

PRAIRIE ISLAND iUCLEAR GENERATING PLANT . *.. PLANT SAFETY PROCEDURE.

A. NUMBER:

<. . LONG TERM CORE COOLING F3-17.2

. REV: 0 Attachment 1 Background infornation for F3-17.2, Long Term Core Cooling (Continued) 4.; Stop any train RHRand Si 'purhp with iriadequatb flow and refer to guidance of ECA-1-1 .'Dama§ to ECCS pumps can occur within a matter of minutes if

-operated with inadequateflow: F6r flow blockage in the SI pump discharge, the RHR pumps should have adequate recirculation flow, however the SI pump mini-flow path is isolated. Although this action may stop recirculation

-flow, it is necessary to preservd'the pumps for future use. Loss of emergency sump recirculation procedure provides guiidance for RCS

-cobldown to establish RHR recirculation flow to the core.

. '. . . .o rt. f. w.t Fuel Asse:(bly'Blockage(Bottom) - These strategies can be broken down as follows: .

1. Aligning SI flow to the reactor vessel rather than the cold legs provides an

-alternate flow path to the core.. If fuel assembly bottom nozzles are sufficiently blocked, then flow may be backing up into the cold legs and

.. spilling to the floor rather than making it into the core region. Opening the SI K. vessel injection path and closing the cold leg injection path will provide flow

- directly on the top of the fuel assemblies and may reverse the flow path through the fuel assemblies helping to push debris off the bottom of the fuel

• -.. assembly.nozzles., .'

2. Re-initiating cold leg injection flow after a settling period: Actions taken to

-direct flow to the core outlet may have allowed .debris to settle or flow

. reversal may have' cleared .the bottom nozzles of debris.. Re-establishing the normal SI cold leg flow'path allows'staff to'determine if bottom nozzle debris

- - t -has been removed. . .i.

3.; Expedite post-LOCAcooidoWn arid depressurization.' Post-LOCA cooldown

• and depressurizatio cools' down the RCS,, wkhifpbrrnits RCS depressurization by sequentially stopping the SI p umnps 'tore cooling is supplied by the RHR pumps and/or~charging pumps.when SI pumps are stoqppd. RHR flow i§ directed to th'es'ssel outlatand thtuisprovides a diff6rent flow path 'for core'coling ah'd allows'settling of any debris on the

. ,.... .-. bottom nozzles.: The procedure may:aIso ptaceone:or;b.oth-trains of RHR in a shutdown cooling alignrment.I,.rF RHR can be placed'in shutdown cooling

. ... alignment, then the need forrecirculation is eliminated.,-This eliminates the

- . potential for debris ingestion to further degrade post-LOCA recirculation

• t .~ operation. *,, * .

/ '-- . ; .m -

Page 19 of 25

PRAIRIE ISLAND NUCLEAR GENERATING PLANT PLANT SAFETY PROCEDURE

.1: ad .NUMBER:

LONG TERM CORE COOLING F3-17.2 REV: 0 Attachment 1 Background Information for F3-17.2, Long Term Core Cooling (Continued)'

Table 2 Table 2 provides strategies for use during low head recirculation operation. The most susceptible blockage locations are the RHR heatexchangers and the fuel assembly top nozzles. Indications of and strategies for sump screen blockage are also provided.

Sump screen blockage is not expected to occur due to the significant amount of screen blockage needed to result in any degradation of RHR suction and the size of screen openings. The indications and trends provided are those primary indications expected to show degrading conditions. Other parameters are also being monitored and may provide valuable trend information when making decisions about recirculation performance.

RHR Heat Exchanger Blockage - These strategies can be broken down as follows:

1. Switch to the other train if both trains are available. Since one RHR pump is idle, that pump may not have sustained any significant debris blockage.

Stopping the operating train may allow accumulated debris to settle.

Operation of the train following a period of time for settling could later be attempted if necessary or desired.

2. Stopping one pump, if both pumps are running, may allow debris to settle.

Operation of the secured pump following a period of time for settling could later be attempted if necessary or desired.

3. If only one pump is available then stopping the running pump and restarting the pump may clear the debris by allowing it to settle or pushing it through the opening with the flow/pressure surge when the pump is started. To avoid excessive core heat-up, flow to the core should not be stopped for more than six minutes. (EOP Setpoint calculation J.5 provides the basis for the six minutes maximum flow interruption time).

4. Throttling RHR flow as low a possible will reduce debris ingestion and may result in maintaining adequate core cooling flow while limiting further degradation of flow capability.

5. Open the RHR heat exchanger bypass line. Establishing this flow path will provide sump water to the vessel with limited or no cooling in the RHR heat exchanger. Flow is provided to maintain the core covered and sump liquid temperature would increase to saturation. Decay heat is then removed by the containment fan coil units. Analysis has demonstrated that no unacceptable conditions would re3sult from this scenario. (NMC Internal Correspondence, dated 1/17/03, "Containment Response to LOCA with loss of CC at 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />," also see NAD File#OC.PX.2003.06.

Page 20 of 25

PRAIRIE ISLANDNUCLEAR GENERATING PLANT PLANT SAFETY PROCEDURE NUMBER:

l LONG TERM CORE COOLING F3-17.2 REV: 0 Attachment 1 Background Information for F3-17.2, Long Term Core Cooling (Continued)

• Fuel Assembly Blockage (Top) -JThese strategies can be broken down as follows:

Blockage at the top of the fU'e-.a'sembiie's is'ne 'ofthe most difficult scenarios to

. detect based on existing in'dictions.; DepenldWing on the break location, recirculation system parameters can appeartbenormal, but the flow would not be reaching the core. As top n6zzelplu'gging occurs and'flow to the'core is hindered, RCS hot leg level may rise.': Level could rise such 'that RHR flow spills to the floor rather than

'finding an alternate route to thb core via the cold legs. -Due to thermocouple location,'the'thermocouples may not show elevated temperatures indicative of fuel heat-up. However, if the core is covered and adequate cooling is provided, then the thermocouples are.expected to be indicating a relatively uniform core exit temperature. If the core exit thermocouples start to diverge or'show unexpected indications or trends, then blockage at the fuel assembly top nozzle should be

• -.suspected.. .

1 Aligning for high head recirculation provides an alternate flow path to the core by directing flow through the vessel via the cold legs to the bottom of the fuel assemblies. The'change in flow path should provide core cooling and help to'push the debris frodm-the top of the fuel assemblies.

2. Aligning RHR to the Loop B cold leg return line directs the flow through the vessel via the cold leg. -The alternate flow path may provide core cooling and help to push the debris from the top of the fuel assemblies.'

' SumpB Screen Blockage - These strategies can .be broken down as follows:

i Bockage of the sump B screen is not expebted due to the relative.large size opening. Analysis (ENG-ME-005 Rev.-'3) has'de'morafted that a substantial portion of the screen-would need~to becompletely.biockeq§toadversely affect recirculation.cperatio n. Blockageof thesu'mip.-B screen is e6idenced by loss of adequat6 NPSHto the RHR pump(s) or...erratic RHR pump/motor indications. Since the-sump screens are common to' both trains, both trains are expected to show the

• ,same symptorms-and!alternating trains provides nb benefit.':

-. -'.- -; a Ad 1..i  ; A, , ',: 4 "1,'9  ; I] .';' ' .

• -1. Stopping.one RHR pump -ifboth a're"running reduces the total flow through the.sump.,screens: The:reduction in total flow may be-sufficient to restore

NPSH for the remainingpmp 'One train of RHR on recirculation is also sufficient to provide adequate core cooling flow. ' -

B y .- , , *-. . A. , .*i:X .. . , ,: , . ,

-1 .. -f  ! ......................

Page 21 of 25 -

PRAIRIE ISLAND NUCLEAR GENERATING PLANT PLANT SAFETY PROCEDURE Attachment 1 BackgrQundllnformation for F3-17.2, Long Term Core Cooling I. (Continued)

2. Reducing RHR pump flow reduces the required NPSH and may restore adequate NPSH margin for the pumps and provide sufficient flow for core cooling. Flow reduction below the value needed for stable core temperature may be necessary and in that case vessel inventory should be maintained above the fuel. This flow reduction may be sufficient to maintain RHR pump NPSH and decay heat removal by phase change and subsequent heat removal by the containment fan coil units.

3. If only one pump is available then stopping the running pump and restarting the pump may clear the debris by allowing it to settle. To avoid excessive core heat-up, flow to the core should not be stopped for more than six minutes. (EOP Setpoint calculation J.5 provides the basis for the six minutes maximum flow interruption time).

4. Aligning RHR suction to the RWST provides a source of injection flow to the core and raises containment level to improve RHR pump NPSH. This action is only possible if RWST level has been raised following transfer to recirculation. ECA-1.1 provides guidance for required flow to maintain core cooling. RHR pumps should be stopped if a suction source is not available (RWST not available and NPSH lost regardless of RHR flow).

Step 7.2 Step 7.2 provides guidance to adjust/reduce RHR flow rate during long-term recirculation operation. Full RHR flow rate is not required for long-term recirculation operation. Evaluations have demonstrated that adequate core cooling can be provided with much lower flow rates within the first few hours following the reactor trip. Reducing RHR flow rate during long term recirculation has several benefits:

1. Better control of plant parameters such as RCS temperature and temperature trends, RCS inventory, etc.

2. Reduces the flow velocities at the basement of containment. This increases the likelihood that potential debris sources can settle to the containment floor prior to reaching the containment sump B.

3. Reduces the head loss across the sump B screen, which increases the available NPSH to the RHR pumps.

4. Decreases adverse affects of debris that has been ingested into the recirculation system.

5. May establish conditions that allow aligning for RHR shutdown cooling; which would alleviate the need to draw liquid from the containment sump.

Page 22 of 25

PRAIRIE ISLAND NUCLEAR GENERATING PLANT . PLANT SAFETY PROCEDURE 4,~r *~. ,,' j'< .NUMBER: -

LONG TERM CORE COOLINGs. F3-17.2 REV: 0 Attachment 1 'Backgroundjlnforffiation f'r'F3'72,Long Terirn Core Cooling

(on inued)

Note at beginning of Step 7.2 ., ..

Minimum RHR pump flow rate for continuous operation is 600 gpm.. Specifying a minimum RHR flow rate of 750 gpm is consistent with bther plant procedures (such as reduced inventoryoperration and ensures at continuous operation of the pump is at

-greater than 600 gpm.

Steps 7.2.1.and 7.2.2 .

These steps reduce RHR flow based on the number of RHR trains operating and system conditions. ln lieu of specifying a minimum flow rate vs. time as provided in other procedures,'RHR flow rate is reduced based on maintaining specific system

-conditions such as RCS temperature'and RCS inventory. Stopping one RHR pump and reducing RHR flow are routine activities performed with RHR in a shutdown cooling alignment. As decay heat load reduces, the requirement for RHR flow also reduces.

The actions of throttling flow or stopping one pump account for the reduced decay heat removal requirements. 'Performance of these steps is only performed after both SI pumps are stopped to' ensure an RHR pump supplying SI is not stopped.

Step 7.2.3 Establishment of RHR shutdown cooling is performed based on operating equipment and system conditions. Based on the break elevation, it may be possible to maintain RCS elevation at the-break location with one-train and align the other RHR train for shutdown coling.i If shutdown cooling is established, it maybe subsequently possible, based on break location, to stop recirculation flow and have RCS inventory maintain at the break location.- The combination'of listed conditions and use of ES-I1.1 provide the rnecessary plant conditions and instructions'to-co'mplete the shutdown cooling alignment, if plant staff decide this should be attempted.

• .  ; *;*!., I vbt4SS8

' -tt-a*

-;I 'o Pe: 3, 2`5

-. S. .;A Page 23 of 25'

PRAIRIE ISLAND NUCLEAR GENERATING PLANT PLANT SAFETY PROCEDURE NUMBER:

3

't aF 1 LONG TERM CORE COOLING F3-17.2

. r REV: 0 Attachment 1 BackgrQund 1lnformation for.F3-17.2, Long Term Core Cooling (Continued)

Step 7.3 This step provides guidance for refilling the RWST after recirculation has been initiated.

Refilling the RWST provides an alternate water source to re-establish injection in the event that recirculation flow is degraded or lost. Refilling the RWST also provides a source of water to the charging pumps. Evaluations have shown that the liquid contents equivalent to two. complete RWSTs can be injected into containment without flooding required equipment. In addition to providing an injection water source to maintain core cooling, re-establishing injection can also provide the following benefits:

1. Increases the liquid level inside of containment, which (for a given recirculation flow rate) results in a reduced flow velocity. The reduced flow velocity increases the likelihood that potential debris sources can settle to the containment floor prior to reaching sump B.

2. Increases the liquid level inside of containment, which increases the available NPSH to the RHR pumps.

Note at beginning of Step 7.3 See above discussion.

1st Caution at beginning of Step 7.3 Radiation levels following a LOCA are dependent upon the amount of fuel damage that has occurred. If significant fuel damage has occurred, then refilling the RWST may involve personnel entry into areas with potentially very high radiation fields. Based on analysis the areas near SI-17-1 and 2SI-17-1 are anticipated to be very high radiation fields. Strategies to minimize exposure such as minimizing time, knowing the high radiation field locations through surveys, or the use of multiple people.

2 nd Caution at beginning of Step 7.3 In order to maintain long-term sub-criticality of the core, a minimum boron concentration of the sump liquid is required. To ensure that this minimum boron concentration is maintained, with some margin, a minimum boron concentration of 2000 ppm should be established in the RWST during refill. This minimum boron concentration can be established by knowing the boron concentrations and volumes in the water sources used to refill the RWST; that is, it is not necessary to sample the RWST unless there is uncertainty that the 2000 ppm minimum concentration is satisfied. The upper limit of 3500 ppm is to be consistent with the sump pH evaluation.

Page 24 of 25

AttachmentI Backgroundlnrifforatibon'oiF'-i'.2, Lontg TernM Core Cooling

vt`'(Continued)

Step 7.3 The RWST is refilled using current 'standing prodddures;: The first'three methods (transfer from other unit'RWST, CVCS Hold-Up Taniks or CVCS Monitor Tanks) require

'no actions from the Control Roo6m. The fourth' rmithod (refill using th6 BA Blender) requires control room'operator acti6ns. Consideration'that the control rooM'operators may have other duties at this time should be factored into deciding which method(s) will be used. Transfer of water frorm the other unit RWST may require some justification, or

ILCO Condition entry, for openin'g'cross tie valve's on lines without adequate seismic evaluation.

Step 7.4 '

This step provides guidance for adjusting the pH of the liquid in the containment sump.

Containment spray may have been secured during the injection phase, or may not '

have started for break conditions that did not result in containment pressure reaching 23 psig. To reduce the potential for stress corrosion cracking of stainless steel and to retain Iodine in the sump liquid solution, it is necessary to establish a pH > 7.0.

1st Note at beginning of Step 7.4 See above discussion 2nd Note at beginning of step 7.4 Caustic standpipe level indication < 45% indicates that adequate NaOH has bee n injected to assure a'pH > 7.0 if the 'entire liquid contents of one6RWST have been injected. ,If liquid from a'second RWST are to be injected' (dr has been injected) it will be necessary to -inject additional"NaOH to assure the s'urip -iiqqid pH is maintained greater than 7.0.; '

g*0 .h ,.*.! i ; sr..--

Step 7.4 ,. - -

The combination of caustic standpie'lev'el'and -RW8Tlevel provides NPSH for the containment'spray pumps anda'dedate c~lt'fic f6r`6ntainment s pH umpadjustment.

Initiation of flow is by maiually!abtua ingiontainment spray. The containment spray pump discharge valves open when containment spray is'actuated and are maintained

.operi while -the associated spray pumrp is runiting. -An opti6inis'pr6vided to stop one train of containment spray. This may be desirable to lengthen the timeior RWST inventory depletion fdr the case when the RWST is alsobe'ing'used to 're-supply ECCS

-purnps with an injectidn source. ' ' '

Page 25 of 25'