NRC 2002-0107, Emergency Plan Implementing Procedure Revisions for Point Beach Nuclear Plant, Units 1 & 2

Emergency Plan Implementing Procedure Revisions for Point Beach Nuclear Plant, Units 1 & 2

ML023520514
Person / Time
Site:	Point Beach
Issue date:	12/06/2002
From:	Cayia A Nuclear Management Co
To:	Document Control Desk, Office of Nuclear Security and Incident Response
References
NRC 2002-0107
Download: ML023520514 (166)

Text

Nuclear Management Company, LLC N MC Point Beach Nuclear Plant Committed to NuclearEicellence 6610 Nuclear Road Two Rivers, WI 54241 NRC 2002-0107 December 6, 2002 Document Control Desk U.S. Nuclear Regulatory Commission Mail Station P 1-137 Washington DC 20555 Ladies/Gentlemen:

Dockets 50-266 and 50-301 Emergency Plan Implementing Procedure Revisions Point Beach Nuclear Plant, Units 1 And 2 Enclosed are copies of revised procedures to the Point Beach Nuclear Plant Emergency Plan Implementing Procedures. The revised procedures dated October 25 and Nov 15, 2002, should be filed in your copy of the manual. We apologize for the delay.

Sincerely, Site FAF/kmd Enclosures cc: NRC Resident Inspector (w/o/e)

Incident Response Center, Region III

EPIP INDEX POINT BEACH NUCLEAR PLANT Revision 86 EMERGENCY PLAN IMPLEMENTING PROCEDURES October 25, 2002 INDEX REVISION EFFECTIVE PERIODIC PROCEDURE DATE REVIEW DATE NUMBER PROCEDURE TITLE NUMBER 42 C 10/25/02 08/09/01 EPIP 1.1 Course of Actions ................................................. R 07/12/02 02/06/02 EPIP 1.2 Emergency Classification ...................................... 36 EPIP 1.3 Dose Assessment and Protective Action R 06/24/02 05/03/02 Recommendations ............................................ 29 C 07/12/02 12/14/01 EPIP 1.4 Credible High or Low Security Threat ......................... 1 R 06/25/02 02/06/02 EPIP 2.1 Notifications - ERO, State & Counties, and NRC ..... 25 EPIP 4.1 Technical Support Center (TSC) Activation R 07/26/02 12/14/01 and Evacuation ........................................................ 33 EPIP 4.2 Operations Support Center (OSC) Activation R 12/14/01 12/14/01 and Evacuation .................................................... 15 EPIP 4.3 Emergency Operations Facility (EOF) Activation R 07/26/02 12/14/01 and Evacuation .................................................... 27 EPIP 4.7 Offsite Radiation Protection Facility (OSRPF) R 07/12/02 12/14/01 Activation and Evacuation .................................... 3 R 02/06/02 02/06/02 EPIP 5.1 Personnel Emergency Dose Authorization ............ 14 EPIP 5.2 Radioiodine Blocking and Thyroid Dose R 03/20/01 02/06/02 Accounting ......................................................... 13 EPIP 6.1 Assembly and Accountability, Release and R 07/26/02 12/14/01 Evacuation of Personnel ......................................... 23 R 07/12/02 07/12/02 EPIP 7.3.1 Offsite Radiation Sampling and Survey ................ 22 R 07/12/02 07/12/02 EPIP 7.3.2 Post-Accident Isokinetic Stack Sampling System ....... 7 R 07/12/02 07/12/02 EPIP 7.3.7 Estimating Radioiodine Air Concentrations ........... 5 EPIP 8.4.1 Post-Accident Sampling and Analysis of 16 C 10/25/02 07/12/02 Potentially Bicgh Activity Reactor Coolant ......

EPIP 8.4.2 Post-Accident Sampling of Containment 11 C 09/13/02 07/12/02 A tm osphere ............................................................. 07/12/02 07/12/02 EPIP 8.4.3 Emergency Containment Sump "A" Sampling ..... 7 07/12/02 07/12/02 Emergency Reentry ............................................... 22 R EPIP 10.1 R 07/26/02 01/30/01 EPIP 10.2 Core Damage Estimation ...................................... 19 R 05/19/00 05/19/00 EPIP 10.3 Post Accident Containment Hydrogen Reduction ....... 5 C = Continuous Use R = Reference Use (T - Temporary Change) I = Information Use Page 1 of 2

EPIP INDEX POINT BEACH NUCLEAR PLANT Revision 86 EMERGENCY PLAN IMPLEMENTING PROCEDURES October 25, 2002 INDEX REVISION EFFECTIVE PERIODIC PROCEDURE NUMBER DATE REVIEW DATE NUMBER PROCEDURE TITLE 17 R 09/20/02 07/12/02 EPIP 11.2 Medical Emergency ...................................................

EPIP 12.1 Emergency Event De-Escalation, Termination, or 07/12/02 07/12/02 Recovery Operations ................................................. 8 R 07/12/02 07/12/02 Recovery Implementation .......................................... 14 R EPIP 12.2 C = Continuous Use R = Reference Use (T - Temporary Change) I = Information Use Page 2 of 2

EPIP 1.1 COURSE OF ACTIONS DOCUMENT TYPE: Technical NNSR CLASSIFICATION:

REVISION: 42 EFFECTIVE DATE: October 25, 2002 REVIEWER: N/A APPROVAL AUTHORITY: Department Manager Group Head PROCEDURE OWNER (title):

OWNER GROUP: Emergency Preparedness Controlling Work Document Numbers List pages used for Partial Performance

POINT BEACH NUCLEAR PLANT EPIP 1.1 EMERGENCY PLAN IMPLEMENTING PROCEDURES NNSR Revision 42 COURSE OF ACTIONS October 25, 2002 TABLE OF CONTENTS TITLE PAGE SECTION PU R P O SE ...................................................................................................................... 3 1.0 2.0 PRERE QU ISIT ES ........................................................................................................... 3 PRECAUTIONS AND LIMITATIONS ................................................................... 3 3.0 INITIAL CONDITIONS ................................................................................................ 3 4.0 5.0 PROCEDURE ....................................................................................................... 4 REFEREN C ES ............................................................................................................... 9 6.0 7.0 BA SE S ........................................................................................................................... 9 ATTACHMENT A ANNOUNCEMENT OF CLASSIFIED EVENT AND PROTECTIVE ACTION .................................................................... 10 KATTACHMENT B NOTIFICATION OF PERSONNEL OUTSIDE THE PROTECTED AREA... 12 ATTACHMENT C ACTIVATION OF EMERGENCY RESPONSE DATA SYSTEM (ERDS) .............................................................................. 15 Page 2 of 15 CONTINUOUS USE

POINT BEACH NUCLEAR PLANT EPIP 1.1 NNSR EMERGENCY PLAN IMPLEMENTING PROCEDURES Revision 42 October 25, 2002 COURSE OF ACTIONS 1.0 PURPOSE This procedure provides instructions for Control Room personnel responding to an off-normal event at the Point Beach Nuclear Plant (PBNP).

2.0 PREREQUISITES 2.1 Responsibilities 2.1.1 The Shift Manager (SM) is responsible for this procedure.

2.1.2 The SM is responsible for taking immediate actions to mitigate the consequences of the emergency.

2.1.3 The SM is responsible for implementing Emergency Plan Implementing Procedures (EPIPs) as referenced by this procedure until formally relieved by key personnel in emergency response facilities.

2.1.4 The SM may delegate assignments to qualified personnel as necessary.

2.2 Equipment None 3.0 PRECAUTIONS AND LIMITATIONS None 4.0 INITIAL CONDITIONS An off-normal occurrence with the potential for an emergency classification exists (or has existed) at PBNP.

Page 3 of 15 CONTINUOUS USE

POINT BEACH NUCLEAR PLANT EPIP 1.1 EMERGENCY PLAN IMPLEMENTING PROCEDURES NNSR Revision 42 COURSE OF ACTIONS October 25, 2002 5.0 PROCEDURE Date/Time entered /

Reason entered Initials Time NOTE 1: Steps may be completed out of sequence, as appropriate.

NOTE 2: Steps that do not apply, may be marked "not applicable."

NOTE 3: Steps already formally turned over to Emergency Response Facilities (per Step 5.19) may be marked "not applicable:'

of NOTE 4: Control Room is responsible for all steps prior to transfer responsibility. Text in [ ]'s denotes the Emergency Response Facility responsible for implementation of that step AFTER the formal transfer of responsibilities from the Control Room.

NOTE 5: Classifications are to be made consistent with the goal of 15 minutes once plant parameters reach an Emergency Action Level (EAL). (Ref NEI 99-02) 5.1 Implement procedures, or take other actions as required, to place the affected unit(s) in a safe condition. [CR]

5.2 Verify the on-shift Shift Technical Advisor (STA) is in the Control Room. [CR]

5.3 IF a credible HIGH or LOW security threat, THEN go to EPIP 1.4. (N/A if attack in progress)

[CR primary, EOF support]

5.4 Direct the Security Shift Commander to report to the Control Room. [CR]

5.5 Initiate EPIP 1.2, Emergency Classifications, to determine if an emergency classification is required. [EOF primary, TSC/CR support]_

WARNING If event is security related, then discuss the consequences of to conducting an assembly and accountability with Security prior implementation.

5.6 Complete Attachment A, Announcement of Classified Event and Protective Actions, and make the plant Gai-Tronics announcement. [CR]

Page 4 of 15 CONTINUOUS USE

POINT BEACH NUCLEAR PLANT EPIP 1.1 EMERGENCY PLAN IMPLEMENTING PROCEDURES NNSR Revision 42 COURSE OF ACTIONS October 25, 2002 Initials Time 5.7 IF normal working hours, THEN implement EPIP 1.1, Attachment B, to notify personnel outside the protected area of the event, ELSE implement EPIP 1.1, Attachment B, after notifications in Step 5.8 to notify personnel outside the protected area. [TSC]

NOTE 1: IF the event is classified as a GENERAL EMERGENCY, THEN minimum protective action recommendations are required.

"* Evacuate 0-2 miles, all sectors

"* Evacuate 2-5 miles, 3 or 4 downwind sectors NOTE 2: IF wind speed is less than three (3) mph or lake breeze conditions exist, THEN recommend protective actions for all sectors (360'F) 0-5 miles. Lake breeze conditions exist if the difference between actual wind direction values for inland and near shore meteorological towers is greater than 90'F.

NOTE 3: IF wind direction is on or near (-2°) the sector line, THEN use 4 downwind sectors 5.8 Initiate EPIP 2.1, "Notifications - ERO, State & Counties, and NRC."

[EOF]

NOTE: Assign personnel to make each notification.

5.8.1 Emergency Response Organization (pager activation)

Use EPIP 2.1, Section 5.1. [CR]

5.8.2 State & Counties (within 15 minutes of declaration)

Use EPIP 2.1, Section 5.2 and EPIP 2.1 Attachment B,

- Nuclear Accident Reporting Form. [EOF]

5.8.3 KNPP Control Room.

Use EPIP 2.1, Section 5.3. [EOF]

5.8.4 Nuclear Regulatory Commission (NRC)

(immediately after State & County notifications, NOT to exceed 60 minutes from declaration).

Use EPIP 2.1, Section 5.4. [TSC]

Page 5 of 15 CONTINUOUS USE

EPIP 1.1

-POINT BEACH NUCLEAR PLANT NNSR EMERGENCY PLAN IMPLEMENTING PROCEDURES Revision 42 October 25, 2002 COURSE OF ACTIONS Initials Time 5.9 IF emergency involves plant conditions which suggest a radioactive release is in progress or anticipated, THEN initiate EPIP 1.3, Dose Assessment and Protective Action Recommendations. [EOF]

5.10 IF a backshift or weekend, THEN contact the onshift Radiation Protection Technologist and Radiochemical Technician to report to the Control Room for further instructions in support of the event. [CR] (Ref EPIP 10 1, Emergency Reentry) 5.11 Dispatch and track reentry teams until this responsibility is assumed by the OSC. [OSC]

5.12 IF the event is an Unusual Event and additional staff is desired, THEN call in personnel using the Emergency Response Organization (ERO) Call List, ETD 01. [CR]

5.13 Implement the remaining sections of EPIP 6.1, Assembly and Accountability, Release and Evacuation of Personnel, for the K>following:

5.13.1 Accountability (ref EPIP 6.1, Step 5.2.2). [TSC]

5.13.2 Release of personnel (no radiological impediments) (ref EPIP 6.1, Step 5.3). [TSC]

5.13.3 Evacuation of site to offsite assembly areas (includes radiological monitoring prior to leaving the plant site) (ref EPIP 6.1, Step 5.4). [TSC]

5.14 IF Alert or higher, THEN:

5.14.1 Activate Emergency Response Data System (ERDS) per EPIP 1.1, Attachment C, Activation of Emergency Response Data System (ERDS) (Within 60 minutes of declaration). [CR]

5.14.2 Issue high range dosimetry to Control Room personnel. [CR]

5.14.3 Ensure backshift RPTs and RCTs have high range dosimetry. [CR]

Page 6 of 15 CONTINUOUS USE

POINT BEACH NUCLEAR PLANT EPIP 1.1 EMERGENCY PLAN IMPLEMENTING PROCEDURES NNSR Revision 42 COURSE OF ACTIONS October 25, 2002 Initials Time 5.15 IF the event is a General Emergency AND ALL the following criteria are met, THEN implement expanded PARS of evacuation for 0-5 miles all sectors and 5-10 miles downwind sectors. [EOF] (Ref Step 6 15) 5.15.1 Substantial core damage in progress or projected (>20%)

(> 30,000 R/hr in containment high radiation monitors) 5.15.2 Large fission product inventory in containment (more than GAP) (LOSS criteria for RCS barrier in EPIP 1.2,_

Attachment C, exceeded) 5.15.3 Imminent projected containment failure or release underway (LOSS criteria for containment barrier in EPIP 1.2,'Attachment C, exceeded) 5.16 IF event involves a liquid release to the lake, THEN notify local water utilities per Offsite Agency Call List, ETD 02. [EOF]

5.17 IF TSC and/or EOF are NOT activated, THEN ensure periodic status updates are provided to the State, Counties, and NRC per EPIP 2.1. [CR]

5.18 IF activating the Emergency Response Facilities, THEN provide a turnover briefing to TSC Manager upon arrival in the Control Room. [CR]

5.18.1 Plant status 5.18.2 Notifications status and current EPIP 2.1, Attachment B form 5.18.3 Protected Worker Status 5.18.4 Assembly and accountability status 5.19 Conduct a formal turnover of responsibilities to Emergency Response Facilities as they are activated. [CR]

5.19.1 Technical Support Center (TSC)

a. Plant assessment on classification recommendations per EPIP 1.2 K>b. Onsite protective actions Page 7 of 15 CONTINUOUS USE

POINT BEACH NUCLEAR PLANT EPIP 1.1 EMERGENCY PLAN IMPLEMENTING PROCEDURES NNSR Revision 42 COURSE OF ACTIONS October 25, 2002 Initials Time

c. Onsite radiological assessment

d. NRC Notifications per EPIP 2.1

e. Assembly and Accountability, Release and Evacuation of Personnel per EPIP 6.1 5.19.2 Operations Support Center (OSC)

a. Tracking dispatched reentry teams (repair, fire, search and rescue)

b. Tracking of non-PBNP, We Energies personnel, Nuclear Management Company fleet personnel, or contractor teams that may be doing repairs inside and outside protected area (switchyard, etc.)

c. Medical emergencies per EPIP 11.2 5.19.3 Emergency Operations Facility (EOF)

>a. Classification of emergencies per EPIP 1.2

b. Offsite protective action recommendations per EPIP 1.3

c. State and County notifications per EPIP 2.1

d. Overall management of ERO activities

e. Request for Federal Assistance, if needed

f. Authorize the use of potassium iodide per EPIP 5.2

g. Authorize emergency radiation exposures in excess of 10 CFR 20 requirements per EPIP 5.1

h. Review and approval of news releases.

5.20 Initiate an action request associated with the event and insert a copy in the Operations Notebook (reference Step 6.14) and exit this procedure, returning it to Emergency Preparedness. [CR]

Pag~e 8 of 15 CONTINUOUS USE

POINT BEACH NUCLEAR PLANT EPIP 1.1 EMERGENCY PLAN IMPLEMENTING PROCEDURES NNSR Revision 42 COURSE OF ACTIONS October 25, 2002

6.0 REFERENCES

6.1 EPIP 1.2, Emergency Classifications 6.2 EPIP 1.3, Dose Assessment and Protective Action Recommendations 6.3 EPIP 1.4, Credible High or Low Security Threat 6.4 EPIP 2.1, Notifications - ERO, State & Counties, and NRC 6.5 EPIP 4.1, Attachment E, Activation of Emergency Response Data System (ERDS) 6.6 EPIP 5.1, Personnel Emergency Dose Authorizatiofi 6.7 EPIP 5.2, Radioiodine Blocking and Thyroid Dose Accounting 6.8 EPIP 6.1, Assembly and Accountability, Release and Evacuation of Personnel 6.9 EPIP 10.1, Emergency Reentry 6.10 EPIP 11.2, Medical Emergency 6.11 ETD 01, Point Beach Emergency Response Organization (ERO) Call List 6.12 ETD 02, Offsite Agency Call List 6.13 NEI 99-02, Regulatory Assessment Performance Indicator Guideline 6.14 IR 94-013, NPNPD-94-014, Response to Notice of Violation, October 5, 1994 6.15 NUREG-0150, Volume 1, Revision 4, RTM-96, Response Technical Manual, Figures A-5 and A-6, March 1996 7.0 BASES B-1 10 CFR 50.47(b), Emergency Plans B-2 10 CFR 50.47, Appendix E.IV, Content of Emergency Plans Plans B-3 NUREG-0654, Criteria for Preparation and Evaluation of Radiological Response 1980 and Preparedness in Support of Nuclear Power Plants, Revision 1, November Page 9 of 15 CONTINUOUS USE

POINT BEACH NUCLEAR PLANT EPIP 1.1 EMERGENCY PLAN IMPLEMENTING PROCEDURES NNSR Revision 42 COURSE OF ACTIONS October 25, 2002 ATTACHMENT A ANNOUNCEMENT OF CLASSIFIED EVENT AND PROTECTIVE ACTION SOUND THE FISHERMAN'S ALARM.

SOUND THE EVACUATION ALARM.

MAKE THE FOLLOWING ANNOUNCEMENT.

"ATTENTION ALL PERSONNEL. ATTENTION ALL PERSONNEL.

A (AN) 1.0 El THERE ARE CONDITIONS AT THE PLANT THAT WARRANT

[] UNUSUAL EVENT El ALERT El SITE EMERGENCY 0l GENERAL EMERGENCY 0l TERMINATION OF CLASSIFIED EVENT THESE CONDITIONS ARE (EAL chart classification/condition):

RESPONSE FACILITY.

El ALL ERO PERSONNEL REPORT TO YOUR ASSIGNED EMERGENCY (required at an ALERT or Higher, Optional Earlier)

El THESE CONDITIONS ALSO WARRANT A:

2.0 0l LIMITED PLANT EVACUATION OF THE FOLLOWING AREAS:

THE:

ALL REMAINING PERSONNEL IN THESE AREA(S) REPORT TO El (In RCA) RP STATION El (Outside RCA) NORTH SERVICE BUILDING CAFETERIA 0l (Other)

AND AWAIT FURTHER INSTRUCTIONS."

EMERGENCY, Optional Earlier):

3.0 0l FULL SITE ASSEMBLY AND ACCOUNTABILITY (Required at SITE ALL REMAINING PERSONNEL REPORT TO:

E3 YOUR ASSIGNED ASSEMBLY AREA "El NORTH SERVICE BUILDING CAFETERIA "El ADMIN BUILDING EL 26' OFFICE AREA "El ENGINEERING BUILDING CAFETERIA El TRAINING BUILDING NORTH FOYER El WAREHOUSE #4 El (OTHER) 4.0 El OFFSITE ASSEMBLY AND ACCOUNTABILITY (In lieu of Step 3.0):

ALL REMAINING PERSONNEL REPORT TO:

El TWO CREEKS TOWN HALL El TWO RIVERS NATIONAL GUARD ARMORY 5.0 El EXIT THROUGH El THE SOUTH GATEHOUSE El THE NORTH VEHICLE GATE (Only if the South Gatehouse Unavailable) 6.0 l (IF filled in, THEN announce:) AVOID THE FOLLOWING AREA(S):

REPEAT ALARMS AND ANNOUNCEMENT Return the completed form to Emergency Preparedness or TSC Manager.

Page 10 of 15 CONTINUOUS USE

POINT BEACH NUCLEAR PLANT EPIP 1.1 EMERGENCY PLAN IMPLEMENTING PROCEDURES NNSR Revision 42 COURSE OF ACTIONS October 25, 2002 ATTACHMENT A ANNOUNCEMENT OF CLASSIFIED 2EVENT AND PROTECTIVE ACTION Page of 2 NOTE: Ensure each section that should be included in the announcement has a check in that check box. The individual making the announcement should read each of those sections when making the actual announcement.

1.0 CLASSIFICATION 1.1 Check the appropriate event classification level box.

1.2 Log the EAL chart classification number and condition of the event.

1.3 Emergency Response Facilities must be activated at an Alert or higher classification.

NOTE: Select Step 2.0, Step 3.0, or Step 4.0 based on the event in progress and the appropriate personnel protective actions required.

2.0 LIMITED PLANT EVACUATION 2.1 Log the areas where a limited plant evacuation is required.

if 2.2 Check the appropriate location box(s) where the evacuated personnel should report. Use "other" a different assembly area is needed due to the unavailability of the ones listed.

NOTE: An evacuation of non-essential personnel to offsite assembly areas (Step 4.0) shall be conducted in lieu of Step 3.0 under emergency conditions that may endanger human life and health (i.e., fire, flooding, toxic gases, etc.) and the SM or TSC Manager has determined that non-essential personnel shall be evacuated to offsite assembly areas..

3.0 FULL SITE ASSEMBLY AND ACCOUNTABILITY (Required at Site Emergency, Optional Earlier - Ref EPIP 6.1, Step 4.2) 3.1 Check "Your Assigned Assembly Area" if all locations are available.

3.2 Check the appropriate location box(s) where the personnel should assemble if any normal assembly areas are unavailable. Use "other" if a different assembly area is needed due to the unavailability of the ones listed.

4.0 OFFSITE ASSEMBLY AND ACCOUNTABILITY (In lieu of Step 3.0 - Ref EPIP 6. 1, Step 4.3)

Check the appropriate offsite location box where personnel should assemble.

5.0 GATEHOUSE Check the appropriate gatehouse that personnel should exit through.

6.0 AVOID AREAS Log the hazardous area(s) that should be avoided by personnel assembling.

Pag~eIIofl15 CONTINUOUS USE

POINT BEACH NUCLEAR PLANT EPIP 1.1 EMERGENCY PLAN IMPLEMENTING PROCEDURES NNSR Revision 42 COURSE OF ACTIONS October 25, 2002 ATTACHMENT B NOTIFICATION OF PERSONNEL OUTSIDE THE PROTECTED AREA Page 1 of 3 1.0 PBNP AUTOMATED NOTIFICATION SYSTEM NOTE 1: Listen carefully because the system will give you other options. To expedite the notification process, the following steps have been written to only list the specific voice prompt you need, at which time you can immediately respond without listening to the remaining prompt.

NOTE 2: If at any point you want to exit the system and start over, you should keep slowly pressing the "#" key until the system says "goodbye" and restart the entire process.

1.1 PBNP Automated Notification System is unavailable, THEN go to Step 2.0 of this attachment.

1.2 From any on-site telephone, dial ext. 7158 to access the PBNP Automated Notification System.

1.3 When prompted "Please enter your scenario activation password," enter "111222333#"

using the keypad on the telephone.

1.4 When prompted "To start a scenario enter the scenario ID....," enter "700#."

1.5 Press "2" to select recording a new message. Other voice prompts will be given but you do not have to listen to the options.

1.6 When prompted "After the tone, speak the new message....," read Attachment A, Announcement of Classified Event, and Protective Actions, and press "#." (Your message will automatically play back).

1.7 Press "3" to start the scenario. Other voice prompts will be given if you want to replay your message or rerecord it.

1.8 When prompted, "The scenario is building," press "#", listen to "good-bye" and hang up:

1.9 IF normal working hours and Step 1.0 was successful, THEN return to procedure Step 5.8.

1.10 IF outside normal working hours and Step 1.0 was successful, THEN return to procedure Step 5.9.

1.11 IF Step 1.0 was NOT successful, THEN go to next Step 2.0 of this attachment.

Completed By Date/Time Page 12 of 15 CONTINUOUS USE

POINT BEACH NUCLEAR PLANT EPIP 1.1 EMERGENCY PLAN IMPLEMENTING PROCEDURES NNSR Revision 42 COURSE OF ACTIONS October 25, 2002 ATTACHMENT B NOTIFICATION OF PERSONNEL OUTSIDE THE PROTECTED AREA Page 2 of 3 2.0 PBX Broadcast System 2.1 Notification of Personnel Outside Protected Area on South Side of Plant From a touch-tone phone, dial 2.1.1 7666 to access voice system 2.1.2 5623# when prompted for mailbox 2.1.3 5623# when prompted for password 2.1.4 75 to compose message 2.1.5 8003# and 8004## when prompted 2.1.6 5 when prompted 2.1.7 Read Attachment A, Announcement of Classified Event and Protective Actions, and press # when done recording.

2.1.8 79 to send message 2.1.9 83 to exit PBX Broadcast System.

2.2 Notification of Personnel Outside Protected Area on North Side of Plant From a touch-tone phone, dial 2.2.1 6666 to access voice system 2.2.2 5152# when prompted for mailbox 2.2.3 5152# when prompted for password 2.2.4 75 to compose message 2.2.5 8003## when prompted 2.2.6 5 when prompted Pagye 13 of 15 CONTINUOUS USE

POINT BEACH NUCLEAR PLANT EPIP 1.1 EMERGENCY PLAN IMPLEMENTING PROCEDURES NNSR Revision 42 COURSE OF ACTIONS October 25, 2002 ATTACHMENT B NOTIFICATION OF PERSONNEL OUTSIDE THE PROTECTED AREA Page 3 of 3 2.2.7 Read Attachment A, Announcement of Classified Event and Protective Actions, and press # when done recording.

2.2.8 79 to send message 2.2.9 83 to exit PBX Broadcast System.

2.3 IF normal working hours, THEN return to procedure Step 5.8 2.4 IF outside normal working hours, THEN return to procedure Step 5.9.

Completed By Date/Time Return the completed form to Emergency Preparedness or TSC Manager Page 14 of 15 CONTINUOUS USE

POINT BEACH NUCLEAR PLANT EPIP 1.1 EMERGENCY PLAN IMPLEMENTING PROCEDURES NNSR Revision 42 COURSE OF ACTIONS October 25, 2002 ATTACHMENT C ACTIVATION OF EMERGENCY RESPONSE DATA SYSTEM (ERDS)

NOTE 1: ERDS shall be activated within one hour of declaration of an Alert or higher emergency.

being a NOTE 2: ERDS must be activated and deactivated from the Control Room due to controlled environment and the TSC PPCS configuration.

transmit NOTE 3: ERDS can be activated from terminals PPCS 101 or 102. Only one Unit can

'connected' through 179 then the data at a time through a connection. If one Unit is Units are other Unit to be 'connected' must be activated through 182. When both inadvertently activated through a single line 179(182) then only one unit displays

'connected'. The other unit displays 'on line' with an 'error' message. The NRC receives Both Units alarms indicating validity of the data being transmitted should be checked.

connection. The status can transmit simultaneously by activating one Unit on each displayed would be 'connected' in this case.

1.0 TO INITIATE THE ERDS 1.1 From Control Room drop'PPCS 101 or 102, click on the MENU icon.

1.2 Click on Operator Station Programs.

1.3 Click on ERDS for the appropriate Unit.

1.4 Click on the Startup button.

NOTE: When ERDS is activated due to an Emergency the NRC must provide verbal concurrence before ERDS can be deactivated from PBNP.

2.0 TO DEACTIVATE THE ERDS 2.1 From Control Room drop PPCS 101 or 102, click on the MENU icon.

2.2 Click on Operator Station Programs.

2.3 Click on ERDS for the appropriate Unit.

2.4 Click on the Shutdown button.

Page 15 of 15 CONTINUOUS USE

EPIP 8.4.1 POST-ACCIDENT SAMPLING AND ANALYSIS OF POTENTIALLY HIGH ACTIVITY REACTOR COOLANT DOCUMENT TYPE: Technical CLASSIFICATION: Safety Related REVISION: 16 EFFECTIVE DATE: October 25, 2002 REVIEWER: Qualified Reviewer APPROVAL AUTHORITY: Department Manager Head PROCEDURE OWNER (title): Group OWNER GROUP: Emergency Preparedness Verified Current Copy:

Signature Date Time List pages used for Partial Performance Controlling Work Document Numbers

POINT BEACH NUCLEAR PLANT EPIP 8.4.1 EMERGENCY PLAN IMPLEMENTING PROCEDURES SAFETY RELATED Revision 16 POST-ACCIDENT SAMPLING AND ANALYSIS OF October 25, 2002 POTENTIALLY HIGH ACTIVITY REACTOR COOLANT TABLE OF CONTENTS SECTION TITLE PAGE PU RP O SE ...................................................................................................................... 4 1.0 PRERE Q U ISITES .......................................................................................................... 4 2.0 PRECAUTIONS AND LIMITATION ..................................................................... 4 3.0 INITIAL CONDITIONS .......................................................................................... 5 4.0 6

5.0 PROCEDURE ......................................................................................................

Preliminary Evaluation ............................................................................................... 6 5.1 9

5.2 -Preparation Of Sampling Equipment And Laboratory Setup ....................................

Approvals For Implementing Sampling And Analyses ............................................ 13 5.3 Reactor Coolant Sampling ....................................................................................... 14 5.4 5.5 Collecting Gaseous Sample From Pressurized Sample Vessel ................................ 33 Analyzing Gaseous Sample ...................................................................................... 35 5.6 5.7 Collecting Liquid Sample From-Sample Vessel ..................................................... 37 5.8 Analyzing Liquid Sample ........................................................................................ 38 Submittal Of Sampling/Analyses Documents ......................................................... 40 5.9 Sampling/Analysis Report And Continuation Of Procedure - Chemistry Supervision 40 5.10 5.11 Labeling Samples ..................................................................................................... 40 Sample Storage ........................................................................................................ 40 5.12 System Normal Valve Lineup ................................................................................. 41 5.13 5.14 Re-Installation Of Shielded Sample Vessel ............................................................ 43 Preparation/Transfer Of Sample(S) To Off-Site Laboratory ................................... 46 5.15 Post-Analysis Documentation/Submittal ................................................................ 47 5.16 REFEREN CE S ............................................................................................................. 48 6.0 B A SE S .......................................................................................................................... 48 7 .0 Page 2 of 54 CONTINUOUS USE

POINT BEACH NUCLEAR PLANT EPIP 8.4.1 SAFETY RELATED EMERGENCY PLAN IMPLEMENTING PROCEDURES Revision 16 October 25, 2002 POST-ACCIDENT SAMPLING AND ANALYSIS OF POTENTIALLY HIGH ACTIVITY REACTOR COOLANT TABLE OF CONTENTS PAGE SECTION TITLE ATTACHMENT A PRIMARY COOLANT HIGH-LEVEL SAMPLING FLOW DIAGRAM ...... 49 SETUP FOR TRANSFER OF SAMPLE FROM SAMPLE VESSEL ...... 50 ATTACHMENT B ATTACHMENT C PBNP REACTOR COOLANT POST-ACCIDENT SAMPLING AND AN ALYSIS REPORT ................................................................................... 51 Page 3 of 54 CONTINUOUS USE

POINT BEACH NUCLEAR PLANT EPIP 8.4.1 EMERGENCY PLAN IMPLEMENTING PROCEDURES SAFETY RELATED Revision 16 POST-ACCIDENT SAMPLING AND ANALYSIS OF October 25, 2002 POTENTIALLY HIGH ACTIVITY REACTOR COOLANT Reason for performance: [] Drill E] Training En Fuel Failure Unit Being Sampled: El Unit 1 En Unit 2 1.0 PURPOSE high This procedure provides instructions for collecting, handling, and analyzing a (potentially radioactivity) reactor coolant sample with due to fuel failure.

may This procedure may also be used for training and drills or exercises. Chemistry supervision of unit.

exclude any section or step where performance may adversely affect operation 2.0 PREREQUISITES 2.1 Responsibilities 2.1.1 The Shift Manager (SM) Operations Support Center Coordinator if TSC is sample activated) is responsible for the management of the reactor coolant collection and analysis.

2.1.2 Radiation Protection (RP) Supervision (Rad/Chem Coordinator if TSC is activated) is responsible for the radiological monitoring controls.

2.1.3 Chemistry Supervision (Chemistry Leader is OSC is activated) is responsible for sample collection and analysis.

2.2 Equipment 2.2.1 Emergency Preparedness Radiation Protection Equipment maintained in OSC.

2.2.2 Refer to Step 5.2 for Chemistry equipment.

3.0 PRECAUTIONS AND LIMITATIONS be subjected to 3.1 During performance of radiological sampling activities, personnel may radioactivity.

radiological hazards including high radiation, contamination, and airborne other 3.2 Personnel shall wear prescribed protective clothing, dosimetry devices, and (RWP) or protective equipment, as required by the applicable Radiation Work Permit Re-entry Permit, when performing radiological sampling.

3.3 Improper handling of radioactive material can result in personnel contamination, radioactive material uptake, and unplanned personnel exposure.

Page 4 of 54 CONTINUOUS USE

POINT BEACH NUCLEAR PLANT EPIP 8.4.1 EMERGENCY PLAN IMPLEMENTING PROCEDURES SAFETY RELATED Revision 16 POST-ACCIDENT SAMPLING AND ANALYSIS OF October 25, 2002 POTENTIALLY HIGH ACTIVITY REACTOR COOLANT 3.4 SM approval is required for connecting DI-148 (DI Header To High Radiation Sample shall System) to the DI (deionized) water manifold, and Unit Control Operator (Unit CO) affect be informed prior to initiating DI flush because flush will dilute reactor coolant and reactivity. Unit CO shall also be notified upon completion of flush.

time a 3.5 Reactor Coolant system samples shall be taken within three (3) hours from the decision is made to dispatch a sample team.

NOTE: SC-956C and SC-990 (Step 3.6) may be required for containment isolation should SC-955 or SC-959 (Step 3.7) fail to shut.

being used for 3.6 SM approval is required for opening the following valves when they are containment isolation (i.e., only one other valve operable upstream):

3.6.1 SC-946, HX-14C RC Hot Leg Sample HX DI Flush 3.6.2 SC-956C, HX-14C RC Hot Leg Sample HX Inlet 3.6.3 SC-990, HX-14C RC Hot Leg Sample HX Inlet RHR boundary valves 3.7 Unit CO shall be notified immediately should the following containment fail to SHUT within 15 seconds of moving valve switch to CLOSE position:

3.7.1 SC-955, RC Hot Leg Sample Isolation 3.7.2 SC-959, RHR Loop Sample Isolation drill, or 3.8 Sample vessel shall be re-installed with shielding upon completing training, exercise.

drill, or 3.9 Flush and blow dry laboratory stripping apparatus upon completing training, exercise.

4.0 INITIAL CONDITIONS Indications of fuel failure.

Pag~e 5 of 54 CONTINUOUS USE

POINT BEACH NUCLEAR PLANT EPIP 8.4.1 EMERGENCY PLAN IMPLEMENTING PROCEDURES SAFETY RELATED Revision 16 POST-ACCIDENT SAMPLING AND ANALYSIS OF October 25, 2002 POTENTIALLY HIGH ACTIVITY REACTOR COOLANT INITIALS 5.0 PROCEDURE 5.1 Preliminary Evaluation NOTE: This section shall be completed prior to entering the Auxiliary Building OR Sample Room to obtain Reactor Coolant samples under emergency conditions.

NOTE: Equipment collection and laboratory setup (Section 5.2) may be performed concurrently with this section.

5.1.1 Indications of Possible Fuel Damage NOTE: Fuel damage may be the cause when any of the conditions in Steps 5.11.a - 5.11.h exist.

Circle step number(s) and initial to indicate existing condition(s), and enter N/A for conditions that do NOT exist:

a. Letdown Valve Gallery Monitor (RE-1 16) reading is unusually high OR offscale high. (Containment isolation may invalidate readings.)

b. Post Accident Sample Line Monitor, l(2)RE-109 reading is unusually high OR offscale high. (Containment isolation may invalidate readings.)

c. Containment Air Particulate Monitor, 1(2)RE-21 I and Containment Noble Gas Monitor, 1(2)RE-212 readings are unusually high OR offscale high. (RE-211/212 readings may be invalid if containment is isolated.)

d. Containment Low-Range Monitor, 1(2)RE- 102, and Seal Table Monitor, 1(2)RE-107 readings are unusually high OR offscale high.

e. Auxiliary Building Vent Stack Monitor, RE-214 is indicating a significant increase due to Auxiliary Building airborne activity from the Letdown and Charging Pump areas.

f. Charging Pump Room Low Range Monitor, l(2)RE-104 is indicating increased radioactivity in the Auxiliary Building.

Page 6 of 54 CONTINUOUS USE

POINT BEACH NUCLEAR PLANT EPIP 8.4.1 EMERGENCY PLAN IMPLEMENTING PROCEDURES SAFETY RELATED Revision 16 POST-ACCIDENT SAMPLING AND ANALYSIS OF October 25, 2002 POTENTIALLY HIGH ACTIVITY REACTOR COOLANT INITIALS

g. Sample Rbom Low Range Monitor, 1(2)RE-106 is indicating increased radioactivity in the Sample Room and Background Monitor, 1(2)RE-219B is indicating increased radioactivity outside Sample Room.

h. Chem Lab Low Range Monitor, RE-103 is indicating increased radioactivity in laboratory (prior to sample analysis).

5.1.2 Verification of Component Cooling and Instrument Air NOTE: This section (5.1.2) is NOT required for training.

NOTE: Sample cannot be obtained without component cooling to Sample Room heat exchangers and instrument air for operating air operated sample valves.

Contact SM and confirm following (circle as appropriate):

a. Component Cooling In-Service in affected unit (Yes No)

b. Instrument Air In-Service in affected unit (Yes No) 5.1.3 Pre-Job Briefing and Evaluation of Potential Radiological Hazards for Sampling and Analysis Functions Attend pre-job briefing which includes Chemistry supervision, Radiation Protection (RP) supervision (or designee), and affected Chemistry technicians to evaluate the following, based on Radiation Work Permit(s) OR reentry permit(s), radiological survey/monitor data, Calculation 98-0058 (Post-LOCA Reactor Coolant Sample System Dose Rate), AND Attachments A-B.

Page 7 of 54 CONTINUOUS USE

POINT BEACH NUCLEAR PLANT EPIP 8.4.1 EMERGENCY PLAN IMPLEMENTING PROCEDURES SAFETY RELATED Revision 16 POST-ACCIDENT SAMPLING AND ANALYSIS OF October 25, 2002 POTENTIALLY HIGH ACTIVITY REACTOR COOLANT INITIALS NOTE: RP determines radiological hazards in access and sampling/analysis areas, dosimetry/respiratory requirements, stay times, and dose estimates, based on radiation dose rate (area and contact), estimated time required for each task, and shielding/distance(s) indicated by chemistry.

a. Chemistry supervision:

1. Provide input as requested by RP and Chemistry Technician(s) regarding sampling and analysis functions.

2. Determine which valve/instrument/component position verification(s) OR other function(s) in procedure may be omitted as a means for minimizing radiation exposure, based on knowledge of pre-event lineups, laboratory readiness, historical data, value gained/lost by performing/NO._ performing, etc., i.e.:

(a) Omit section(s) NOT to be performed (e.g., If radiation dose from pre-sample flush may negate any reduction in over-all dose, then section should be omitted).

(b) Omit steps NOT to be performed (e.g.,

sampling equipment Steps 5.2.1.b and 5.2.1.c may already be in location for use, AND Steps 5.4.2.i through 5.4.2.k should already be in desired position.

(c) Choose flow path to be used, i.e., if SC-955/966C OR SC-959/990 used, omit steps/sub-steps for flow path NO.T used.

3. Ensure N/A entered and initialed for any section/steps/data lines omitted AND affected selection boxes are checked.

4 Provide other suggestions for reducing exposure time and completing procedure in timely manner, as applicable.

Page 8 of 54 CONTINUOUS USE

POINT BEACH NUCLEAR PLANT EPIP 8.4.1 EMERGENCY PLAN IMPLEMENTING PROCEDURES SAFETY RELATED Revision 16 POST-ACCIDENT SAMPLING AND ANALYSIS OF October 25, 2002 POTENTIALLY HIGH ACTIVITY REACTOR COOLANT INITIALS

5. Assign sampling/analysis functions among chemistry personnel, as necessary to maintain exposure ALARA.

b. Chemistry Technician(s): Provide input as requested by Chemistry supervision OR RP, AND ask questions, as necessary to ensure all concerns are adequately addressed.

"5.2 Preparation Of Sampling Equipment And Laboratory Setup 5.2.1 Sampling Equipment List

a. Obtain the following equipment and supplies:

1. Two-way radio and spare battery

2. Waste container for collecting possible drips from High Rad Sampling Station (HRSS) sample vessel fittings_

3. Polyethylene bags for trash, and large bag(s) to cover sample vessel during transport

4. Paper towels to control/contain potential leaks or spray during connection/disconnection of sample vessel NOTE: Items in Step 5.2.1.b AND 5.2.1.c are normally maintained for immediate use at locations specified in each step. Chemistry supervision may enter N/A for these steps when items are known to be positioned as indicated.

b. Ensure the following equipment is positioned adjacent to Unit I Sample Room (normal location) AND ready for use:

I. Sample cart designated for transport of sample vessel

2. Wrenches, 5/8" and 11/16", open-end (or adjustable) for connecting and disconnecting sample vessel

3. Remote valve turning tool _

4. Syringe for evacuating liquid from sample vessel fittings_

5. Shielded waste bottle Page 9 of 54 CONTINUOUS USE

POINT BEACH NUCLEAR PLANT EPIP 8.4.1 EMERGENCY PLAN IMPLEMENTING PROCEDURES SAFETY RELATED Revision 16 POST-ACCIDENT SAMPLING AND ANALYSIS OF October 25, 2002 POTENTIALLY HIGH ACTIVITY REACTOR COOLANT INITIALS

c. Ensure the following are positioned for use in the Primary Sample Hood:

1. Equipment detailed in Attachment B, excluding sample vessel

2. Remote handling tools located in cabinet below the hood NOTE: Preparation of Laboratory Equipment can be completed in any logical sequence as deemed applicable by the performers.

5.2.2 Laboratory (Analysis) Equipment

a. Gas Chromatograph (G.C.)

b. Ion Chromatograph (I.C.)

c. Multichannel Analyzer (MCA)

d. pH/mv meters (120v AC or battery powered) and pH probe

e. Magnetic stirrers; (120v AC, battery, or water/air powered),

one 50 ml polyethylene beaker, and a 50 ml beaker

f. Piston burette; manual or auto

g. Automatic pipets; continuously adjustable through 5 ml, or equivalent

h. One, 1.0 liter sidearm flask with correctly sized solid stopper and rubber septum over sidearm

i. Gas Collection Vessel, glass

j. Gas syringe, pressure-lock, or equivalent

k. Beaker, 50 ml

1. Mini-lab analyzer, as required by Chemistry supervision

m. Lead brick barricade (wall) of sufficient size to temporarily shield sample in analytical hood Page 10 of 54 CONTINUOUS USE

POINT BEACH NUCLEAR PLANT EPIP 8.4.1 EMERGENCY PLAN IMPLEMENTING PROCEDURES SAFETY RELATED Revision 16 POST-ACCIDENT SAMPLING AND ANALYSIS OF October 25, 2002 POTENTIALLY HIGH ACTIVITY REACTOR COOLANT INITIALS 5.2.3 Reagents, Standards, and Bottled Gases

a. NaOH, 0.1N for boron (May be obtained from normal boron analysis reagents.)

b. Mannitol for boron analysis

c. Boron standard, appropriate for expected boron concentration

d. Calibration gases for G.C.

e. Carrier and valve gas (if applicable)

NOTE: Preliminary Equipment checks can be completed in any logical sequence as deemed applicable by the performers.

5.2.4 Preliminary Equipment Check and Preparation NOTE: Chemistry supervision selects instruments to use.

(Step 5.2.4.a)

a. Ensure current calibration on following instruments and startup instruments (if necessary), in accordance with applicable procedures, and as directed by Chemistry supervision:

1. G.C.

2. I.C.

3. MCA

b. Ensure daily QA check performed and results-are within limits specified in applicable procedure for instrument selected (Step 5.2.4.b) and analysis parameters listed below:

1. G.C. for hydrogen analysis and hydrogen result

2. I.C. for chloride analysis and chloride result

3. MCA for noble gas and iodine analyses and results Page I11 of 54 CONTINUOUS USE

EPIP 8.4.1 POINT BEACH NUCLEAR PLANT SAFETY RELATED EMERGENCY PLAN IMPLEMENTING PROCEDURES Revision 16 October 25, 2002 POST-ACCIDENT SAMPLING AND ANALYSIS OF POTENTIALLY HIGH ACTIVITY REACTOR COOLANT INITIALS NOTE: In the event of loss of AC power to laboratory, a battery-powered pH meter and a battery, air, OR water-powered magnetic stirrer may be used.

c. Standardize pH meter and ensure mid-point check within +/- 0.1 su.

d. Check and prime piston burette with 0.1N NaOH solution.

e. Perform QA check on boron standard and ensure result within +/- 1 ppm for 10 - 19 ppm standard OR +/- 1% or +/- I ppm (whichever is greater) for 20 - 2000 ppm standard.

f. Organize equipment behind lead brick barricade to allow view of operations with aid of mirror.

g. Set up Gas Collection Vessel, ensuring the following:

1. New septum installed AND Valve 2 OPEN

2. Vacuum line attached to vessel at Valve 1

3. Valve 1 OPEN AND valve 3 SHUT

4. Outlet line to liquid sample vessel capped AND vacuum established

5. Valve I SHUT AND vacuum stable

6. IF vacuum decreases, TH-EN ensure the following:

(a) Valves SHUT (b) Rubber septum replaced (c) Fittings are tight (d) Vacuum re-applied AND Steps 5.2.4.g.3 - 5.2.4.g.5 repeated until vacuum stable

7. Remove valve 3 and from gas collection vessel to facilitate easier connection after sampling (N/A if not performed).

.h. Rinse sample vessel drain tube, receiving beaker, and tips of transfer pipets with DI water to remove any chloride contamination.

Page 12 of 54 CONTINUOUS USE

POINT BEACH NUCLEAR PLANT EPIP 8.4.1 SAFETY RELATED EMERGENCY PLAN IMPLEMENTING PROCEDURES Revision 16 October 25, 2002 POST-ACCIDENT SAMPLING AND ANALYSIS OF POTENTIALLY HIGH ACTIVITY REACTOR COOLANT INITIALS

i. Perform two-way radio check with Control Room AND Operation Support Center (OSC), as applicable.

j. IF DI water flush desired by Chemistry supervision, THEN obtain SM permission for connecting DI water to DI water manifold.

5.3 Approvals For Implementing Sampling And Analyses NOTE: Steps 5.3.1 and 5.3.2 are NOT required for training.

Obtain verbal approval to implement sampling and analyses from:

5.3.1 Shift Manager (SM) (Operations Support Center Coordinator if TSC is activated).

5.3.2 Duty and Call Radiation Protection Supervisor (Rad/Chem Coordinator if TSC is activated)

Page 13 of 54 CONTINUOUS USE

EPIP 8.4.1 POINT BEACH NUCLEAR PLANT SAFETY RELATED EMERGENCY PLAN IMPLEMENTING PROCEDURES Revision 16 October 25, 2002 POST-ACCIDENT SAMPLING AND ANALYSIS OF POTENTIALLY HIGH ACTIVITY REACTOR COOLANT UNII 1AL INIT IALS 5.4 Reactor Coolant Sampling NOTE: The terms, demineralized and deionized (DI) water are both used in CHAMP component names for the same water. The term DI is used in this procedure except when demineralized is used in the CHAMP name.

NOTE: For drills and training, DI water flush and sample dilutions are NOT to be performed unless otherwise directed by Chemistry supervision.

CAUTION Sampling (fuel failure only) shall NOT begin until preliminary evaluation is completed and all approvals for implementation have been received.

CAUTION Performance of a DI water flush will dilute the reactor coolant, and affect reactivity.

,5.4.1 Preparation for Collecting Pressurized Sample

a. Contact Unit CO AND perform following:

1. Inform Unit CO of following, as applicable:

(a) PASS sampling (including installation of shielded sample vessel, as applicable) to begin.

(b) RE-109, Post Accident Sample Line Monitor to be removed from service while sampling (c) SC-955, RC Hot Leg Sample Isolation valve will be SHUT.

2. Have Unit CO SHUT SC- 966C, RC Hot Leg Sample (isolation valve), as applicable.

Page 14 of 54 CONTINUOUS USE

POINT BEACH NUCLEAR PLANT EPIP 8.4.1 EMERGENCY PLAN IMPLEMENTING PROCEDURES SAFETY RELATED Revision 16 POST-ACCIDENT SAMPLING AND ANALYSIS OF October 25, 2002 POTENTIALLY HIGH ACTIVITY REACTOR COOLANT INITIALS

b. Proceed with sample cart (loaded with sampling equipment) to HRSS outside appropriate Sample Room, 26' elevation, Primary Auxiliary Building (PAB).

c. Ensure SC-955 SHUT.

d. IF shielded sample vessel installed, THEN enter N/A for Steps 5.4.l.e AND GO TO Step 5.4.1 .f

e. IF shielded sample vessel NOT installed, THEN perform following:

1. Place a waste container under sample vessel fittings to collect potential drips.

CAUTION System pressure shall be relieved by opening valves in Step 5.4.1.e.2 before loosening sample vessel fittings.

2. OPEN following valves to relieve possible pressure in system.

ID Name Initials SC-939 Z-5E High Rad Sample Vessel Inlet SC-940 Z-5E High Rad Sample Vessel Bypass SC-941 Z-5E High Rad Sample Vessel Outlet

3. Ensure the following valves are SHUT:

Page 15 of 54 CONTINUOUS USE

POINT BEACH NUCLEAR PLANT EPIP 8.4.1 SAFETY RELATED EMERGENCY PLAN IMPLEMENTING PROCEDURES Revision 16 "October 25, 2002 POST-ACCIDENT SAMPLING AND ANALYSIS OF POTENTIALLY HIGH ACTIVITY REACTOR COOLANT INITIALS INITIALS CAUTION Paper towel shall be used to contain potential spray when performing Step 5.4.1.e.4.

4. Disconnect (carefully) sample vessel fittings using care to contain potential spray and prevent possible contamination.

5. Allow any liquid in line to drain into waste container.

6. Install shielded sample vessel on HRSS sample vessel fittings.

f. Perform following:

1. Ensure the sample vessel fittings are tight.

2. Ensure the following valves are SHUT:

ID Name Initials SC-939 Z-5E High Rad Sample Vessel Inlet SC-940 Z-5E High Rad Sample Vessel Bypass SC-941 Z-5E High Rad Sample Vessel Outlet 8A Sample Vessel Valve 8B Sample Vessel Valve 9A Sample Vessel Valve 9B Sample Vessel Valve 9B Page 16 of 54 CONTINUOUS USE

EPIP 8.4.1 POINT BEACH NUCLEAR PLANT SAFETY RELATED EMERGENCY PLAN IMPLEMENTING PROCEDURES Revision 16 October 25, 2002 POST-ACCIDENT SAMPLING AND ANALYSIS OF POTENTIALLY HIGH ACTIVITY REACTOR COOLANT INITIALS NOTE: N/A Section 5.4.2 if NOT performing a presample flush.

5.4.2 Pre-Sampling Flush

a. Connect DI water line quick-connect from valve DI-148 to the DI water manifold, AND ensure the quick-connect is unlocked.

b. Ensure the following valves are SHUT:

ID Name Initials DI-148 DI Header To High Radiation Sample System SC-946 HX-14C RC Hot Leg Sample HX DI Flush SC-947 Z-5E High Rad Sample Vessel DI Flush Inlet SC-948 Z-5E High Rad Sample Vessel DI Flush Outlet NOTE: Next step i§ performed only to determine if pump will energize. Pump is NOT to be left on after performance.

c. IF P-166, Demineralized Water Flush Pump to be used, THEN perform following:

1. Ensure the motor starter switch for P-166, is unlocked.

2. Check that P-166 is energized by momentarily switching pump ON, then OFF.

d. IF P-166 did NOT energize, THEN have Unit CO ensure Breaker No. 4 on Panel 1Y-1 14/2Y-1 13 (U 1/2 Rod Drive Room) CLOSED, then repeat Step 5.4.2.c.

e. IF P-166 still did NOT energize, THEN discontinue effort to start pump. (DI water header pressure to be increased at later step.)

f. IF N/A entered by Chemistry supervision for all Steps 5.4.2.i through 5.4.2.k, 5.4.2.1.

THEN GO TO note/caution preceding Step Page 17 of 54 CONTINUOUS USE

POINT BEACH NUCLEAR PLANT EPIP 8.4.1 SAFETY RELATED EMERGENCY PLAN IMPLEMENTING PROCEDURES Revision 16 October 25, 2002 POST-ACCIDENT SAMPLING AND ANALYSIS OF POTENTIALLY HIGH ACTIVITY REACTOR COOLANT INITIALS

g. IF Chemistry supervision did NOT omit all Sample Room valve position verifications, THEN review Steps 5.4.2.i through 5.4.2.k for familiarization with valves needing verification.

h. Obtain SM approval to open SC-946/SC-956C, as applicable AND enter Sample Room.

NOTE: Sign-offs (below) may be initialed after all positions verified.

i. Ensure valves are positioned as indicated in position column, as follows:

ID Name Position Initials SC-961C HX-14C RC Hot Leg Sample HX Outlet SHUT SC-964C RC Hot Leg Sample Throttle SHUT SC-965C Z-5C RC Hot Leg Sample Vessel Bypass SHUT SC-971 RC Hot Leg Grab Sample SHUT SC-968 Z-5C Sample Pressure Vessel Common Outlet SHUT SC-938A RE-109 Failed Fuel Monitor Inlet SHUT SC-938B RE-109 Failed Fuel Monitor Outlet SHUT SC-938C RE-109 Failed Fuel Monitor Flow Throttle SHUT SC-938 HX-14C RC Hot Leg HX Out To High Rad Sample OPEN System SC-969A FI-903 Flow Indicator Purge Line To T-4 VCT Isolation OPEN SC-956C HX-14C RC Hot Leg Sample HX Inlet OPEN

j. Ensure component cooling flow to Sample Room Heat Exchangers by viewing FI-603, Sample System HX Shell Side Outlet Flow Indicator on Sample Panel.

CAUTION SC-990 shall NOT be used for drill OR training.

k. Ensure SC-990, HX-14C RC Hot Leg Sample HX Inlet RHR Sample OPEN, then exit to low radiation area.

Page IS of 54 CONTINUOUS USE

POINT BEACH NUCLEAR PLANT EPIP 8.4.1 SAFETY RELATED EMERGENCY PLAN IMPLEMENTING PROCEDURES Revision 16 October 25, 2002 POST-ACCIDENT SAMPLING AND ANALYSIS OF POTENTIALLY HIGH ACTIVITY REACTOR COOLANT INITIALS CAUTION Unit CO shall be informed prior to initiating DI flush because this evolution can affect reactivity.

NOTE: Steps 5.4.2.1 - 5.4.2.x required only if DI water flush is performed, as directed by Chemistry supervision.

1. Ensure the following valves OPEN:

ID Initials SC-939 SC-940 SC-941 DI-144 NOTE: SM approval required to open SC-946 following containment isolation if SC-966C fails to shut.

m. OPEN following valves:

Page 19 of 54 CONTINUOUS USE

EPIP 8.4.1 POINT BEACH NUCLEAR PLANT EMERGENCY PLAN IMPLEMENTING PROCEDURES SAFETY RELATED Revision 16 October 25, 2002 POST-ACCIDENT SAMPLING AND ANALYSIS OF POTENTIALLY HIGH ACTIVITY REACTOR COOLANT INITIALS NOTE: To minimize radiation dose during an emergency event, Chemistry supervision may have omitted Step 5.4.2.n

n. IF N/A NOT entered by Chemistry supervision for Steps 5.4.2.n.1 - 5.4.2.n.2, THEN check DI water system lineup for applicable unit, as follows:

1. Unit 1: Ensure DI flow by momentarily OPENING, checking flow, then SHUTTING DI water valve in Primary Sample Hood.

2. Unit 2: Ensure OPEN DI water valve (DI-57) located on east wall of Sample Room.

o. IF P-166 to be used, THEN perform following:

1. Ensure valve DI-142 SHUT.

2. Ensure valve DI-143 OPEN.

3. Turn pump switch ON AND record time:

Time:

4. Exit to low radiation area.

p. IF P-166 would NOT energize at Step 5.4.2.o, THEN perform following:

1. Ensure OPEN DI-142, P-166 Flush Pump Bypass.

2. Ensure SHUT DI-143.

3. Record system pressure as indicated on PI-4145:

psi

4. Exit to low radiation area.

Page 20 of 54 CONTINUOUS USE

POINT BEACH NUCLEAR PLANT EPIP 8.4.1 EMERGENCY PLAN IMPLEMENTING PROCEDURES SAFETY RELATED Revision 16 POST-ACCIDENT SAMPLING AND ANALYSIS OF October 25, 2002 POTENTIALLY HIGH ACTIVITY REACTOR COOLANT INITIALS

5. IF system pressure (Step 5.4.2.p.3) is less than 80 psi, THEN contact Unit CO, AND request that DI water header pressure be maximized and the time pressure maximized be provided.

6. Record time pressure maximized:

Time:

q. Time flush, as follows:

1. IF flushed using P- 166, I THEN wait approximately 2 minutes.

2. IF flushed by maximized DI water header pressure, THEN wait approximately 15 minutes after header pressure maximized.

r. Return to HRSS AND measure radiation levels in Sample Room OR general area, as applicable.

s. Record data below AND exit to low radiation area:

1. Time:

2. Radiation Level: mR/Hr; Location:

3. PI-4145 pressure indication: - psi

t. Notify Chemistry supervision of radiation dose rate, AND wait for instructions on system flush.

u. IF Chemistry OR RP supervision directs discontinuation of flush, THEN N/A Step 5.4.2.v AND GO TO Step 5.4.2.w.

Page 21 of 54 CONTINUOUS USE

EPIP 8.4.1 POINT BEACH NUCLEAR PLANT SAFETY RELATED EMERGENCY PLAN IMPLEMENTING PROCEDURES Revision 16 October 25, 2002 POST-ACCIDENT SAMPLING AND ANALYSIS OF POTENTIALLY HIGH ACTIVITY REACTOR COOLANT IN11IAL INIT2IALS

v. IF Chemistry OR RP supervision directs continuation of flush, THEN repeat Steps 5.4.2.q - 5.4.2.t, AND record data in Table A (below) until Chemistry OR, RP supervision directs discontinuation of flush, as applicable:

Table A Flush # PI-4145 Time mR/hr psig After Flush 2

3 4

w. Secure flush according to following:

1. IF flushed by P-166, THEN turn P-166 OFF AND record time flush STOPPED, AND total flush time:

Total Flush Time: minutes Time:

2. IF flushed by maximized DI water header pressure, THEN SHUT DI-142, AN..D record time flush STOPPED, AND total flush time:

Time: Total Flush Time: minutes

3. SHUT following valves:

x. Notify Unit CO DI water flush completed.

Page 22 of 54 CONTINUOUS USE

POINT BEACH NUCLEAR PLANT EPIP 8.4.1 EMERGENCY PLAN IMPLEMENTING PROCEDURES SAFETY RELATED Revision 16 POST-ACCIDENT SAMPLING AND ANALYSIS OF October 25, 2002 POTENTIALLY HIGH ACTIVITY REACTOR COOLANT INITIALS 5.4.3 Pressurized Sample Collection CAUTION Valve SC-959 shall NOT be opened for drill OR training.

a. Contact Control Room AND perform following:

1. Obtain SM permission to OPEN one of following AND check block to indicate valve to be opened:

D-SC-955, RC Hot Leg Sample Isolation OR L-i SC-959, RHR Loop Sample Isolation NOTE: Cycling of Containment Isolation valves may be necessary to reset isolation signal to allow sampling.

2. Have Unit CO reset containment isolation signal on valve selected to enable sampling, OR N/A if NOT applicable.

NOTE: Valve positions are indicated on switch panels by green AND red lights, as follows:

- Green light (only) lit = SHUT

- Red light (only) lit = OPEN

- Both lights lit = Intermediate Position

b. OPEN isolation valve selected (Step 5.4.3.a.1), using local control switch on switch panel outside Sample Room, AND check block for valve OPENED:

D SC-955 OR F-1 SC-959 Page 23 of 54 CONTINUOUS USE

POINT BEACH NUCLEAR PLANT EPIP 8.4.1 SAFETY RELATED EMERGENCY PLAN IMPLEMENTING PROCEDURES Revision 16 October 25, 2002 POST-ACCIDENT SAMPLING AND ANALYSIS OF POTENTIALLY HIGH ACTIVITY REACTOR COOLANT INITIALS

c. IF valve selected (Step 5.4.3.a.1) will NOT open because containment isolation signal NOT reset, THEN perform following:

1. Have Unit CO ensure the containment isolation signal for SC-955 OR SC-959 (as applicable) has been reset.

2. Ensure SC-955 OR SC-959, as applicable SHUT by placing applicable control switch on panel outside of Sample Room in CLOSE position, AND check block to indicate valve shut.

F SC-955 OR F SC-959

3. Ensure SC-955 OR SC-959, as applicable OPEN by placing applicable control switch in OPEN position, AND check block to indicate valve opened.

F SC-955 OR F SC-959

d. Ensure valve positions as indicated in position column, as follows:

ID Name Position Initials SC-961C HX-14C RC Hot Leg Sample HX Outlet SHUT SC-964C RC Hot Leg Sample Throttle SHUT SC-965C Z-5C Hot Leg Sample Vessel Bypass SHUT SC-971 RC Hot Leg Grab Sample SHUT SC-968 - Z-5C Sample Pressure Vessel Common Outlet SHUT SC-938A RE-109 Failed Fuel Monitor Inlet SHUT SC-938B RE-109 Failed Fuel Monitor Outlet SHUT SC-938C RE409 Failed Fuel Monitor Flow Throttle SHUT SC-938 HX-14C RC Hot Leg Hx out to High Rad Sample OPEN System SC-969A FI-903 Flow Indicator Purge Line to T-4 VCT OPEN Isolation I SC-956C HX-14C RC Hot Leg Sample Hx Inlet OPEN Page 24 of 54 CONTINUOUS USE

EPIP 8.4.1 POINT BEACH NUCLEAR PLANT SAFETY RELATED EMERGENCY PLAN IMPLEMENTING PROCEDURES Revision 16 October 25, 2002 POST-ACCIDENT SAMPLING AND ANALYSIS OF POTENTIALLY HIGH ACTIVITY REACTOR COOLANT INITIALS

e. Ensure following valves OPEN:

ID Name Initials 9B Sample Vessel Valve 9A Sample Vessel Valve 8A Sample Vessel Valve 8B Sample Vessel Valve CAUTION For drill OR training exercises only. SC-941 shall be cracked OPEN slightly AND flow maintained at 0.5 - 1.0 gpm by THROTTLING SC-941 to minimize N-16 gamma.

NOTE: FI-903 (located inside Sample Room) can be used as aid for setting flow rate if dose rates permit.

NOTE: Observe pressure indication PI-995 when throttling SC-941 or opening SC-966C.

f. Ensure OPEN SC-939 AND throttle SC-941.

g. IF sample being taken off Hot Leg (i.e., only SC-955 being used)

THEN have Unit CO OPEN SC-966C.

h. Ensure sample flow by confirming one O__R both of following indications, as desired (n/a if needed):

1. Increase in radiation level at Sample Vessel

2. Pressure increase/decrease indication on PI-995 when SHUTTING/OPENING Valve SC-941

i. Exit to low radiation area.

Page 25 of 54 CONTINUOUS USE

POINT BEACH NUCLEAR PLANT EPIP 8.4.1 EMERGENCY PLAN IMPLEMENTING PROCEDURES SAFETY RELATED Revision f6 POST-ACCIDENT SAMPLING AND ANALYSIS OF October 25, 2002 POTENTIALLY HIGH ACTIVITY REACTOR COOLANT INITIALS NOTE: Next step should be performed immediately before returning to HRSS to obtain sample.

j. Contact Unit CO AND obtain/record following pressure indications, as applicable:

(RCS): - psig OR RHR: - psig VCT _ .psig

k. WHEN sample has recirculated for 15 minutes, THEN return to HRSS to collect sample.

1. IF reading on PI-995 NOT between RCS AND VCT pressure, THEN exit to low radiation area, notify Chemistry supervision, AND wait for instructions.

NOTE: Use the remote valve operating tool only as needed to reduce dose.

m. SHUT following valves IN EXACT ORDER listed.

ID Name Initials 9B Sample Vessel Valve 9A Sample Vessel Valve 8B Sample Vessel Valve 8A Sample Vessel Valve

n. Record sample time, AND check that PI-995 indicates less than 100 psig AND is NOT increasing:

Sample time:

o. IF PI-995 equal to or greater than 100 psig, OR is increasing, THEN notify Chemistry supervision, AND wait for instructions.

Page 26 of 54 CONTINUOUS USE

POINT BEACH NUCLEAR PLANT EPIP 8.4.1 EMERGENCY PLAN IMPLEMENTING PROCEDURES SAFETY RELATED Revision 16 POST-ACCIDENT SAMPLING AND ANALYSIS OF October 25, 2002 POTENTIALLY HIGH ACTIVITY REACTOR COOLANT INITIALS NOTE: Sample vessel is NOT to be disconnected until sample flow is secured and DI flush (if being performed) is completed.

NOTE: Valve positions are indicated on switch panels by green and red lights, as follows:

Green light (only) lit = SHUT Red light (only) lit = OPEN

- Both lights lit = Intermediate Position

p. SHUT isolation valve opened at Step 5.4.3.b OR 5.4.3.c.3, as applicable, using local control switch on switch panel outside Sample Room, AND check block for valve SHUT.

- SC-955 OR

- SC-959

q. IF valve SC-955 OR SC-959 (as applicable) did NOT SHUT (green light on panel lit) within 15 seconds, THEN notify Unit CO.

r. IF sample being taken off Hot Leg (i.e., only SC-955 being used)

THEN have Unit CO SHUT SC-966C.

s. IF DI flush NOT to be performed, THEN perform following:

1. Ensure the following valves SHUT:

ID Initials SC-939 SC-940 SC-941

2. Enter N/A for Steps 5.4.4.a - 5.4.4.p, AND GO TO Step 5.4.4.q.

Page 27 of 54 CONTINUOUS USE

EPIP 8.4.1 POINT BEACH NUCLEAR PLANT SAFETY RELATED EMERGENCY PLAN IMPLEMENTING PROCEDURES Revision 16 October 25, 2002 POST-ACCIDENT SAMPLING AND ANALYSIS OF POTENTIALLY HIGH ACTIVITY REACTOR COOLANT uN I.I IALŽ IINITIALS.

NOTE: Post-sampling flush to be performed only as directed by Chemistry supervision, OR N/A entered for sign-offs when NOT performed.

5.4.4 Post-Sampling Flush

a. Ensure following valves OPEN:

CAUTION Unit CO shall be informed prior to initiating DI flush because this evolution can affect reactivity.

NOTE: SM approval required to open SC-946 following containment isolation if SC-966C fails to shut.

b. Ensure following valves OPEN:

ID Initials DI-140 DI-145 DI-148 SC-947 SC-948

c. IF P-166 to be used, THEN perform following:

1. Ensure valve DI-142 SHUT.

2. Ensure valve DI-143 OPEN.

Page 28 of 54 CONTINUOUS USE

POINT BEACH NUCLEAR PLANT EPIP 8.4.1 EMERGENCY PLAN IMPLEMENTING PROCEDURES SAFETY RELATED Revision 16 POST-ACCIDENT SAMPLING AND ANALYSIS OF October 25, 2002 POTENTIALLY HIGH ACTIVITY REACTOR COOLANT INITIALS

3. Turn pump switch ON AND record time:

Time:

4. Exit to low radiation area.

d. IF P-166 NOT to be used, THEN perform following:

1. Ensure OPEN DI-142, P-166 Flush Pump Bypass.

2. Ensure SHUT DI-143.

3. Record system pressure as indicated on PI-4145:

_ psi

4. Exit to low radiation area.

5. IF system pressure (Step 5.4.4.d.3) is less than 80 psi, THEN contact Unit CO, AN.DD request that DI water header pressure be maximized and the time pressure maximized be provided.

6. Record time pressure maximized:

Time:

e. Time flush, as follows:

1. IF flushed by P-166, THEN wait approximately 2 minutes.

2. IF flushed by maximized DI water header pressure, THEN wait approximately 15 minutes after header pressure maximized.

f. Return to HRSS AND measure radiation levels AND DI water pressure at Sample Vessel.

Pace 29 of 54 CONTINUOUS USE

POINT BEACH NUCLEAR PLANT EPIP 8.4.1 SAFETY RELATED EMERGENCY PLAN IMPLEMENTING PROCEDURES Revision 16 October 25, 2002 POST-ACCIDENT SAMPLING AND ANALYSIS OF POTENTIALLY HIGH ACTIVITY REACTOR COOLANT INITIALS

g. Record data below AND exit to low radiation area:

1. time:

2. Radiation Level: mR/Hr

3. PI-4145 pressure indication: - psi

h. Notify Chemistry supervision of radiation dose rate, AND wait for instructions on system flush.

i. IF Chemistry OR RP supervision directs discontinuation of flush, Step 5.4.4.k.

THEN N/A Step 5.4.4.j AND GO TO

j. IF Chemistry OR RP supervision directs continuation of flush, THEN repeat Steps 5.4.4.e - 5.4.4.h AND record data in Table B (below) until Chemistry OR RP supervision directs discontinuation of flush:

Table B Flush # PI-4145 Time mR/hr I psig m' After Flush 2

3 4

k. WHEN directed by Chemistry OR3 RP supervision, THEN STOP flush according to following:

1. OPEN valve SC-946.

2. SHUT following valves:

3. OPEN valve SC-939, using remote valve tool.

Page 30 of 54 CONTINUOUS USE

EPIP 8.4.1 POINT BEACH NUCLEAR PLANT SAFETY RELATED EMERGENCY PLAN IMPLEMENTING PROCEDURES Revision 16 October 25, 2002 POST-ACCIDENT SAMPLING AND ANALYSIS OF POTENTIALLY HIGH ACTIVITY REACTOR COOLANT INITIALS

1. Exit to low radiation area AND wait 5 minutes.

I. IF flushed by P-166, THEN turn P-166 OFF, AND record time flush STOPPED:

Time:

2. IF flushed by maximized DI water header pressure, THEN SHUT DI-142 AND record time flush STOPPED AND total flush time:

Time: Total Flush Time:

m. SHUT following valves, using remote valve tool:

n. SHUT following valves:

ID Initials DI-148 DI-140 DI-143 DI-145 SC-947 SC-948

o. Notify Unit CO DI water flush completed.

p. Disconnect quick-connect between Valve DI-148 AND the DI water manifold, using a paper towel to prevent spraying.

Page 31 of 54 CONTINUOUS USE

POINT BEACH NUCLEAR PLANT EPIP 8.4.1 EMERGENCY PLAN IMPLEMENTING PROCEDURES SAFETY RELATED Revision 16 POST-ACCIDENT SAMPLING AND ANALYSIS OF October 25, 2002 POTENTIALLY HIGH ACTIVITY REACTOR COOLANT INITIALS NOTE: The bottom fitting (Step 5.4.4.q) to be loosened is the fitting closest to the tee.

NOTE: Disconnect the bottom fitting first and then the top fitting.

q. Disconnect sample vessel by loosening the top AN____D bottom fittings using appropriate wrench (5/8", 11/16", or adjustable) AND a paper towel to prevent spraying.

r. Ensure sample cart attached to sample vessel.

s. Remove shielded sample vessel (carefully) from support while ensuring vessel fittings are directly above catch basin.

NOTE: Any water remaining in fittings will dilute sample.

CAUTION Approximately 4-5 mls of liquid will drip out when the syringe needle is inserted into the bottom of the vessel.

t. Remove liquid from top AND bottom vessel fittings with syringe (with long needle) AND EJECT liquid into shielded waste bottle.

u. Replace Swagelok plugs on Sample Vessel (finger tight).

v. Place large polyethylene bag over vessel for transport, AND back sample cart away from sample area.

w. Replace Swagelok plugs on wall fittings (for Sample Vessel), using appropriate tool.

x. IF directed by Chemistry supervision, THEN GO TO Section 5.13, restore system to normal valve lineup, AND RETURN TO Step 5.4.4.

Page 32 of 54 CONTINUOUS USE

POINT BEACH NUCLEAR PLANT EPIP 8.4.1 EMERGENCY PLAN IMPLEMENTING PROCEDURES SAFETY RELATED Revision 16 POST-ACCIDENT SAMPLING AND ANALYSIS OF October 25, 2002 POTENTIALLY HIGH ACTIVITY REACTOR COOLANT INITIALS

y. Transport vessel, remote valve tool, AND wrenches to Chemistry laboratory on sample cart.

z. Record any unusual conditions observed OR conditions that interfered with sampling, including step number(s)

AND affected component ID in comments block, below, AND notify Chemistry supervision of unusual condition(s):

COMMENTS (Sampling) aa. Document completion of sampling on PBNP Reactor Coolant Post-Accident Sampling and Analysis Report, Attachment C, Section 3.0.

5.5 Collecting Gaseous Sample From Pressurized Sample Vessel NOTE: A diagram of setup for collecting sample is provided on Attachment B.

5.5.1 If necessary, attach valve 3 and tubing to shielded sample vessel. (N/A if not performed).

5.5.2 Place shielded sample vessel in sample holder in Primary Sample Hood.

5.5.3 Obtain contact, radiation reading on sample vessel AND record reading for future reference in determining giCi content:

Dose Rate: /Hr NOTE: Sample vessel fittings should be cleaned, but only when radiation levels permit. Thread compounds are NOT to be used anytime.

5.5.4 Ensure sample vessel AND shielded gas collection vessel are adequately shielded.

Page 33 of 54 CONTINUOUS USE

POINT BEACH NUCLEAR PLANT EPIP 8.4.1 EMERGENCY PLAN IMPLEMENTING PROCEDURES SAFETY RELATED Revision 16 POST-ACCIDENT SAMPLING AND ANALYSIS OF October 25, 2002 POTENTIALLY HIGH ACTIVITY REACTOR COOLANT INITIALS NOTE: If vacuum is still established in collection vessel, it will be lost when connecting the sample vessel in the following step.

Vacuum will be reestablished by the procedure.

5.5.5 Connect sample vessel to the shielded gas collection vessel by means of the fittings provided.

5.5.6 Connect drain tube to opposite end of sample vessel, AND ensure Valve 11 on drain tube is OPEN.

5.5.7 Ensure vacuum line is attached to gas collection vessel at valve I location.

5.5.8 Ensure Gas Collection Vessel valves (Attachment B), as follows:

a. Valves 1 AND 2 OPEN.

b. Valve 3 SHUT.

K>t 5.5.9 Evacuate Gas Collection Vessel and connecting lines.

5.5.10 SHUT valve 1 AND check stable vacuum.

5.5.11 IF vacuum unstable, THEN perform following:

a. REPAIR leak(s).

b. OPEN valve 1, AND repeat Steps 5.5.9 AND 5.5.10 until stable.

5.5.12 Secure vacuum AND disconnect vacuum line.

NOTE: Other methods of pressure adjustment (Step 5.5.13) may be used, as directed by Chemistry supervision.

5.5.13 Adjust vacuum to 12-14" Hg, using valve 1.

NOTE: Use the remote valve tool only as needed to reduce dose.

', y 5.5.14 OPEN valves 9A AND 9B.

Page 34 of 54 CONTINUOUS USE

EPIP 8.4.1 POINT BEACH NUCLEAR PLANT SAFETY RELATED EMERGENCY PLAN IMPLEMENTING PROCEDURES Revision 16 October 25, 2002 POST-ACCIDENT SAMPLING AND ANALYSIS OF POTENTIALLY HIGH ACTIVITY REACTOR COOLANT UNi1IAL IN ITIALSb 5.5.15 OPEN valve 8B one-quarter turn.

NOTE: Vacuum drop should be no more than 5 inches Hg/min.

5.5.16 OPEN (crack slightly) valve 8A AND control rate of degassing by throttling valve 8A.

5.5.17 Allow system to degas for 2-4 minutes.

5.5.18 Ensure that the following sample vessel valves are OPEN.

ID Initials 9A 9B 8A 8B 5.5.19 CLOSE valve 2.

5.6 Analyzing Gaseous Sample NOTE: Steps 5.6.1 and 5.6.2 can be done concurrently qOi in opposite order.

5.6.1 Hydrogen

a. Draw an appropriate size sample (typically lcc) using a gas tight syringe AND inject sample through injection port of instrument designated by Chemistry supervision, AND perform analysis, in accordance with procedure for I

designated instrument.

b. Record percent (%) Hydrogen from instrument output in space provided on Attachment C, Table 1, Line A.

c. Record analysis data for Lines B, C, AND D in Attachment C, Table 1.

d. Calculate cc/Kg hydrogen in accordance with equation provided in of Attachment C, Section 1.0, AND record result in space provided on Table 1.

Page 35 of 54 CONTINUOUS USE

POINT BEACH NUCLEAR PLANT EPIP 8.4.1 EMERGENCY PLAN IMPLEMENTING PROCEDURES SAFETY RELATED Revision 16 POST-ACCIDENT SAMPLING AND ANALYSIS OF October 25, 2002 POTENTIALLY HIGH ACTIVITY REACTOR COOLANT INITIALS 5.6.2 Radioactive Noble Gas

a. Draw 0.5 cc sample into gas tight syringe (lor 2 cc'for drill/training).

NOTE: Dilution should be performed if contact reading on syringe is >5 mRthour OR >10% detector spectroscopy amplifier dead time.

b. IF syringe contact radiation reading is >5 mR/hr OR >10% detector dead time, THEN dilute as follows:

1. Inject sample into side-arm flask prepared at Step 5.2.2.h, AND allow 15 minutes for thermal mixing.

2. Draw 1 cc of sample of dilution into syringe.

3. IF contact reading is still >5 mR/hr OR >10% detector dead time, THEN perform following:

(a) Repeat Step 5.6.2.b until reading is 55 mR/hr OR 510% detector dead time.

(b) Record number of dilutions performed below, AND label each side-arm flask as second, third, dilution, etc.:

Dilutions performed for this analysis:

c. IF syringe contact reading (original sample or dilution) is

• _5 mR/hr OR *10% detector dead time, THEN prepare sample (normally) for analysis, in accordance with procedure for instrument to be used.

d. Perform MCA analysis in accordance with procedure for instrument designated by Chemistry supervision, AND obtain MCA printout.

Page 36 of 54 CONTINUOUS USE

POINT BEACH NUCLEAR PLANT EPIP 8.4.1 SAFETY RELATED EMERGENCY PLAN IMPLEMENTING PROCEDURES Revision 16 October 25, 2002 POST-ACCIDENT SAMPLING AND ANALYSIS OF POTENTIALLY HIGH ACTIVITY REACTOR COOLANT INITIALS

e. Perform MCA analysis in accordance with procedure for instrument designated by Chemistry supervision.

f. Record analysis data AND, document noble gas results on Attachment C.

5.7 Collecting Liquid Sample From Sample Vessel 5.7.1 Ensure valve 2 SHUT.

5.7.2 OPEN (carefully), valve 3.

5.7.3 Allow liquid sample to drain into 50 ml beaker.

5.7.4 IF necessary to recover total liquid sample, THEN direct a slow stream of air through vent line on valve 3.

5.7.5 WHEN all sample collected, THEN SHUT following:

ID Initials 8A 8B 9A 9B 3 _____

Pag~e 37 of 54 CONTINUOUS USE

POINT BEACH NUCLEAR PLANT EPIP 8.4.1 EMERGENCY PLAN IMPLEMENTING PROCEDURES SAFETY RELATED Revision 16 POST-ACCIDENT SAMPLING AND ANALYSIS OF October 25, 2002 POTENTIALLY HIGH ACTIVITY REACTOR COOLANT INITIALS 5.8 Analvzing Liquid Sample 5.8.1 pH and Boron Analysis

a. pH Measurement

1. Transfer approximately 2-5 mls of sample into an appropriate beaker for boron analysis.

2. Measure pH AN.D record pH in space provided in, Attachment C, Table 3.

b. Boron Analysis Perform boron analysis in accordance with applicable plant procedure; e.g. CAMP 205, AND record- appropriate analysis parameters/results in Attachment C, Table 3.

5.8.2 Chloride Analysis

a. IF PBNP I.C. OR laboratory is NOT available, THEN perform following:

I. Obtain a suitable aliquot of sample (as directed by Chemistry supervision) for analysis by WE Laboratory Services (LS) using WE Laboratory Services I.C.

2. Store sample, as directed by Chemistry supervision in laboratory vent hood until other analyses are completed.

3. Contact TSC Manager to arrange for transport of necessary WE Laboratory Services personnel AND equipment to PBNP to perform Chloride analysis.

b. IF PBNP I.C. analyzer OR laboratory is available, THEN obtain a suitable aliquot of sample (as directed by Chemistry supervision) AND analyze for Chloride.

Page 38 of 54 CONTINUOUS USE

POINT BEACH NUCLEAR PLANT EPIP 8.4.1 SAFETY RELATED EMERGENCY PLAN IMPLEMENTING PROCEDURES Revision 16 October 25, 2002 POST-ACCIDENT SAMPLING AND ANALYSIS OF POTENTIALLY HIGH ACTIVITY REACTOR COOLANT INITIALS

c. Record ppm chloride in space provided in Attachment C, Table 4, 5.8.3 Iodine Analysis and Gamma Scan NOTE: More than 0.3 cc may be used for drill/training.

a. Transfer 0.3 cc (or less, based on activity) of sample from the beaker to appropriate container for iodine and gamma scan analysis.

b. IF the contact radiation reading is >5 mR/hr OR >10%

detector dead time, THEN dilute the sample as necessary to achieve a contact radiation reading <mR/hr OR detector dead time <10%.

c. IF reading (original sample OR dilution) is <5 mR/hr OR *10% detector dead time, THEN count sample in accordance with procedure for instrument designated by Chemistry supervision.

d. Record necessary data and analysis results on Attachment C, Table 5.

5.9 Submittal Of Sampling/Analyses Documents Remove (temporarily) Sections 1.0 - 5.10 AND submit to Chemistry supervision.

Page 39 of 54 CONTINUOUS USE

POINT BEACH NUCLEAR PLANT EPIP 8.4.1 EMERGENCY PLAN IMPLEMENTING PROCEDURES SAFETY RELATED Revision 16 POST-ACCIDENT SAMPLING AND ANALYSIS OF October 25, 2002 POTENTIALLY HIGH ACTIVITY REACTOR COOLANT INITIALS 5.10 Sampling/Analysis Report And Continuation Of Procedure - Chemistry Supervision 5.10.1 Review/sign/date completed procedure and Sampling and Analysis Report (Attachment C), as applicable.

5.10.2 Forward completed sections of procedure (Sections 1.0- 5.10)

AND Attachments A-C to the Rad/Chem Coordinator.

5.10.3 Direct completion of Sections 5.11- 5.16 of procedure, as appropriate.

NOTE: Sections 5.11- 5.16 may be performed in any order, as directed by Chemistry supervision.

5.11 Labeling Samples Ensure chloride, noble gas, iodine AND gamma scan samples, including sample to be sent to Off-Site analysis laboratory are labeled with following information (as minimum):

5.11.1 Sample number 5.11.2 Name of sample 5.11.3 Date and time of sampling 5.11.4 Sample volume 5.11.5 Dilution 5.11.6 Technician initials 5.11.7 Contact radiation dose rate 5.12 Sample Storage WHEN analysis completed AND samples have been properly labeled, THEN store all samples (liquid and gas) NOT to be transported to Off-Site Laboratory in the designated high level shielded storage area located in the Auxiliary Building area outside of Gas Decay Tank Room.

Page 40 of 54 CONTINUOUS USE

POINT BEACH NUCLEAR PLANT EPIP 8.4.1 SAFETY RELATED EMERGENCY PLAN IMPLEMENTING PROCEDURES Revision 16 October 25, 2002 POST-ACCIDENT SAMPLING AND ANALYSIS OF POTENTIALLY HIGH ACTIVITY REACTOR COOLANT IINiI IAI.Ži IN 1ITIALS, 5.13 System Normal Valve Lineup NOTE: This section is to be performed during training 2R drill exercise, OR when directed by Chemistry supervision.

NOTE: Sign-offs (below) (5.13.1/5.13.2) may be initialed after all positions verified.

5.13.1 Ensure valves are positioned as indicated in position column, as follows:

ID Name Position Initials SC-961C HX-14C RC Hot Leg Sample HX Outlet SHUT SC-964C RC Hot Leg Sample Throttle SHUT SC-965C Z-5C RC Hot Leg Sample Vessel Bypass SHUT SC-971 RC Hot Leg Grab Sample SHUT SC-938B RE-109 Failed Fuel Monitor Outlet SHUT SC-939 Z-5E High Rad Sample Vessel Inlet SHUT SC-940 Z-5E High Rad Sample Vessel Bypass SHUT SC-941 Z-5E High Rad Sample Vessel Outlet SHUT NOTE: Step 5.13.2 can be omitted for training if no flushing evolutions performed. N/A in initials column.

5.13.2 Ensure DI water flush valves positioned as indicated in position column, as follows:

5.13.3 IF DI water header pressure has been maximized OR is still maximized, THEN have Unit CO return DI water header pressure to normal.

Page 41 of 54 CONTINUOUS USE

POINT BEACH NUCLEAR PLANT EPIP 8.4.1 EMERGENCY PLAN IMPLEMENTING PROCEDURES SAFETY RELATED Revision 16 POST-ACCIDENT SAMPLING AND ANALYSIS OF October 25, 2002 POTENTIALLY HIGH ACTIVITY REACTOR COOLANT INITIALS 5.13.4 IF DI flush was performed, THEN RED-LOCK P-166 switch AND DI water jumper to following, as applicable:

a. Unit 1: 1RE-219 SG Blowdown Liquid Monitor Demineralized Water Inlet" pipe

b. Unit 2: HX-14C RC Hot Leg Sample HX Demineralized Water Flush" pipe NOTE: Sign-offs (below) may be initialed after all positions verified.

5.13.5 IF RE-109 is to be put in service, THEN ensure valves are positioned as indicated in position column, as follows:

NOTE: Valves SC-955 AND SC-956C require SM approval to open following containment isolation AND SC-966C fails to shut.

ID Name Position Initials SC-938A RE-109 Failed Fuel Monitor Inlet OPEN SC-956C HX-14C RC Hot Leg Sample HX Inlet OPEN SC-968 Z-5C Sample Pressure Vessel Common Outlet OPEN SC-969A FI-903 Flow Indicator Purge Line To T-4 VCT Isolation OPEN 5.13.6 OPEN valve for SC-955, RC Hot Leg Sample Isolation OR SC-959, RHR Loop Sample Isolation, as applicable, using valve switch on switch panel outside Sample Room.

5.13.7 Have Unit CO OPEN SC-966C, RC Hot Leg Sample.

5.13.8 Establish 0.2 to 0.4 gpm flow through RE-109, Post Accident Sample Line Monitor by throttling SC-938C.

5.13.9 Notify Unit CO RE-109 back In-Service.

Page 42 of 54 CONTINUOUS USE

EPIP 8.4.1 POINT BEACH NUCLEAR PLANT SAFETY RELATED EMERGENCY PLAN IMPLEMENTING PROCEDURES Revision 16 October 25, 2002 POST-ACCIDENT SAMPLING AND ANALYSIS OF POTENTIALLY HIGH ACTIVITY REACTOR COOLANT INITIALS 5.13.10 Record any unusual conditions observed OR conditions that

-interfered with restoring system to normal valve lineup, including step number(s) AND affected component ID in comments block, below, AND notify Chemistry supervision of unusual condition(s):

COMMENTS (Normal Valve Lineup) 5.13.11 Sign AND date in Section 5.16 for restoring system to normal valve lineup.

NOTE: Section 5.16 to be performed during drill OR training, OR as directed by Chemistry supervision during post-accident conditions.

5.14 Re-Installation Of Shielded Sample Vessel NOTE: The RP survey and release (Steps 5.14.1 and 5.14.2) may be omitted for drill OR training.

5.14.1 Have RP perform radiological survey of sample vessel AND associated tools.

5.14.2 WHEN sample vessel AND tools have been released by RP, THEN perform following:

a. Notify Unit CO that RE-109 will be temporarily removed from service for installation of sample vessel.

b. Transport sample vessel AND tools to affected unit HRSS.

Page 43 of 54 CONTINUOUS USE

POINT BEACH NUCLEAR PLANT EPIP 8.4.1 EMERGENCY PLAN IMPLEMENTING PROCEDURES SAFETY RELATED Revision 16 POST-ACCIDENT SAMPLING AND ANALYSIS OF October 25, 2002 POTENTIALLY HIGH ACTIVITY REACTOR COOLANT INITIALS NOTE: Valve positions are indicated on switch panels by green AND red lights, as follows:

- Green light (only) lit = SHUT

- Red light (only) lit = OPEN

- Both lights lit = Intermediate Position 5.14.3 SHUT SC-955 AND ensure SC-959 SHUT.

5.14.4 IF SC-955 (OR SC-959) did NOT SHUT (green light on panel lit) within 15 seconds, THEN notify Unit CO.

5.14.5 Place a waste container under sample vessel fittings to collect possible liquid when sample vessel fittings are removed.

CAUTION System pressure shall be relieved by opening valves in Step 5.14.6 before removing sample vessel fittings.

5.14.6 OPEN following valves to relieve possible pressure in system.

ID Name Initials SC-939 Rad Sample Vessel Inlet SC-940 Z-5E High Rad Sample Vessel Bypass SC-941 Z-5E High Rad Sample Vessel Outlet 5.14.7 SHUT following:

ID Initials SC-939 SC-940 SC-941 Page 44 of 54 CONTINUOUS USE

EPIP 8.4.1 POINT BEACH NUCLEAR PLANT SAFETY RELATED EMERGENCY PLAN IMPLEMENTING PROCEDURES Revision 16 October 25, 2002 POST-ACCIDENT SAMPLING AND ANALYSIS OF POTENTIALLY HIGH ACTIVITY REACTOR COOLANT INITIALS CAUTION Paper towel shall be used to contain potential spray when performing Step 5.14.8.

5.14.8 Remove (carefully) Swaglok plugs from HRSS sample vessel inlet AND outlet lines, using care to contain potential spray AND prevent, contamination.

5.14.9 Allow any liquid in line to drain into waste container.

5.14.10 Remove plugs from sample vessel, AND install sample vessel on HRSS fittings.

5.14.11 Ensure sample vessel fittings are tight.

5.14.12 Ensure following valves are SHUT:

ID Name Initials SC-939 Z-5E High Rad Sample Vessel Inlet SC-940 Z-5E High Rad Sample Vessel Bypass SC-941 Z-5E High Rad Sample Vessel Outlet 8A Sample Vessel Valve 8B Sample Vessel Valve 9A Sample Vessel Valve 9B Sample Vessel Valve 5.14.13 OPEN SC-955.

5.14.14 IF SC-955 did NOT OPEN (red light on panel lit) within 15 seconds, THEN notify Unit CO.

5.14.15 Notify Unit CO that RE-109 back In-Service.

5.14.16 Dispose of any liquid collected in waste container, as directed by Chemistry supervision.

Page 45 of 54 CONTINUOUS USE

POINT BEACH NUCLEAR PLANT EPIP 8.4.1 EMERGENCY PLAN IMPLEMENTING PROCEDURES SAFETY RELATED Revision 16 POST-ACCIDENT SAMPLING AND ANALYSIS OF October 25, 2002 K POTENTIALLY HIGH ACTIVITY REACTOR COOLANT INITIALS 5.14.17 Return sample cart AND tools to original storage location AND configuration, as directed by Chemistry supervision.

5.14.18 Flush and blow dry laboratory stripping apparatus upon completing training, drill or exercise.

COMMENTS (Sample Vessel Installation) 5.14.19 Sign AND date in Section 5.16 for sample vessel installation.

5.15 Preparation/Transfer Of Sample(S) To Off-Site Laboratory NOTE: The Kewaunee Nuclear Plant, Chemistry laboratory is the Off-Site Laboratory that will perform analyses of the K> PBNP Post-Accident sample(s). Refer to Post-accident counting agreement with Wisconsin Public Service, Kewaunee Nuclear Plant.

NOTE: Kewaunee Nuclear Plant does NOT utilize the gas serum vial and 1 cc test tube geometries. Therefore, "normal" sample has to be diluted AND placed in one liter poly bottle for transport to Kewaunee Nuclear Plant.

5.15.1 IF 1-liter sample used for Iodine analysis AND Gamma scan is to be transferred to Kewaunee Nuclear Plant, THEN enter N/A for Step 5.15.2 AND GO TO Step 5.15.3.

5.15.2 IF separate dilution to be prepared for Kewaunee Nuclear Plant, THEN prepare dilution, as directed by Chemistry supervision, AND label in accordance with Section 5.11.

5.15.3 Contact Radioactive Material Shipping personnel to ensure completion of any required shipping papers.

5.15.4 Complete documents, AND submit documents, sample, AND copies of all analysis data obtained for sample being shipped to Kewaunee to Radioactive Material Shipping personnel for transport to Kewaunee Nuclear Plant.

Page 46 of 54 CONTINUOUS USE

EPIP 8.4.1 POINT BEACH NUCLEAR PLANT SAFETY RELATED EMERGENCY PLAN IMPLEMENTING PROCEDURES Revision 16 October 25, 2002 POST-ACCIDENT SAMPLING AND ANALYSIS OF POTENTIALLY HIGH ACTIVITY REACTOR COOLANT INITIALS INITIALS 5.15.5 Sign AND date in Section 5.16 for preparation of sample for off-site analysis.

5.16 Post-Analysis Documentation/Submittal 5.16.1 Sign/date for task performed/completed below AND submit to Chemistry supervision:

Post-Sampling Normal Valve Lineup:

Performer:

Printed Name Signature Date Sampling Vessel Installation:

Performer:

Printed Name Signature Date Preparation of Sample(s) For Off-Site Analysis:

Performer:

Printed Name Signature Date Approval:

Supervisor: Date Signature Printed Name 5.16.2 Forward Sections 5.11 - 5.16 to Rad/Chem Coordinator for re-attachment to Sections 1.0- 5.10 (submitted to Chemistry supervision at Step 5.9).

Page 47 of 54 CONTINUOUS USE ,

POINT BEACH NUCLEAR PLANT EPIP 8.4.1 SAFETY RELATED EMERGENCY PLAN IMPLEMENTING PROCEDURES Revision 16 October 25, 2002 POST-ACCIDENT SAMPLING AND ANALYSIS OF POTENTIALLY HIGH ACTIVITY REACTOR COOLANT

6.0 REFERENCES

6.1 Calculation 98-0058, Post-LOCA Reactor Coolant Sampling System Dose Rates, 08/10/98 6.2 Calculation 98-0018, Evaluation of WEPCO PBNP Post Accident Sampling System 6.3 Calculation 98-0175, Post-LOCA Post-Accident Sampling System I

6.4 NPM 94-0030, Compilation of WEPCO Commitments Associated with NUREG 0737, clarification of TMI Action Plan Requirements, 01/12/94.

7.0 BASES B-1 10 CFR 50.47, Code of Federal Regulations, Emergency Plans B-2 NUREG-0654, Criteria for Preparation and Evaluation of Radiological Emergency Response Plans and Preparedness in Support of Nuclear Power Plants, Rev. 1, November 1980 B-3 NUREG-0737, Clarification of TMI Action Plan Requirements, November 1980 Page 48 of 54 CONTINUOUS USE

EPIP 8.4.1 POINT BEACH NUCLEAR PLANT SAFETY RELATED EMERGENCY PLAN IMPLEMENTING PROCEDURES Revision 16 October 25, 2002 POST-ACCIDENT SAMPLING AND ANALYSIS OF POTENTIALLY HIGH ACTIVITY REACTOR COOLANT ATTACHMENT A PRIMARY COOLANT HIGH-LEVEL SAMPLING FLOW DIAGRAM Page 49 of 54 CONTINUOUS USE

EPIP 8.4.1 POINT BEACH NUCLEAR PLANT SAFETY RELATED EMERGENCY PLAN IMPLEMENTING PROCEDURES Revision 16 October 25, 2002 POST-ACCIDENT SAMPLING AND ANALYSIS OF POTENTIALLY HIGH ACTIVITY REACTOR COOLANT ATTACHMENT B SETUP FOR TRANSFER OF SAMPLE FROM SAMPLE VESSEL Vacuum Side Gas View Sample j& Vessel End 12 x View 20' Lead Lead forPig Samples Top View Gas Hollow Collection Vessel Lead 9B 9A 8B Bricks Staggered for Stability Page 50 of 54 CONTINUOUS USE

EPIP 8.4.1 POINT BEACH NUCLEAR PLANT SAFETY RELATED EMERGENCY PLAN IMPLEMENTING PROCEDURES Revision 16 October 25, 2002 POST-ACCIDENT SAMPLING AND ANALYSIS OF POTENTIALLY HIGH ACTIVITY REACTOR COOLANT ATTACHMENT C REPORT PBNP REACTOR COOLANT POST-ACCIDENT SAMPLING AND ANALYSIS Page 1 of 4 1.0 ANALYSIS OF GASEOUS SAMPLE 1.1 Hydrogen Analysis Hydrogen @ STP (using following equation with Table 1, Values A-D):

cA) 280 (C) 273 cc/KgH2 @*STP = ( x- x x1000 100 (B) 1013 (D)

Table 1: Hydrogen A Percent (%) hydrogen from instrument output B Volume of sample bomb (ml)

I C ' At,,,,-cnhp~rr nreqI~rP.{ hr r, At

- - V1-r;c I I t,ressure (mbaOI I ID -, t.-nn tor 4-,?771)

c/Kg hydrogen in coolant

• STP I I 1.2 Radioactive Noble Gases Corrected volume used for MCA analysis = - A B x(D) =E 280 oCt

-C=lIif no dilutions are necessary Table 2: Radioactive. Nloble Gases B. Noble Gases A. Analysis/Calculation Data Isotope gCi/cc Isotope pCi/cc A Vol of Sample vessel (ml)

Xe-133 Xe-138 B Vol of original Sample removed with syringe (ml)

Kr-85m Kr-87 C Vol of sidearm flask (ml) Ar-41 Kr-88 D rumber ofdilutions berformed Kr-85 Xe-133m E Corrected volume used for MCA analysis Xe -131m Xe-I Page 51 of 54 CONTINUOUS USE

EPIP 8.4.1 POINT BEACH NUCLEAR PLANT SAFETY RELATED EMERGENCY PLAN IMPLEMENTING PROCEDURES Revision 16 October 25, 2002 POST-ACCIDENT SAMPLING AND ANALYSIS OF POTENTIALLY HIGH ACTIVITY REACTOR COOLANT ATTACHMENT C REPORT PBNP REACTOR COOLANT POST-ACCIDENT SAMPLING AND ANALYSIS Page 2 of 4 2.0 ANALYSIS OF LIQUID SAMPLE 2.1 pH and Boron pH and Boron (with Boron calculated, using following equation with Table 3, Values A-C):

x (C) x 10810

• ppm Boron = (A) 2.2 Chloride Table 4: Chloride Chloride concentration I - ppm Page 52 of 54 CONTINUOUS USE

EPIP 8.4.1 POINT BEACH NUCLEAR PLANT SAFETY RELATED EMERGENCY PLAN IMPLEMENTING PROCEDURES Revision 16 October 25, 2002 POST-ACCIDENT SAMPLING AND ANALYSIS OF POTENTIALLY HIGH ACTIVITY REACTOR COOLANT ATTACHMENT C PBNP REACTOR COOLANT POST-ACCIDENT SAMPLING AND ANALYSIS REPORT Page 3 of 4 2.3 Iodine Analysis NOTE: Isotopes in Table 5, Column C, are NOT required. However, if time permits, analysis results for as many as possible should be provided to aid in completing EPIP 10.2, "Core Damage Estimation."

Table 5: Iodine and Other Isotopes for EPIP 10.2 B. Iodine i C. Other (for EPIP 10.2)

A. Analysis/Calculation Data Isotope Mult. Fact. Final *. Isotope R.Ci/cc A Volume of Oriinal Sample Corrected ,

(ml) m.-Ci/cc _

1130 x 4.5 = Rb-88 I B Dilution factors (if dilution factors are required)

Corrected volume used for 1-131 x 4.5 = Cs-134 MCA analysis (ml) 1-132 x 4.5 = Cs-137 D

1-133 x 4.5= Te-132 E

134 x 4.5 =, Ba-140 F

- 1-135 x 4.5= La-140 G

..... .. . ........... Total La-142 Pag.e 53 of 54 CONTINUOUS USE

POINT BEACH NUCLEAR PLANT EPIP 8.4.1 SAFETY RELATED EMERGENCY PLAN IMPLEMENTING PROCEDURES Revision 16 October 25, 2002 POST-ACCIDENT SAMPLING AND ANALYSIS OF POTENTIALLY HIGH ACTIVITY REACTOR COOLANT ATTACHMENT C PBNP REACTOR COOLANT POST-ACCIDENT SAMPLING AND ANALYSIS REPORT Page 4 of 4 3.0 PERFORMANCE (SAMPLING and ANALYSES)

SAMPLING:

Performer: Signature Date Printed Name Verifier:

Signature Date Printed Name ANALYSES (including calculations, as applicable)

Hydrogen: Date Performer Printed Name Performer Signature Noble Gas: Date Performer Printed Name Performer Signature pH/Boron: Date Performer Printed Name Performer Signature Chloride: Date Performer Printed Name Performer Signature Iodine/Other: Date Performer Printed Name Performer Signature 4.0 ROUTING AND APPROVAL Chemistry Leader Signature Date Printed Name Rad/Chem Coordinator: Signature Date Printed Name TSC Manager: Signature Date Printed Name Pace 54 of 54 CONTINUOUS USE

EPIP 1.2 EMERGENCY CLASSIFICATION DOCUMENT TYPE: Technical CLASSIFICATION: NNSR REVISION: 37 EFFECTIVE DATE: November 15, 2002 REVIEWER: Management Supervisory Staff APPROVAL AUTHORITY: Department Manager PROCEDURE OWNER (title): Emergency Preparedness OWNER GROUP: Emergency Preparedness Verified Current Copy:

Signature Date Time List pages used'for Partial Performance Controlling Work Document Numbers

POINT BEACH NUCLEAR PLANT EPIP 1.2 EMERGENCY PLAN IMPLEMENTING PROCEDURES NNSR Revision 37 EMERGENCY CLASSIFICATION November 15, 2002 TABLE OF CONTENTS SECTION TITLE PAGE 1.0 PU RPOSE ...................................................................................................................... 3 2.0 PREREQ U ISITES ..................................................................................................... 3 3.0 PRECAUTIONS AND LIMITATIONS ................................................................... 3 4.0 INITIAL CONDITIONS ............................................................................................ 4 5.0 PRO CED URE ........................................................................................................ 4 5.1 Classifying an Em ergency .......................................................................................... 4 5.2 Term inating an Em ergency ....................................................................................... 5 5.3 M issed C lassifications ................................................................................................. 6 6.0 REFEREN CES ...................................................................................................... 6 7.0 B A SE S ........................................................................................................................... 7 ATTACHMENT A EMERGENCY ACTION LEVEL (EAL) OVERVIEW MATRIX .................. 8 ATTACHMENT B EMERGENCY ACTION LEVELS (EALs) ................................................... 9 ATTACHMENT C FISSION PRODUCT BARRIER (FPB) MATRIX ........................................ 84 ATTACHMENT D SAFETY AND SAFETY-RELATED SYSTEMS .......................................... 92 Pag:e 2 of 94 REFERENCE USE

POINT BEACH NUCLEAR PLANT EPIP 1.2 EMERGENCY PLAN IMPLEMENTING PROCEDURES NNSR Revision 37 EMERGENCY CLASSIFICATION November 15, 2002 1.0 PURPOSE This procedure provides instructions to classify off-normal occurrences at PBNP into one of four standardized emergency classes.

2.0 PREREQUISITES 2.1 Responsibilities 2.1.1 This procedure is intended for immediate use by the Shift Manager (SM).

Following the activation of the Emergency Operations Facility (EOF) the overall responsibility for classification is assumed by the Emergency Director.

He is supported in this effort by Control Room, TSC, and EOF personnel.

2.1.2 When relieved of Emergency Director duties by the Emergency Director, the Shift Manager shall no longer be responsible for performance of actions specified in this procedure, however as an NRC licensee the SM shall bring to the attention of the Emergency Director changing plant conditions which may affect the emergency classification.

2.1.3 Upon activation of the TSC, the Operations Coordinator shall monitor plant conditions and provide event classification recommendations to the Emergency Director.

2.2 Equipment None 3.0 PRECAUTIONS AND LIMITATIONS 3.1 Notifications to state and local authorities must be made within 15 minutes of the declaration of an emergency. Notifications to the NRC must be made immediately following these, not to exceed one hour from declaration.

3.2 Category 8 EALs (Judgment) provide the ability to classify any set of plant conditions based on the Emergency Class definitions, derived from NUREG-0654.

3.3 Certain conditions or occurrences, while not meeting the threshold for classification as an emergency, may nonetheless be reportable to the NRC per 10 CFR 50.72. (Guidance on interpretation of the 10 CFR 50.72 criteria may be found in NUREG-1022.)

3.4 Continuously reference both plant conditions and the EALs in this procedure for potential re-classification.

3.5 When Emergency conditions exist on both Units due to separate events then each Unit should be classified according to the plant conditions and EALS. Units are independent of each other unless the event affects both units. If an event affects both units a single Emergency Classification is adequate.

Page 3 of 94 REFERENCE USE

POINT BEACH NUCLEAR PLANT EPIP 1.2 EMERGENCY PLAN IMPLEMENTING PROCEDURES NNSR Revision 37 EMERGENCY CLASSIFICATION November 15, 2002 4.0 INITIAL CONDITIONS EPIP 1.1 has been initiated by the Control Room because an off-normal occurrence exists (or has existed) at PBNP.

5.0 PROCEDURE 5.1 Classifying an Emergency 5.1.1 Record the time this procedure is entered in the station log (Emergency Director should log subsequent use in his log).

NOTE: A large version of Attachment A is available in the Control Room, TSC, and EOF.

5.1.2 Determine the category (or categories) of the event. (Column 1 of Attachment A). The categories are:

1. Fission Product Barriers

2. System Malfunction

3. Electrical Power

4. Radiological

5. Internal Events

6. External Events

7. Fuel Handling/ISFSI Events

8. Emergency Director Judgment 5.1.3 Make an initial EAL selection from Attachment A.

If the EAL relates to Category 1 (Fission Product Barriers), Attachment C provides additional information on the CHALLENGE and LOSS criteria.

NOTE: Do not "anticipate" challenge or loss of a barrier unless the trend is rapid, and the values are close to the threshold/criteria.

5.1.4 Reference the individual EAL page(s) in Attachment B for the EAL(s) selected. Read all fields on the page to determine/confirm that the EAL applies.

Page 4 of 94 REFERENCE USE

POINT BEACH NUCLEAR PLANT EPIP 1.2 EMERGENCY PLAN IMPLEMENTING PROCEDURES NNSR Revision 37 EMERGENCY CLASSIFICATION November 15, 2002 5.1.5 Also reference the individual EAL pages for the next higher and lower emergency class - in that category- (if such EALs exist). This should further confirm the initial selection and specific EAL.

5.1.6 IF an event has been categorized on Attachment A, and the threshold of the EAL and surrounding conditions verified to have been met or exceeded (Attachments B and C),

THEN declare the emergency.

a. Record the time of declaration, the emergency classification, and the EAL number on EPIP 2.1, Attachment B.

b. Make an announcement to your facility of the emergency and that you are assuming the duties of Emergency Director.

NOTE: IF this procedure is being implemented from the EOF, THEN verify Control Room is assisting with Gaitronics announcements.

c. Return to EPIP 1.1, Step 5.6 to ensure all appropriate actions are taken and coordinated with actions of the other ERFs if activated.

5.1.7 IF it is determined that no EAL is met, THEN review plant conditions against the criteria of 10 CFR 50.72 for one-hour and four-hour notifications to the NRC (Ref. NUREG-1022).

AND THEN return to EPIP 1.1, Step 5.20.

5.2 Terminating an Emergency IF conditions have improved where an EAL is no longer met, and it is believed that the plant is stable, i.e., the EAL is not anticipated to be exceeded again, THEN the emergency may be terminated per EPIP 12.1, WITH THE FOLLOWING CAVEATS:

5.2.1 IF Emergency Response Facilities have been activated, or personnel have been called to activate these facilities, THEN the event shall NOT be terminated until the TSC and EOF have been activated, and the TSC Manager concurs with the assessment of plant conditions.

5.2.2 IF any General Emergency has been declared, or any Protective Action Recommendation made to or by off-site authorities, THEN the emergency shall NOT be terminated until the NRC (for any General Emergency) and/or off-site authorities (for Protective Action Recommendations) concur.

Page 5 of 94 REFERENCE USE

POINT BEACH NUCLEAR PLANT EPIP 1.2 EMERGENCY PLAN IMPLEMENTING PROCEDURES NNSR Revision 37 EMERGENCY CLASSIFICATION November 15, 2002 5.3 Missed Classifications A missed classification is defined as a set of circumstances or events, which although no longer existing, if recognized at the time of their existence would have resulted in an emergency classification (i.e., met or exceeded an EAL of this procedure). This definition does not include conditions described in EALs which are based on expected plant response which does not occur, but where operator action was successful- such as failure of RPS.

NOTE: In ALL cases, the SM is vested with unilateral authority to classify an emergency and initiate any actions deemed appropriate to place the plant in a safe condition (per NUREG-0654, II.A.1.d, II.B.2).

5.3.1 If the missed classification would have been one classification, but current plant conditions warrant a lower classification, the lower classification shall be declared, but parties notified shall be informed of the temporary higher classification during the notification process.

5.3.2 If NO current plant conditions meeting any EAL exist at the time of discovery of the missed classification, the actual declaration of the emergency is not required; however an ENS notification should be made within one hour of the discovery of the undeclared event. Notify the Emergency Preparedness staff to ensure courtesy calls are made to offsite agencies.

6.0 REFERENCES

6.1 Technical Specifications 6.2 Final Safety Analysis Report (FSAR) Chapter 14, Appendix A 6.3 Point Beach Nuclear Plant Emergency Plan 6.4 Point Beach Design Basis Document (DBDs) 6.5 Abnormal Operating Procedures (AOPs) 6.6 Emergency Operating Procedures (EOPs) 6.7 Emergency Contingency Actions (ECAs) 6.8 Critical Safety Procedures (CSPs) 6.9 Point Beach Setpoint Document (STPT) 6.10 Security and Safeguards Contingency Plan Page 6 of 94 REFERENCE USE

POINT BEACH NUCLEAR PLANT EPIP 1.2 EMERGENCY PLAN IMPLEMENTING PROCEDURES NNSR Revision 37 EMERGENCY CLASSIFICATION November 15, 2002 6.11 WCAP 7525-L, Likelihood and Consequences of Turbine Overspeed at the Point Beach Nuclear Plant.

6.12 Reg Guide 1.115, Protection Against Low-Trajectory Turbine Missiles 6.13 EPRI Document, "Guidelines for Nuclear Plant Response to an Earthquake," dated October, 1989 6.14 Probabilistic Safety Assessment - High Winds, and Others Sec 9, Rev 0, Dated July 1995 6.15 Bechtel Corporation, "Westinghouse Electric Corporation-Wisconsin Michigan Power Company-Point Beach Atomic Power Station-Design Criteria for Nuclear Power Plants Against Tornadoes," March 12, 1970, B-TOP-3.

6.16 SOER 85-5, Internal Flooding of Power Plant Buildings 6.17 NUREG/CR-4982, "Severe Accident in Spent Fuel Pools in Support of Generic Safety Issue 82" 6.18 NRC Information Notice 90-08, "Kr-85 Hazards from Decayed Fuel" 6.19 NUREG-1022, Rev. 2, Event Reporting Guidelines 10CFR50.72 and 10CFR50.73.

7.0 BASES B-1 Code of Federal Regulation, 10 CFR 50 B-2 NUREG-0654/FEMA-REP-1, Criteria for Preparation and Evaluation of Radiological Emergency Response Plans and Preparedness in Support of Nuclear Plants, Revision 1, published November, 1980.

B-3 NUMARC NESP-007, Methodology for Development of Emergency Actions Levels, Revision 2, January 1992.

B-4 U.S. Regulatory Commission Position Paper, Branch Position on Acceptable Deviations to Appendix 1 to NUREG-0654/FEMA-REP-1, dated July 11, 1994.

Page 7 of 94 REFERENCE USE

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U z0 U z

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POINT BEACH NUCLEAR PLANT EPIP 1.2 NNSR EMERGENCY PLAN IMPLEMENTING PROCEDURES Revision 37 November 15, 2002 EMERGENCY CLASSIFICATION ATTACHMENT B EMERGENCY ACTION LEVELS (EALs)

Category: Fission Product Barriers EAL 1.1.1.1 Sub-Category: None Emergency Classification: UNUSUAL EVENT Emergency Action Level:

Reactor coolant sample activity greaterthan Technical Specification TS 3.4.16.

Basis:

This EAL is related to a Fission Product Barrier challenge. See Attachment C for additional information.

Elevated reactor coolant activity represents a potential degradation in the level of safety of the plant and is a potential precursor of more serious problems. This EAL addresses reactor coolant samples exceeding coolant Technical Specifications.

Technical Specifications allow exceeding normal coolant activities for limited time periods. This EAL does not apply while operating within these allowances.

Because RCS leakage and coolant activity are considered precursors to more serious events, and because they should be treated alike (each relating to a Fission Product Barrier) declaration shall be upon validation and shall NOT be delayed until Technical Specification's actions are taken.

References:

NUREG 0654, Appendix 1 Initiating Condition: Unusual Event 3b Page 9 of 94 REFERENCE USE

POINT BEACH NUCLEAR PLANT EPIP 1.2 EMERGENCY PLAN IMPLEMENTING PROCEDURES NNSR Revision 37 EMERGENCY CLASSIFICATION November 15, 2002 ATTACHMENT B EMERGENCY ACTION LEVELS (EALs)

Category: Fission Product Barriers EAL 1.1.1.2 Sub-Category: Loss of One Barrier Emergency Classification: ALERT Emergency Action Level:

Exceeding the LOSS threshold of either Fuel Clad OR Reactor Coolant System (RCS) barrierbased on FPB Matrix (See Attachment Cfor thresholds).

Basis:

This Fission Product Barrier (FPB) EAL refers to exceeding the LOSS threshold of either the Fuel Cladding or Reactor Coolant System barrier by comparing plant conditions to the thresholds outlined in the FPB Matrix (Attachment C).

The FPB Matrix LOSS criteria indicate values at which either the Fuel Cladding or RCS barrier has been breached to the point that it no longer serves as an effective barrier to the travel of fission products.

This value is not intended to represent total loss, however one of these two essential barriers is no longer serving its function. A substantial reduction in the level of safety at the plant exists, therefore an Alert classification is appropriate.

Loss of the Containment barrier (by itself) does not create an immediate transport of fission products as the Containment is designed to be a backup to the cladding and RCS barriers. Therefore, if only the Containment barrier is lost, it will be dealt with by Technical Specification action statements. However, if either the Fuel Cladding or RCS barrier is lost, the Containment barrier will be considered at the same level as these.

References:

NEI 97-03 Rev. 2 NUREG 0654, Appendix 1 Initiating Condition: Alert lb,lc, 5 Page 10 of 94 REFERENCE USE

POINT BEACH NUCLEAR PLANT EPIP 1.2 EMERGENCY PLAN IMPLEMENTING PROCEDURES NNSR Revision 37 EMERGENCY CLASSIFICATION November 15, 2002 ATTACHMENT B EMERGENCY ACTION LEVELS (EALs)

Category: Fission Product Barriers EAL 1.1.1.3 Sub-Category: Loss of Two Barriers Emergency Classification: SITE EMERGENCY Emergency Action Level:

Exceeding the LOSS threshold of any 2 fission product barriersbased on FPB Matrix (See Attachment C for thresholds).

Basis:

This Fission Product Barrier (FPB) EAL refers to exceeding the LOSS threshold of any two of the three fission product barriers; fuel cladding, reactor coolant system, or containment by comparing plant conditions to the thresholds outlined in the FPB Matrix (Attachment C).

The third barrier must remain INTACT. If challenged or lost, a General Emergency exists.

The FPB Matrix LOSS criteria indicate values at which barriers have been breached to the point that they no longer serve as effective barriers to the travel of fission products. These values are not intended to represent total loss, however two important barriers are no longer serving their function. This represents a major failure in plant systems needed to protect the public, therefore a Site Emergency classification is appropriate.

References:

NEI 97-03 NUREG 0654, Appendix 1 Initiating Condition: Site Emergency 5 Page 11 of 94 REFERENCE USE

POINT BEACH NUCLEAR PLANT EPIP 1.2 EMERGENCY PLAN IMPLEMENTING PROCEDURES NNSR Revision 37 EMERGENCY CLASSIFICATION November 15, 2002 ATTACHMENT B EMERGENCY ACTION LEVELS (EALs)

Category: Fission Product Barriers EAL 1.1.1.4 Sub-Category: Loss of Three Barriers Emergency Classification: GENERAL EMERGENCY Emergency Action Level:

Exceeding the LOSS threshold of any 2 fission product barriersAND exceeding the loss OR challenge threshold of the 3rd barrierbased on the FPBMatrix (See Attachment Cfor thresholds).

Basis:

This Fission Product Barrier (FPB) EAL refers to exceeding the LOSS threshold of any TWO of the three fission product barriers; fuel cladding, reactor coolant system, or containment AND also exceeding EITHER the loss OR challenge threshold on the third barrier by comparing plant conditions to the thresholds outlined in the FPB Matrix (Attachment C).

The FPB Matrix LOSS criteria indicate values at which barriers have been breached to the point that they no longer serve as effective barriers to the travel of fission products. These values are not intended to represent total loss, however the barriers are no longer serving their function. The loss of two and a loss or challenge of the third available barrier represents major failures to plant systems needed to protect the public with the actual or potential release of significant amounts of radioactive materials offsite, therefore a General Emergency classification is appropriate.

References:

NEI 97-03 Rev.2 NUREG 0654, Appendix 1 Initiating Condition: General Emergency 2 Page 12 of 94 REFERENCE USE

POINT BEACH NUCLEAR PLANT EPIP 1.2 EMERGENCY PLAN IMPLEMENTING PROCEDURES NNSR Revision 37 EMERGENCY CLASSIFICATION November 15, 2002 ATTACHMENT B EMERGENCY ACTION LEVELS (EALs)

Category: Fission Product Barriers EAL 1.1.2.1 Sub-Category: None Emergency Classification: UNUSUAL EVENT Emergency Action Level:

Failedfitelmonitor [1(2) RE-109] readinggreaterthan 120 mRem/hr, or 2 of 3 containment high range monitors readgreaterthan 1000 Rem/hr.

Basis:

This EAL is related to a Fission Product Barrier challenge. See Attachment C for additional information.

Other indications should accompany this indication, such as increased radiation on RE-106 or on hand-held instruments.

Elevated reactor coolant activity as indicated by the failed fuel monitor [1(2) RE-109] represents a potential degradation in the level of safety of the plant and is a potential precursor of more serious problems. This EAL addresses failed fuel monitor readings exceeding approximately 0.1 % fuel clad failures.

Because RCS leakage and coolant activity are considered precursors to more serious events, and because they should be treated alike (each relating to a Fission Product Barrier) declaration shall be upon validation and shall NOT be delayed until Technical Specification's actions are taken.

References:

NUREG 0654, Appendix 1 Initiating Condition: Unusual Event 3c Page 13 of 94 REFERENCE USE

POINT BEACH NUCLEAR PLANT EPIP 1.2 EMERGENCY PLAN IMPLEMENTING PROCEDURES NNSR Revision 37 EMERGENCY CLASSIFICATION November 15, 2002 ATTACHMENT B EMERGENCY ACTION LEVELS (EALs)

Category: Fission Product Barriers EAL 1.1.2.2 Sub-Category: Loss of One Barrier Emergency Classification: ALERT Emergency Action Level:

Unisolable steam line break outside containment with greater than 10 gpm , but less than 50 gpm, primary to secondary leakage.

Basis:

This EAL reflects a unique Initiating Condition from NUREG-0654. It does not meet the loss of one barrier criteria from Attachment C (Fission Product Barrier matrix), yet will be classified as an Alert.

Because an unisolable steam line break is evaluated under the Containment section of the Fission Product Barrier matrix, it would not result in an Alert by itself. Because the primary to secondary leakage rate (10 gpm) is less than the LOSS criteria for RCS, it would not result in an Alert. The 10 gpm does meet the CHALLENGE criteria therefore is an Unusual Event. However, there is no logic in the FPB matrix for combinations of LOSS of Containment with CHALLENGE of another barrier.

Due to the unique, specific criteria of NUREG-0654, Appendix 1 criteria, this EAL covers the unique condition of an unisolable steam line break, combined with a small primary to secondary leak.

If the steam line can be isolated, no emergency is warranted. If the steam line cannot be isolated, and the other Fission Product Barriers are INTACT (No leakage, or leakage below 10 gpm) then no emergency is applicable UNLESS the SM determines a potential degradation in the level of safety.

If the steam line cannot be isolated AND primary to secondary leakage is greater than 10 gpm, but less than 50 gpm, then this EAL applies and an Alert must be declared.

If the primary to secondary leak rate exceeds 50 gpm, then the LOSS criteria for RCS Fission Product Barrier has been met. This would constitute LOSS of two barriers, and would be a Site Emergency on EAL 1.1.1.3.

References:

NEI 97-03 Rev. 2 NUREG 0654, Appendix 1 Initiating Condition: Alert 4 Page 14 of 94 REFERENCE USE

POINT BEACH NUCLEAR PLANT EPIP 1.2 EMERGENCY PLAN IMPLEMENTING PROCEDURES NNSR Revision 37 EMERGENCY CLASSIFICATION November 15, 2002 ATTACHMENT B EMERGENCY ACTION LEVELS (EALs)

Category: Fission Product Barriers EAL 1.1.3.1 Sub-Category: None Emergency Classification: UNUSUAL EVENT Emergency Action Level:

Primary to secondary leakage greaterthan Technical Specification Reference TS 3.4.13.d (500 gallons

! per

! day in either steam generator). I Basis:

This EAL is related to a Fission Product Barrier challenge. See Attachment C for additional information.

Leakage from the RCS in excess of Technical Specifications is considered by the NRC to be a precursor to more serious events. Therefore, an Unusual Event must be declared even if Technical Specification actions are taken.

Because RCS leakage and coolant activity are considered precursors to more serious events, and because they should be treated alike (each relating to a Fission Product Barrier) declaration shall be upon validation and shall NOT be delayed until Technical Specification's actions are taken.

References:

NUREG 0654, Appendix 1 Initiating Condition: Unusual Event 5 Page 15 of 94 REFERENCE USE

POINT BEACH NUCLEAR PLANT EPIP 1.2 EMERGENCY PLAN IMPLEMENTING PROCEDURES NNSR Revision 37 EMERGENCY CLASSIFICATION November 15, 2002 ATTACHMENT B EMERGENCY ACTION LEVELS (EALs)

Category: Fission Product Barriers EAL 1.1.4.1 Sub-Category: None Emergency Classification: UNUSUAL EVENT Emergency Action Level:

Unisolableprimary system leakage greaterthan Technical Specification Reference TS 3.4.13.c (10 gallons per minute).

Basis:

This EAL is related to a Fission Product Barrier challenge. See Attachment C for additional information.

Leakage from RCS in excess of Technical Specifications which cannot be isolated is considered by the NRC to be a precursor to more serious events. Therefore, an Unusual Event must be declared even if Technical Specification actions are taken.

Because RCS leakage and coolant activity are considered precursors to more serious events, and because they should be treated alike (each relating to a Fission Product Barrier) declaration shall be upon validation and shall NOT be delayed until Technical Specification's actions are taken.

References:

PBNP Technical Specifications NUREG 0654, Appendix 1 Initiating Condition: Unusual Event 5 Page 16 of 94 REFERENCE USE

POINT BEACH NUCLEAR PLANT EPIP 1.2 EMERGENCY PLAN IMPLEMENTING PROCEDURES NNSR Revision 37 EMERGENCY CLASSIFICATION November 15, 2002 ATTACHMENT B EMERGENCY ACTION LEVELS (EALs)

Category: Fission Product Barriers EAL 1.1.5.1 Sub-Category: None Emergency Classification: UNUSUAL EVENT Emergency Action Level:

Excess RCS cooldown or cold overpressutrizationof the RCS (ST-4 Integrity Orange path)

Basis:

The following conditions meet ST-4 Integrity - Orange Path criteria. A challenge to the RCS barrier is present due to excessive cooldown or cold overpressurization as indicated below:

Decrease in temperature in either cold leg greater than 100°F in the last 60 minutes AND temperature in either cold leg less than 315'F.

OR Temperature in either cold leg less than 315'F and RCS pressure greater than 425 psig.

Any actual loss of RCS barrier warrants declaration of an Alert per the FPB matrix, Attachment C.

References:

NUREG 0654, Appendix 1 Initiating Condition: Unusual Event 17 Page 17 of 94 REFERENCE USE

POINT BEACH NUCLEAR PLANT EPIP 1.2 EMERGENCY PLAN IMPLEMENTING PROCEDURES NNSR Revision 37 EMERGENCY CLASSIFICATION November 15, 2002 ATTACHMENT B EMERGENCY ACTION LEVELS (EALs)

Category: System Malfunctions EAL 2.1.1.2 Sub-Category: Failure to Trip Emergency Classification: ALERT Emergency Action Level:

Failureof the reactorprotection system(automatic or manual) to initiate and complete a trip which brings the reactorsubcritical.

Basis:

The reactor protection system may be actuated either by automatic means (exceeding pre-determined thresholds which result in trip signals) or by operator action (manual trip).

The failure of EITHER of these means to cause a trip with subsequent subcriticality meets this EAL (an Alert).

If BOTH these means AND all other means from the Control Room fail, see EAL 2.1.1.3 (a Site Emergency).

References:

NUREG 0654, Appendix 1 Initiating Condition: Alert 11 Page 18 of 94 REFERENCE USE

POINT BEACH NUCLEAR PLANT EPIP 1.2 EMERGENCY PLAN IMPLEMENTING PROCEDURES NNSR Revision 37 EMERGENCY CLASSIFICATION November 15, 2002 ATT~ACHMENT B EMERGENCY ACTION LEVELS (EALs)

Category: System Malfunctions EAL 2.1.1.3 Sub-Category: Failure to Trip Emergency Classification: SITE EMERGENCY Emergency Action Level:

IFailureto rapidly bring the reactorsubcriticalfrom the Control Room. (ST-1 SubcriticalityRed Path) I Basis:

CSFST Subcriticality - RED path is entered based on failure of power range indication (N-41, N-42, N-43, N-44) to decrease below 5% following a reactor trip. This EAL addresses any manual trip or automatic trip signal followed by a manual trip or other Control Room actions which fail to rapidly shut down the reactor.

If any actions must be taken outside the Control Room to effect a reactor trip this EAL is also met.

This condition indicates failure of both the automatic and manual protection systems to trip the reactor, to an extent that emergency boration is required: or actions are needed outside the Control Room to trip the reactor. The failure of both front line and backup protection systems to function in response to a plant transient, along with the continued production of heat, poses a direct threat to fuel clad and RCS integrity and thus warrants declaration of a Site Emergency.

This EAL is synonymous with entry into CSP S-1.

References:

NUREG 0654, Appendix 1 Initiating Condition: Site Emergency 9 Page 19 of 94 REFERENCE USE

POINT BEACH NUCLEAR PLANT EPIP 1.2 EMERGENCY PLAN IMPLEMENTING PROCEDURES NNSR Revision 37 EMERGENCY CLASSIFICATION November 15, 2002 ATTACHMENT B EMERGENCY ACTION LEVELS (EALs)

Category: System Malfunctions EAL 2.2.1.1 Sub-Category: Technical Specification Requirements Emergency Classification: UNUSUAL EVENT Emergency Action Level:

Failureto reach Technical Specification requiredoperating mode or condition within the specified time limit of the LCO action statement.

Basis:

Limiting Conditions of Operation (LCOs) action statements require the plant to be brought to a required condition (often shutdown) when the Technical Specification required configuration cannot be restored.

Depending on the circumstances, this may or may not be an emergency or precursor to a more severe condition. In any case, the initiation of plant shutdown required by the site Technical Specification requires a four hour report under 10 CFR 50.72 (b) non-emergency events. The plant remains within its evaluated safety envelope while changing conditions or being shut down so long as it is accomplished within the completion time for the required action in the Technical Specifications.

An immediate Unusual Event is required when the plant is not brought to the required operating mode or condition within the allowable action statement time of the Technical Specifications. Declaration of an Unusual Event is based on the time at which the LCO-specified action statement time period elapses under the site Technical Specifications and is not related to how long a condition may have existed.

If a Notice of Enforcement Discretion (NOED) is approved by the NRC prior to the LCO action statement time expiration an emergency need not be declared.

References:

NUREG 0654, Appendix 1 Initiating Condition: Unusual Event 15 Page 20 of 94 REFERENCE USE

POINT BEACH NUCLEAR PLANT EPIP 1.2 EMERGENCY PLAN IMPLEMENTING PROCEDURES NNSR Revision 37 EMERGENCY CLASSIFICATION November 15, 2002 ATTACHMENT B EMERGENCY ACTION LEVELS (EALs)

Category: System Malfunctions EAL 2.3.1.1 Sub-Category: Loss of Indications/Communications Emergency Classification: UNUSUAL EVENT Emergency Action Level:

Unplanned loss of most (approximately 75%) safety system annunciatorsor indicationson Control Room Panelsfor greaterthan 15 minutes AND increasedmonitoring is requiredfor safe plant operation.

Basis:

This EAL recognizes the difficulty associated with monitoring changing plant conditions without the use of a major portion of the annunciation or indication equipment. Recognition of the normal availability of computer based indication equipment is considered.

"Unplanned" loss of annunciators or indicators excludes scheduled maintenance and testing activities, which should not disable such large portions of the system(s).

It is not intended that personnel perform a count of the instrumentation or annunciation lost but use the judgment of the SM as the threshold for determining the severity of the plant condition. The increased monitoring portion of this EAL is met if the SM determines that additional personnel are required to provide increased monitoring of system operation to safely operate the plant.

It is recognized that most plant designs provide redundant safety system indication powered from separate uninterruptible power supplies. While failure of a large portion of annunciators is more likely than a failure of a large portion of indications, the concern is included in this EAL due to difficulty associated with assessment of plant conditions. The loss of specific safety system indicators should remain a function of that specific system or component operability status, and is addressed by the specific Technical Specifications.

Safety systems as used here designates systems with safety-related functions. Attachment D lists safety systems and systems with safety-related functions.

Fifteen minutes was selected as a threshold to exclude transient or momentary losses.

References:

NUREG 0654, Appendix 1 Initiating Condition: Unusual Event 14 Page 21 of 94 REFERENCE USE

POINT BEACH NUCLEAR PLANT EPIP 1.2 EMERGENCY PLAN IMPLEMENTING PROCEDURES NNSR Revision 37 EMERGENCY CLASSIFICATION November 15, 2002 ATTACHMENT B EMERGENCY ACTION LEVELS (EALs)

Category: System Malfunctions EAL 2.3.1.2 Sub-Category: Loss of Indications/Communications Emergency Classification: ALERT Emergency Action Level:

Unplanned loss of most (approximately 75%) safety system annunciatorsor indicationson Control Room Panelsfor greaterthan 15 minutes AND Increased monitoring is requiredfor safe plant operation AND either:

A significantplant transientis in progress OR PPCS is unavailable.

Basis:

This EAL recognizes the difficulty associated with monitoring changing plant conditions without the use of a major portion of the annunciation or indication equipment during a transient. Recognition of the normal availability of computer based indication equipment is also considered.

"Unplanned" loss of annunciators or indicators excludes scheduled maintenance and testing activities, which should not disable such large portions of the system(s).

Safety systems as used here designates systems with safety-related functions. Attachment D lists safety systems and systems with safety-related functions.

It is not intended that personnel perform a count of the instrumentation or annunciation lost but the use the judgment of the SM as the threshold for determining the severity of the plant conditions. The increased monitoring portion of this EAL is met if the SM determines that additional personnel are required to provide increased monitoring of system operation to safely operate the plant.

It is recognized that most plant designs provide redundant safety system indication powered from separate uninterruptible power supplies. While failure of a large portion of annunciators is more likely than a failure of a large portion of indications, the concern is included in this EAL due to difficulty associated with assessment of plant conditions. The loss of specific, or several, safety system indicators should remain a function of that specific system or component operability status and is addressed by the specific Technical Specifications.

Page 22 of 94 REFERENCE USE

POINT BEACH NUCLEAR PLANT EPIP 1.2 EMERGENCY PLAN IMPLEMENTING PROCEDURES NNSR Revision 37 EMERGENCY CLASSIFICATION November 15, 2002 ATTACHMENT B EMERGENCY ACTION LEVELS (EALs)

"Significant transient" includes response to automatic or manually initiated functions such as trips, runbacks involving greater than 25% thermal power change, ECCS injections, or thermal power ramps of 10% or greater.

If both a major portion of the annunciation system and all computer monitoring is unavailable to the extent that additional personnel are required to monitor indications, the Alert is required. If the operating crew cannot monitor the transient in progress this will be escalated to a Site Emergency.

References:

NUREG 0654, Appendix 1 Initiating Condition: Alert 14 Page 23 of 94 REFERENCE USE

POINT BEACH NUCLEAR PLANT EPIP 1.2 EMERGENCY PLAN IMPLEMENTING PROCEDURES NNSR Revision 37 EMERGENCY CLASSIFICATION November 15, 2002 (EALs)

B LEVELS ATTACHMENT BACTION ATTACHMENTEMERGENCY EMERGENCY ACTION LEVELS (EALs)

Category: System Malfunctions EAL 2.3.1.3 Sub-Category: Loss of Indications/Communications Emergency Classification: SITE EMERGENCY Emergency Action Level:

Unplanned loss of most (approx. 75%) safety system annunciatorsor indicationson ControlRoom Panels.

AND Loss of ability to monitor criticalsafety function status AND A significantplant transient in progress.

Basis:

This EAL recognizes the INABILITY of the Control Room staff to monitor the plant response to a transient. A Site Emergency is considered to exist if the Control Room staff cannot monitor safety functions needed for protection of the public.

"Significant transient" includes response to automatic or manually initiated functions such as trips, runbacks involving greater than 25% thermal power change, ECCS injections, or thermal power ramps of 10% or greater.

Safety systems as used here designates systems with safety-related functions. Attachment D lists safety systems and systems with safety-related functions.

"Unplanned" loss of annunciators or indicators excludes scheduled maintenance and testing activities, which should not disable such large portions of the system(s).

References:

NUREG 0654, Appendix 1 Initiating Condition: Site Emergency 12 Page 24 of 94 REFERENCE USE

POINT BEACH NUCLEAR PLANT EPIP 1.2 EMERGENCY PLAN IMPLEMENTING PROCEDURES NNSR Revision 37 EMERGENCY CLASSIFICATION November 15, 2002 ATTACHMENT B EMERGENCY ACTION LEVELS (EALs)

Category: System Malfunctions EAL 2.3.2.1 Sub-Category: Loss of Indications/Communications Emergency Classification: UNUSUAL EVENT Emergency Action Level:

Loss of all communications capabilityaffecting the ability to either:

Perform routine operations OR Notify offsite agencies orpersonnel.

Basis:

The purpose of this EAL is to recognize a loss of communications capability that EITHER defeats the plant operation's or staffs ability to perform routine tasks necessary for plant operations OR the ability to communicate problems with offsite authorities. The loss of offsite communications ability anticipated by this EAL is expected to be significantly more comprehensive than the condition addressed by 10 CFR 50.72.

The onsite communications loss must encompass the loss of all means of routine communications (i.e., plant telephone system, Gaitronics page system, portable radios).

The offsite communications loss must encompass the loss of all means of communications with offsite authorities. This should include Emergency Notification System (ENS) for NRC, Microwave lines, and radio. This EAL is also met when extraordinary means are being utilized to make communications possible (relaying of information from radio transmissions, individuals being sent to offsite locations, etc.).

Procedure DCS 2.1.1 describes lesser communications losses which must be reported to the NRC within eight hours.

References:

NUREG 0654, Appendix 1 Initiating Condition: Unusual Event 11 Page 25 of 94 REFERENCE USE

POINT BEACH NUCLEAR PLANT EPIP 1.2 EMERGENCY PLAN IMPLEMENTING PROCEDURES NNSR Revision 37 EMERGENCY CLASSIFICATION November 15, 2002 (EALs)

B LEVELS ATTACHMENT BACTION ATTfACHMENTEMERGENCY EMERGENCY ACTION LEVELS (EALs)

Category: System Malfunctions EAL 2.4.1.2 Sub-Category: Degradation of Safety System Performance Emergency Classification: ALERT Emergency Action Level:

Inability to maintain reactorcoolant temperatureless than or equal to 200 7.

Basis:

This EAL addresses complete loss of functions required for core cooling during refueling and cold shutdown modes. Escalation to Site Emergency or General Emergency would be through other EALs.

An uncontrollable reactor coolant temperature increase that approaches or exceeds the cold shutdown technical specification limit warrants declaration of an Alert. The concern of this EAL is the loss of control resulting in the loss of ability to maintain the plant in cold shutdown which is defined by reactor coolant temperature.

References:

NUREG 0654, Appendix 1 Initiating Condition: Alert 10 Page 26 of 94 REFERENCE USE

POINT BEACH NUCLEAR PLANT EPIP 1.2 EMERGENCY PLAN IMPLEMENTING PROCEDURES NNSR Revision 37 EMERGENCY CLASSIFICATION November 15, 2002 ATTACHMENT B EMERGENCY ACTION LEVELS (EALs)

Category: System Malfunctions EAL 2.4.1.3 Sub-Category: Degradation of Safety System Performance Emergency Classification: SITE EMERGENCY Emergency Action Level:

Primaryto secondary leakage greaterthan 400 gallons per minute AND Inability to power BOTH buses A-05 AND A-06 from offsite sources.

Basis:

400 gpm is also the expected output from a single SI pump @ 1400 psia RCS pressure. (See DBD-09).

Loss of offsite power combined with an RCS leak (from Primary to Secondary) of this magnitude constitute several major challenges to the protection of the public:

1. Operating on diesel generators.

2. Leak (rupture) near the capacity of a single Safety Injection pump.

3. Transport of any fission products from Primary to Secondary.

Therefore, major plant functions needed for the protection of the public have been affected. A Site Emergency is warranted.

References:

NUREG 0654, Appendix 1 Initiating Condition: Site Emergency 3 Page 27 of 94 REFERENCE USE

POINT BEACH NUCLEAR PLANT EPIP 1.2 EMERGENCY PLAN IMPLEMENTING PROCEDURES NNSR Revision 37 EMERGENCY CLASSIFICATION November 15, 2002 (EALs)

B LEVELS ATTACHMENT BACTION ATTACHMENTEMERGENCY EMERGENCY ACTION LEVELS (EALs)

Category: Loss of Electrical Power EAL 3.1.1.1 Sub-Category: Loss of Vital AC Power Emergency Classification: UNUSUAL EVENT Emergency Action Level:

Loss of all offsite AC capability to vital buses as indicatedby the inability to power BOTH buses A-05 AND A-06 of a given unitfrom offsite sourcesfor greaterthan 15 minutes. I Basis:

Prolonged loss of offsite AC power reduces required redundancy and potentially degrades the level of safety of the plant by rendering the plant more vulnerable to a complete loss of AC power (station blackout). Fifteen minutes was selected as a threshold to exclude transient or momentary power losses.

Therefore, this condition (which is indicative of degraded conditions, but with no adverse consequences on the public health and safety) is classified as an UNUSUAL EVENT.

If primary to secondary leakage also exists, see EAL 2.4.1.3.

References:

FSAR Section 8, Electrical Systems DBD-22, 4160 VAC System NUREG 0654, Appendix 1 Initiating Condition: Unusual Event 7a Page 28 of 94 REFERENCE USE

POINT BEACH NUCLEAR PLANT EPIP 1.2 EMERGENCY PLAN IMPLEMENTING PROCEDURES NNSR Revision 37 EMERGENCY CLASSIFICATION November 15, 2002 ATTACHMENT B EMERGENCY ACTION LEVELS (EALs)

Category: Loss of Electrical Power EAL 3.1.1.2 Sub-Category: Loss of Vital AC Power Emergency Classification: ALERT Emergency Action Level:

Loss of all safeguardbus AC power of a given unit as indicatedby the inability to power BOTH buses A-05 AND A-06, OR B-03 AND B-04.

AND Loss isfor less than 15 minutes.

Basis:

Loss of all AC power safeguards buses compromises critical plant safety functions including RHR, ECCS, containment heat removal, and maintaining the ultimate heat sink. Prolonged loss of all AC power safeguards buses may result in uncovering the core and loss of containment integrity, thus this event can escalate to a General Emergency. The site blackout coping analysis assumes that AC power can be restored in one hour.

This condition is entered when there are indications of a total loss of power to the safeguards buses A-05 and A-06 OR B-03 and B-04 from any source (on or off-site) for less than 15 minutes.

This condition is indicative of actual or potential substantial degradation to plant systems with possible adverse consequences on the public health and safety. An ALERT is warranted and must be declared.

This EAL escalates to a SITE EMERGENCY if loss of AC power continues for greater than 15 minutes.

References:

FSAR Section 8, Electrical Systems DBD-22, 4160 VAC System NUREG 0654, Appendix 1 Initiating Condition: Alert 7 Page 29 of 94 REFERENCE USE

POINT BEACH NUCLEAR PLANT EPIP 1.2 EMERGENCY PLAN IMPLEMENTING PROCEDURES NNSR Revision 37 EMERGENCY CLASSIFICATION November 15, 2002 ATTACHMENT B EMERGENCY ACTION LEVELS (EALs)

Category: Loss of Electrical Power EAL 3.1.1.3 Sub-Category: Loss of Vital AC Power Emergency Classification: SITE EMERGENCY Emergency Action Level:

Loss of all safeguardbus AC power of a given unit as indicatedby the inabilityto power BOTH buses A-05 AND A-06, OR B-03 AND B-04.

AND Loss is for greaterthan 15 minutes.

Basis:

Loss of all AC power safeguards buses compromises critical plant safety functions including RHR, ECCS, containment heat removal, and maintaining the ultimate heat sink. Prolonged loss of all AC power safeguards buses may result in the uncovering core and loss of containment integrity, thus this event can escalate to a General Emergency. The site blackout coping analysis assumes that AC power can be restored in one hour.

This condition is entered when there are indications of a total loss of power to the safeguards buses A-05 and A-06 OR B-03 and B-04 from any source (on or off-site) for more than 15 minutes.

Therefore, this condition (which is indicative of serious plant system conditions with adverse consequences on the public health and safety) is classified as a SITE EMERGENCY.

References:

FSAR Section 8, Electrical Systems DBD-22, 4160 VAC System NUREG 0654, Appendix 1 Initiating Condition: Site Emergency 6 Page 30 of 94 REFERENCE USE

POINT BEACH NUCLEAR PLANT EPIP 1.2 EMERGENCY PLAN IMPLEMENTING PROCEDURES NNSR Revision 37 EMERGENCY CLASSIFICATION November 15, 2002 ATTACHMENT B EMERGENCY ACTION LEVELS (EALs)

Category: Loss of Electrical Power EAL 3.1.1.4 Sub-Category: Loss of Vital AC Power Emergency Classification: GENERAL EMERGENCY Emergency Action Level:

Loss of all safeguardbus AC power of a given unit as indicatedby the inability to power BOTH buses A-05 AND A-06, OR B-03 AND B-04.

AND Loss is greaterthan 15 minutes.

AND Both narrow range S/G level less than [51%] 29% AND totalfeedwaterflow to S/Gs less than 200 gpm.

(ST-3 Heat Sink Red path)

Basis:

Loss of all AC power safeguards buses compromises critical plant safety functions including RHR, ECCS, containment heat removal, and maintaining the ultimate heat sink. Prolonged loss of all AC power safeguards buses may result in the uncovering core and loss of containment integrity, thus this event can escalate to a General Emergency. The site blackout coping analysis assumes that AC power can be restored in one hour.

This EAL assures that in the event of a prolonged station blackout, timely recognition of the loss of heat sink occurs.

Therefore, this condition is indicative of grave plant conditions with potential adverse consequences on the public health and safety. A GENERAL EMERGENCY is warranted and must be declared.

References:

FSAR Section 8, Electrical Systems DBD-22, 4160 VAC System NUREG 0654, Appendix 1 Initiating Condition: General Emergency 5d Page 31 of 94 REFERENCE USE

POINT BEACH NUCLEAR PLANT EPIP 1.2 EMERGENCY PLAN IMPLEMENTING PROCEDURES NNSR Revision 37 EMERGENCY CLASSIFICATION November 15, 2002 (EALs)

B LEVELS ATTACHMENT BACTION ATTIACHMENTEMERGENCY EMERGENCY ACTION LEVELS (EALs)

Category: Loss of Electrical Power EAL 3.1.2.1 Sub-Category: Loss of Vital AC Power Emergency Classification: UNUSUAL EVENT Emergency Action Level:

Loss of all onsite AC power capability to power BOTH buses A-05 AND A-06 of a given unitfrom onsite sources (GO] through G04)for greaterthan 15 minutes.

Basis:

Loss of onsite safety related AC power sources reduces required redundancy and potentially degrades the level of safety of the plant by rendering the plant more vulnerable to a complete loss of AC power (station blackout). Therefore, an Unusual Event is warranted and must be declared. Fifteen minutes was selected as a threshold to exclude transient losses.

This condition is entered when there are indications of the unavailability of all the emergency diesel generators (GO 1 through G04) or that none of these sources can be aligned to either A-05 or A-06 for greater than 15 minutes.

Therefore, this condition (which is indicative of degraded conditions, but with no adverse consequences on the public health and safety) is classified as an Unusual Event.

References:

FSAR Section 8, Electrical Systems DBD-22, 4160 VAC System NUREG 0654, Appendix 1 Initiating Condition: Unusual Event 7b Pagye 32 of 94 REFERENCE USE

POINT BEACH NUCLEAR PLANT EPIP 1.2 EMERGENCY PLAN IMPLEMENTING PROCEDURES NNSR Revision 37 EMERGENCY CLASSIFICATION November 15, 2002 ATTACHMENT B EMERGENCY ACTION LEVELS (EALs)

Category: Loss of Electrical Power EAL 3.2.1.2 Sub-Category: Loss of Vital DC Power Emergency Classification: ALERT Emergency Action Level:

Loss of all vital DCpower as indicated by less than 105 vdc on all station battery buses (DO], D02, D03, D04)for less than 15 minutes.

Basis:

Loss of all vital DC power compromises the ability to monitor and control plant safety functions.

Prolonged loss of all DC power may result in uncovering the core and loss of containment integrity.

Loss of DC power to any AC bus creates the following conditions:

1. Associated breakers cannot be electrically opened or closed remotely or locally;

2. Electrical protection/interlock tripping of associated breakers is rendered inoperable including undervoltage stripping. The one exception is the 480 V individual breaker overloads which remain operable;

3. All associated breaker positions remain AS IS.

Loss of all vital onsite DC power may also be indicated by an "Annunciator Power Failure" alarm.

This EAL escalates to a SITE EMERGENCY if the power loss continues for greater than 15 minutes.

References:

FSAR Section 8, Electrical Systems DBD-19, 125 VDC System NUREG 0654, Appendix 1 Initiating Condition: Alert 8 Page 33 of 94 REFERENCE USE

POINT BEACH NUCLEAR PLANT EPIP 1.2 EMERGENCY PLAN IMPLEMENTING PROCEDURES NNSR Revision 37 EMERGENCY CLASSIFICATION November 15, 2002 ATTACHMENT B EMERGENCY ACTION LEVELS (EALs)

Category: Loss of Electrical Power EAL 3.2.1.3 Sub-Category: Loss of Vital DC Power Emergency Classification: SITE EMERGENCY Emergency Action Level:

Loss of all vital DC power as indicatedby less than 105 vdc on all station battery buses (DO], D02, D03, D04)for greaterthan 15 minutes.

Basis:

Loss of all vital DC power compromises the ability to monitor and control plant safety functions.

Prolonged loss of all DC power may result in uncovering the core and loss of containment integrity.

Loss of DC power to any AC bus creates the following conditions:

1. Associated breakers cannot be electrically opened or closed remotely or locally;

2. Electrical protection/interlock tripping of associated breakers is rendered inoperable including undervoltage stripping. The one exception is the 480 V individual breaker overloads which remain operable.

3. All associated breaker positions remain AS IS.

Loss of all vital onsite DC power may also be indicated by an "Annunciator Power Failure" alarm.

This condition (which is indicative of possible loss of control of the reactor coolant and containment barriers, with possible adverse consequences on the public health and safety) is classified as a SITE EMERGENCY.

References:

FSAR Section 8, Electrical Systems DBD-19, 125 VDC System NUREG 0654, Appendix 1 Initiating Condition: Site Emergency 7 Page 34 of 94 REFERENCE USE

POINT BEACH NUCLEAR PLANT EPIP 1.2 EMERGENCY PLAN IMPLEMENTING PROCEDURES NNSR Revision 37 EMERGENCY CLASSIFICATION November 15, 2002 ATTACHMENT B EMERGENCY ACTION LEVELS (EALs)

Category: Radiological Conditions EAL 4.1.1.1 Sub-Category: Off-site Radiological Release Emergency Classification: UNUSUAL EVENT Emergency Action Level:

Vent radiationreading(s)exceed the high alarm setpointsforgreaterthan 60 minutes, OR Liquid release in excess of alarm setpoints which cannot be isolated.

Vent Radiation High Alarm Setpoints Reference RMSARB for current 9/99 setpoint values I RE 212 2.73 E-4 uCi/cc if purging, 1.62E-2 if forced vent.

1 RE 215 2.71 E+0 uCi/cc RE 214 1.02 E-4 uCi/cc RE 221 1.58 E-4 uCi/cc RE 224 2.09 E-3 uCi/cc RE 225 1.36 E+O uCi/cc 2 RE 212 1.78 E-4 uCi/cc if purging, 1.82E-2 if forced vent.

2 RE 215 2.71 E + 0 uCi/cc Liquid Release Limits I Service Water Discharge Waste Water Effluent 1(2) RE-229 High Alarm RE-230 High Alarm AND: AND:

1 Circ. Water pump Release Limit 1 Circ. Water pump AND: (uCi/cc): Release Limit (uCi/cc):

2 Service Water pumps 4.12 E-5 3.70 E-4 3 Service Water pumps 3.27 E-5 4 Service Water pumps 3.03 E-5 5 Service Water pumps 2.87 E-5 6 Service Water pumps 2.78 E-5 2 Circ. Water pump Release Limit 2 Circ. Water pump AND: (uCi/cc): Release Limit (uCi/cc):

2 Service Water pumps 7.00 E-5 6.29 E-4 3 Service Water pumps 5.56 E-5 4 Service Water pumps 5.15 E-5 5 Service Water pumps 4.88 E-5 6 Service Water pumps 4.73 E-5

Reference:

C.H. Onesti to G.J. Maxfield, 11/17/92, RE-229 and RE-230 Alarm Setpoints, NPM 92-1035.

Page 35 of 94 REFERENCE USE

POINT BEACH NUCLEAR PLANT EPIP 1.2 EMERGENCY PLAN IMPLEMENTING PROCEDURES NNSR Revision 37 EMERGENCY CLASSIFICATION November 15, 2002 ATTACHMENT B EMERGENCY ACTION LEVELS (EALs)

Basis:

Unplanned airborne releases in excess of the site technical specifications, that cannot be reduced to within technical specifications within 60 minutes, represent an uncontrolled situation and hence, a potential degradation in the level of safety. The final integrated dose (which is very low in the Unusual Event emergency class) is not the primary concern here; it is the degradation in plant control implied by the fact that the release was not controlled to within Technical Specification limits within 60 minutes.

Therefore, it is not intended that the release be averaged over 60 minutes. For example, a release of 2 times Technical Specifications for 30 minutes, but which is terminated, does not exceed this EAL.

However, the SM should not wait until 60 minutes has elapsed, but should declare the event as soon as it is determined that the release duration has or will likely exceed 60 minutes.

Likewise, liquid release values (which would result in very low integrated dose) are not the primary concern. Rather, the fact that the release cannot be isolated represents a potential degradation in the level of safety.

References:

STPT 13.4, Effluent Monitors NUREG 0654, Appendix 1 Initiating Condition: Unusual Event 2 Pag~e 36 of 94 REFERENCE USE

POINT BEACH NUCLEAR PLANT EPIP 1.2 EMERGENCY PLAN IMPLEMENTING PROCEDURES NNSR Revision 37 EMERGENCY CLASSIFICATION November 15, 2002 ATTACHMENT B EMERGENCY ACTION LEVELS (EALs)

Category: Radiological Conditions EAL 4.1.1.2 Sub-Category: Off-site Radiological Release Emergency Classification: ALERT Emergency Action Level:

Vent radiationreadingsexceed ten times the high alarm setpointsforgreaterthan 15 minutes.

OR Liquid release in excess often times alarmsetpoint which cannot be isolated.

10 times Vent Radiation High Reference RMSARB for current setpoint values Alarm Setpoints 9/99 1 RE 212 2.73 E-3 uCi/cc if purging, 1.62E-1 if forced vent 1 RE 215 2.71 E+1 uCi/cc RE 214 1.02 E-3 uCi/cc RE 221 1.58 E-3 uCi/cc RE 224 2.09 E-2 uCi/cc RE 225 1.36 E+I uCi/cc 2 RE 212 1.78 E-3 uCi/cc if purging, 1.82E-1 if forced vent 2 RE 215 2.71 E+1 uCi/cc Liquid Release Limits Service Water Discharge Waste Water Effluent 1(2) RE-229 High Alarm RE-230 High Alarm AND:

AND:

1 Circ. Water pump Ten times Release 1 Circ. Water pump AND: Limit (uCi/cc): Ten times Release Limit (uCi/cc):

2 Service Water pumps 4.12 E-4 3.70 E-3 3 Service Water pumps 3.27 E-4 4 Service Water pumps 3.03 E-4 5 Service Water pumps 2.87 E-4 6 Service Water pumps 2.78 E-4 2 Circ. Water pump Ten times Release 2 Circ. Water pump AND: Limit (uCi/cc): Ten times Release Limit (uCi/cc):

2 Service Water pumps 7.00 E-4 6.29 E-3 3 Service Water pumps 5.56 E-4 4 Service Water pumps 5.15 E-4 5 Service Water pumps 4.88 E-4 6 Service Water pumps 4.73 E-4

Reference:

C.H. Onesti to G.J. Maxfield, 11/17/92, RE-229 and RE-230 Alarm Setpoints, NPM 92-1035.

Page 37 of 94 REFERENCE USE

POINT BEACH NUCLEAR PLANT EPIP 1.2 EMERGENCY PLAN IMPLEMENTING PROCEDURES NNSR Revision 37 EMERGENCY CLASSIFICATION November 15, 2002 ATTACHMENT B EMERGENCY ACTION LEVELS (EALs)

Basis:

Release rates in excess of ten times technical specifications which continue for 15 minutes or longer represent a serious situation. Ideally, most releases will begin small, then increase, hence will progress through the Unusual Event classification, allowing time to stop or mitigate them. Assuming this is the case, significant time has passed during which attempts to reduce or terminate the release have failed.

Therefore the required release duration for meeting this EAL was reduced to 15 minutes in recognition of the increased severity.

The final integrated dose (which is still expected to be low at these release rates) is not the primary concern here; it is the degradation in plant control implied by the fact that the release cannot be controlled.

Likewise, liquid release values (which would result in very low integrated dose) are not the primary concern. Rather, the fact that the release cannot be isolated represents a potential degradation in the level of safety.

References:

STPT 13.4, Effluent Monitors NUREG 0654, Appendix 1 Initiating Condition: Alert 15 Page 38 of 94 REFERENCE USE

POINT BEACH NUCLEAR PLANT EPIP 1.2 EMERGENCY PLAN IMPLEMENTING PROCEDURES NNSR Revision 37 EMERGENCY CLASSIFICATION November 15, 2002 ATTACHMENT B EMERGENCY ACTION LEVELS (EALs)

Category: Radiological Conditions EAL 4.1.1.3 Sub-Category: Off-site Radiological Release Emergency Classification: SITE EMERGENCY Emergency Action Level:

Release of airborne radioactivematerial resulting in a dose projection at or beyond the site boundary (by any means) of either:

Greaterthan or equal to 100 mRem Total Effective Dose Equivalent (TEDE)

OR Greaterthan or equal to 500 mRem Committed Dose Equivalent (CDE) [thyroid]

OR Closed window dose rate exceeding 100 mRem/hr measured at or beyond the site boundary.

Basis:

The 100 mR integrated dose is based on the 10 CFR 20 annual average population exposure. This value also provides a desirable gradient (one order of magnitude) between the Site Emergency and General Emergency classes. It is deemed that exposures less than this limit are not consistent with the Site Emergency class description. The 500 mR CDE thyroid dose was established in consideration of the 1:5 ratio of the EPA Protective Action Guidelines for whole body to thyroid.

Integrated doses are not monitored in real-time but are projected. In establishing the duration used for the projection, care should be exercised to ensure the time estimates are realistic. If no educated guess can be made regarding estimated duration, the default (4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />) shall be used.

References:

NUREG 0654, Appendix 1 Initiating Condition: Site Emergency 13a Page 39 of 94 REFERENCE USE

POINT BEACH NUCLEAR PLANT EPIP 1.2 EMERGENCY PLAN IMPLEMENTING PROCEDURES NNSR Revision 37 EMERGENCY CLASSIFICATION November 15, 2002 ATITACHMENT B EMERGENCY ACTION LEVELS (EALs)

Category: Radiological Conditions EAL 4.1.1.4 Sub-Category: Off-site Radiological Release Emergency Classification: GENERAL EMERGENCY Emergency Action Level:

Release of airborneradioactive materialresulting in a dose projection at or beyond the site boundary (by any means) of either:

Greaterthan or equal to 1 Rem Total Effective Dose Equivalent (TEDE)

OR Greaterthan or equal to 5 Rem Committed Dose Equivalent (CDE) [thyroid]

OR Closed window dose rate exceeding 1 Rem/hr measured at or beyond the site boundary.

Basis:

The I REM TEDE and the 5 REM CDE thyroid integrated doses are based on the EPA protective action guidance which indicates that public protective actions are indicated. This is consistent with the emergency class description for a General Emergency.

Integrated doses are not monitored in real-time but are projected. In establishing the duration used for the projection, care should be exercised to ensure the time estimates are realistic. If no educated guess can be made regarding estimated duration, the default (4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />) shall be used.

References:

NUREG 0654, Appendix 1 Initiating Condition: General Emergency la Page 40 of 94 REFERENCE USE

POINT BEACH NUCLEAR PLANT EPIP 1.2 EMERGENCY PLAN IMPLEMENTING PROCEDURES NNSR Revision 37 EMERGENCY CLASSIFICATION November 15, 2002 (EALs)

B LEVELS ATTACHMENT BACTION ATTACHMENT EMERGENCY EMERGENCY ACTION LEVELS (EALs)

Category: Radiological Conditions EAL 4.2.1.2 Sub-Category: In-Plant Radiological Conditions Emergency Classification: ALERT Emergency Action Level:

Loss of control of radioactivematerial resulting in area radiationexceeding I00X normal (or expected) levels within the ProtectedArea. Normal may be determined by trend recorderor other relevant data.

Basis:

By themselves indications of increased levels of radiation would only meet the Unusual Event class description (potential degradation in the level of safety). However, there is no specific Unusual Event EAL on increased radiation. This would be a judgment call by the SM. However, when increased radiation of this magnitude (1000x) is combined with "loss of control" a higher classification is warranted. Non-essential personnel should be assembled to ensure their safety. Additional manpower or other resources may be needed. The ALERT classification is appropriate.

The operative phrase in this EAL is "loss of control". Combined with this is the phrase "or expected levels". For most plant evolutions increases of radiation can be estimated, most within a factor of 1000.

If, in the judgment of those concerned, control has been lost, AND radiation levels increase beyond 100OX normal or expected levels, this EAL is met.

References:

NUREG 0654, Appendix 1 Initiating Condition: Alert 6 Page 41 of 94 REFERENCE USE

POINT BEACH NUCLEAR PLANT EPIP 1.2 EMERGENCY PLAN IMPLEMENTING PROCEDURES NNSR Revision 37 EMERGENCY CLASSIFICATION November 15, 2002 (EALs)

B LEVELS ATTACHMENT BACTION ATTACHMENT EMERGENCY EMERGENCY ACTION LEVELS (EALs)

Category: Internal Events EAL 5.1.1.1 Sub-Category: Security Threats Emergency Classification: UNUSUAL EVENT Emergency Action Level:

Bomb, credible bomb threat, indication of sabotage, or attempted entry into the ProtectedArea by a hostile force.

Basis:

This EAL is based on the PBNP Security Plan/ISFSI Security Plan. An actual bomb, credible bomb threat, act of sabotage, or attempted entry into the Protected area by a hostile force indicates a potential degradation in the level of safety at the plant. Therefore an Unusual Event classification is warranted.

The Protected Area Physical Barrier is defined in the Security Plan/ISFSI Plan.

A bomb discovered in or near a Plant Vital Area which could affect Safety-Related Functions would result in escalation of the emergency classification. An actual explosion (of a bomb or other source) would be classified based on EALs 5.3.1.1 through 5.3.1.3 depending upon its effects.

Security events that do not represent at least a potential degradation in the level of plant safety are reported under either 10 CFR 73.71 or 10 CFR 50.72 and do not require implementation of the Emergency Plan. Accidental, non-hostile entry, although reportable as a security event, does not warrant declaration of an.emergency. The operative consideration is 'intent'. If no malicious intent is determined the EAL does not apply.

References:

SSCP - Security and Safeguards Contingency Plan NUREG 0654, Appendix 1 Initiating Condition: Unusual Event 12 Page 42 of 94 REFERENCE USE

POINT BEACH NUCLEAR PLANT EPIP 1.2 EMERGENCY PLAN IMPLEMENTING PROCEDURES NNSR Revision 37 EMERGENCY CLASSIFICATION November 15, 2002 (EALs)

B LEVELS ATTACHMENT ATTACHMENTEMERGENCYBACTION EMERGENCY ACTION LEVELS (EALs)

Category: Internal Events EAL 5.1.1.2 Sub-Category: Security Threats Emergency Classification: ALERT Emergency Action Level:

I Intrusion into the ProtectedArea by a hostileforce. I Basis:

For the purposes of this EAL, the intrusion into the Protected Area can be considered a significant security threat. An Alert classification is warranted. If entry is attempted, but not gained by a hostile force see the Unusual Event EAL.

The Protected Area Physical Barrier is defined in the Security Plan. Note: The Independent Spent Fuel Storage Installation (ISFSI) is a separate Protected Area.

Intrusion into a Plant Vital Area escalates this event to a Site Emergency.

Security events that do not represent at least a potential degradation in the level of plant safety are reported under either 10 CFR 73.71 or 10 CFR 50.72 and do not require implementation of the Emergency Plan. Accidental, non-hostile entry, although reportable as a security event, does not warrant declaration of an emergency. The operative consideration is 'intent'. If no malicious intent is determined the EAL does not apply.

References:

SSCP - Security and Safeguards Contingency Plan NUREG 0654, Appendix 1 Initiating Condition: Alert 16 Pagre 43 of 94 REFERENCE USE

POINT BEACH NUCLEAR PLANT EPIP 1.2 EMERGENCY PLAN IMPLEMENTING PROCEDURES NNSR Revision 37 EMERGENCY CLASSIFICATION November 15, 2002 (EALs)

B LEVELS ATTACHMENT BACTION ATTACHMENT EMERGENCY EMERGENCY ACTION LEVELS (EALs)

Category: Internal Events EAL 5.1.1.3 Sub-Category: Security Threats Emergency Classification: SITE EMERGENCY Emergency Action Level:

Intrusion into a plant Vital Area by hostileforce.

Basis:

Hostile takeover of Vital Areas could lead to loss of physical control of the plant. Therefore a Site Emergency classification is warranted. The Plant Vital Areas are defined in the Security Plan.

Security events that do not represent at least a potential degradation in the level of plant safety are reported under either 10 CFR 73.71 or 10 CFR 50.72 and do not require implementation of the Emergency Plan. Accidental, non-hostile entry, although reportable as a security event, does not warrant declaration of an emergency. The operative consideration is 'intent'. If no malicious intent is determined the EAL does not apply.

References:

SSCP - Security and Safeguards Contingency Plan NUREG 0654, Appendix 1 Initiating Condition: Site Emergency 14 Page 44 of 94 REFERENCE USE

POINT BEACH NUCLEAR PLANT EPIP 1.2 EMERGENCY PLAN IMPLEMENTING PROCEDURES NNSR Revision 37 EMERGENCY CLASSIFICATION November 15, 2002 ATTACHMENT B EMERGENCY ACTION LEVELS (EALs)

Category: Internal Events EAL 5.1.1.4 Sub-Category: Security Threats Emergency Classification: GENERAL EMERGENCY Emergency Action Level:

A Security Event which results in either:

Loss of physical control of the ControlRoom OR Loss of remote shutdown capability.

Basis:

This EAL encompasses conditions under which unauthorized personnel have taken physical control of vital areas required to reach and maintain safe shutdown, with the potential that the intruders can cause a significant event with damage to plant systems, damage to the core, and ultimately a release of large amounts of radioactivity.

References:

SSCP - Security and Safeguards Contingency Plan AOP-10A, Safe Shutdown - Local Control NUREG 0654, Appendix 1 Initiating Condition: General Emergency 3 Page 45 of 94 REFERENCE USE

POINT BEACH NUCLEAR PLANT EPIP 1.2 EMERGENCY PLAN IMPLEMENTING PROCEDURES NNSR Revision 37 EMERGENCY CLASSIFICATION November 15, 2002 ATTACHMENT B EMERGENCY ACTION LEVELS (EALs)

Category: Internal Events EAL 5.2.1.2 Sub-Category: Control Room Habitability Emergency Classification: ALERT Emergency Action Level:

Evacuation of the ControlRoom has been initiatedwith control of shutdown systems establishedfrom local stations.

Basis:

AOP-1OA directs shutdown activities performed outside the Control Room.

This EAL does not imply that all actions associated with Alternate Shutdown shall be completed in order to avoid the higher EAL pertaining to Control Room evacuation (EAL 5.2.1.3). If the reactor successfully trips, if level, pressure, temperature, etc., are being controlled, and no impediments to the associated Shutdown activities are being encountered, this emergency classification is appropriate. If impediments are being encountered in completing critical Shutdown functions, and more than 15 minutes expire, EAL 5.2.1.3 is met.

Located within the Control Room are the controls, indications, annunciators, and communications equipment necessary for the safe operation of the plant. The ability to assess and control plant conditions and abnormal situations is significantly degraded without access to the Control Room.

With the Control Room evacuated, additional support, monitoring, and direction through the resources of the TSC and/or other emergency facilities is assumed to be necessary - therefore, the dedlaration of an Alert is appropriate and required.

References:

AOP-10A, Safe Shutdown - Local Control NUREG 0654, Appendix 1 Initiating Condition: Alert 20 Page 46 of 94 REFERENCE USE

POINT BEACH NUCLEAR PLANT EPIP 1.2 EMERGENCY PLAN IMPLEMENTING PROCEDURES NNSR Revision 37 EMERGENCY CLASSIFICATION November 15, 2002 ATTACHMENT B EMERGENCY ACTION LEVELS (EALs)

Category: Internal Events EAL 5.2.1.3 Sub-Category: Control Room Habitability Emergency Classification: SITE EMERGENCY Emergency Action Level:

Evacuation of the Control Room without establishmentof plant controlfrom remote shutdown stations within approximately 15 minutes.

Basis:

Located within the Control Room are the controls, indications, annunciators, and communications equipment necessary for the safe operation of the plant. The ability to assess and control plant conditions and abnormal situations is significantly degraded without access to the Control Room.

Once the Control Room is evacuated, if control is not established from remote shutdown stations within a reasonable amount of time (approximately 15 minutes), a significant threat to multiple fission product barriers exists should a plant transient or other emergency condition occur. If plant control cannot be established within this time frame, declaration of a Site Emergency is warranted due to extended lack of control of the plant.

Escalation to a higher classification, if appropriate, will be based on system malfunctions, fission product barrier degradation, radiation levels, or Emergency Director judgment.

References:

AOP-OA, Safe Shutdown - Local Control NUREG 0654, Appendix 1 Initiating Condition: Site Emergency 18 Page 47 of 94 REFERENCE USE

POINT BEACH NUCLEAR PLANT EPIP 1.2 EMERGENCY PLAN IMPLEMENTING PROCEDURES NNSR Revision 37 EMERGENCY CLASSIFICATION November 15, 2002 (EALs)

B LEVELS ATTACHMENT BACTION ATTACHMENT EMERGENCY EMERGENCY ACTION LEVELS (EALs)

Category: Internal Events EAL 5.3.1.1 Sub-Category: Fire / Explosion Emergency Classification: UNUSUAL EVENT Emergency Action Level:

Explosion near or in the ProtectedArea affecting permanentplant equipment or structures.

Basis:

Only those explosions of sufficient force to damage permanent plant structures or equipment within the Protected Area meet the threshold of this EAL. Other minor explosions should be evaluated for reportability 10 CFR 50.72. No attempt is made in this EAL to assess the magnitude of damage.

Reports of any damage to permanent structures or equipment is sufficient for declaration.

Specifically excluded from this EAL are stored equipment or non-permanent structures such as trailers, Sea Vans, or skids which are not being relied upon to perform the function of installed plant equipment.

The Protected Area Physical Barrier is defined in the Security Plan. Note: The Independent Spent Fuel Storage Installation (ISFSI) is a separate Protected Area.

As used here, an explosion is a rapid, violent, unconfined combustion or a catastrophic failure of pressurized equipment imparting significant energy to nearby structures and materials. If the explosion damages Safety Systems the event escalates to an Alert or Site Emergency.

The security aspects of the explosion should be considered.

References:

NUREG 0654, Appendix 1 Initiating Condition: Unusual Event 14c SSCP - Security and Safeguards Contingency Plan Page 48 of 94 REFERENCE USE

POINT BEACH NUCLEAR PLANT EPIP 1.2 EMERGENCY PLAN IMPLEMENTING PROCEDURES NNSR Revision 37 EMERGENCY CLASSIFICATION November 15, 2002 (EALs)

B LEVELS ATTACHMENT BACTION ATTACHMENTEMERGENCY EMERGENCY ACTION LEVELS (EALs)

Category: Internal Events EAL 5.3.1.2 Sub-Category: Fire / Explosion Emergency Classification: ALERT Emergency Action Level:

I Explosion affecting operabilityof one (1) train of safety systems. I Basis:

Safety systems as used here designates systems with safety-related functions. Attachment D lists safety systems and systems with safety-related functions.

Only explosions that actually cause damage to equipment required for safe operation AND only damage that renders a single train of a safety system unable to perform its intended safety function meet the threshold of this EAL. A lengthy damage assessment should not be performed. The occurrence of the explosion with evidence of damage likely to prevent one train from performing its intended safety function is sufficient for declaration.

As used here, an explosion is a rapid, violent, unconfined combustion, or catastrophic failure of pressurized equipment that imparts significant energy to nearby structures and equipment.

If the explosion damages more than one train of a Safety System the event escalates to a Site Emergency.

The security aspects of the explosion should be considered.

References:

NUREG 0654, Appendix 1 Initiating Condition: Alert 18c Page 49 of 94 REFERENCE USE

POINT BEACH NUCLEAR PLANT EPIP 1.2 EMERGENCY PLAN IMPLEMENTING PROCEDURES NNSR Revision 37 EMERGENCY CLASSIFICATION November 15, 2002 ATTACHMENT B EMERGENCY ACTION LEVELS (EALs)

Category: Internal Events EAL 5.3.1.3 Sub-Category: Fire / Explosion Emergency Classification: Site Emergency Emergency Action Level:

Explosion affecting operabilityof two (2) trains of safety systems.

Basis:

Safety systems as used here designates systems with safety-related functions. Attachment D lists safety systems and systems with safety-related functions.

Only explosions that actually cause damage to equipment required for safe operation of more than one safety system train AND only damage that affects the systems' ability to perform intended functions meet the threshold of this EAL. A lengthy damage assessment should not be performed. An immediate assessment of the probability of damage making multiple trains incapable of performing their safety function is all that is required. The occurrence of the explosion with evidence of damage likely to prevent the equipment in more than one train of a safety system from performing intended safety functions is sufficient for declaration.

As used here, an explosion is a rapid, violent, unconfined combustion, or catastrophic failure of pressurized equipment that imparts significant energy to nearby structures and equipment.

If only one train of a safety system is affected, see ALERT classification EAL.

The security aspects of the explosion should be considered.

References:

NUREG 0654, Appendix 1 Initiating Condition: Alert 18c Page 50 of 94 REFERENCE USE

POINT BEACH NUCLEAR PLANT EPIP 1.2 EMERGENCY PLAN IMPLEMENTING PROCEDURES NNSR Revision 37 EMERGENCY CLASSIFICATION November 15, 2002 (EALs)

B LEVELS ATTACHMENT BACTION ATTACHMENTEMERGENCY EMERGENCY ACTION LEVELS (EALs)

Category: Internal Events EAL 5.3.2.1 Sub-Category: Fire / Explosion Emergency Classification: UNUSUAL EVENT Emergency Action Level:

Fire within the ProtectedArea lasting more than 10 minutes after use offire extinguishing equipment.

Basis:

The purpose of this EAL is to address fires which are potentially significant precursors to damage to safety systems. This condition applies to buildings or areas contiguous to plant vital areas or other significant buildings or areas.

Specifically excluded are small fires within administration buildings, wastebasket fires, or fires in areas of no safety consequence.

Escalation to a higher emergency class occurs if the fire affects one or more train(s) of a Safety System(s).

References:

NUREG 0654, Appendix 1 Initiating Condition: Unusual Event 10 Page 51 of 94 REFERENCE USE

POINT BEACH NUCLEAR PLANT EPIP 1.2 EMERGENCY PLAN IMPLEMENTING PROCEDURES NNSR Revision 37 EMERGENCY CLASSIFICATION November 15, 2002 ATT7ACHMENT B EMERGENCY ACTION LEVELS (EALs)

Category: Internal Events EAL 5.3.2.2 Sub-Category: Fire / Explosion Emergency Classification: ALERT Emergency Action Level:

Fire affecting operabilityof one (1) train of a safety system.

Basis:

Safety systems as used here designates systems with safety-related functions. Attachment D lists safety systems and systems with safety-related functions.

This condition is entered when the Fire Brigade Leader reports a fire affects one train of a safety system or if Control Room Operators become aware of indications of impact to a safety system after a fire has been reported.

Only those fires that actually cause damage to equipment as reported by the Fire Brigade Leader or as noted by Control Room operators meet this EAL.

Escalation to a higher emergency class, if appropriate, is based on further system malfunctions, fission product barrier degradation, abnormal radiation levels, or Emergency Director judgment.

References:

NUREG 0654, Appendix I Initiating Condition: Alert 13 Page 52 of 94 REFERENCE USE

POINT BEACH NUCLEAR PLANT EPIP 1.2 EMERGENCY PLAN IMPLEMENTING PROCEDURES NNSR Revision 37 EMERGENCY CLASSIFICATION November 15, 2002 ATTACHMENT B EMERGENCY ACTION LEVELS (EALs)

Category: Internal Events EAL 5.3.2.3 Sub-Category: Fire / Explosion Emergency Classification: SITE EMERGENCY Emergency Action Level:

Fire affecting operability of tvo (2) trains of safety systems.

Basis:

Safety systems as used here designates systems with safety-related functions. Attachment D lists safety systems and systems with safety-related functions.

This condition is entered when the Fire Brigade Leader reports a fire that affects more than one train of a safety system or if Control Room Operators become aware of indications of impact on more than one train of a safety system after a fire has been reported.

Only fires that actually cause damage to equipment required for safe operation of more than one safety system train AND only damage that affects the systems' ability to perform intended functions meet the threshold of this EAL. A lengthy damage assessment should not be performed. An immediate assessment of the probability of damage making multiple trains incapable of performing their safety function is all that is required. The occurrence of a fire with evidence of damage likely to prevent the equipment in more than one train of a safety system from performing intended safety functions is sufficient for declaration.

This condition is indicative of severe degradation of the level of safety at the plant with possible adverse consequences on the public health and safety. A Site Emergency is warranted.

Escalation to a higher emergency class, if appropriate, will be based on fission product barrier degradation or emergency management judgment.

References:

NUREG 0654, Appendix 1 Initiating Condition: Site Emergency 11 Page 53 of 94 REFERENCE USE

POINT BEACH NUCLEAR PLANT EPIP 1.2 EMERGENCY PLAN IMPLEMENTING PROCEDURES NNSR Revision 37 EMERGENCY CLASSIFICATION November 15, 2002 ATTACHMENT B EMERGENCY ACTION LEVELS (EALs)

Category: Internal Events EAL 5.4.1.1 Sub-Category: Turbine Rotating Component Failures Emergency Classification: UNUSUAL EVENT Emergency Action Level:

Visual confirmation of turbine housing penetrationby a blade or rotating component.

Basis:

This initiation condition addresses the consequences of turbine failure and turbine missile effects.

Analyses documented in the FSAR on the consequences of turbine overspeed indicate that there would be only a low energy missile generated external to the low pressure turbine casing in the event of a turbine overspeed.

The study determined that the following components are subject to the possible effects of a turbine missile: one main steam line, the condensate storage tanks, reactor makeup water storage tanks, the reactor makeup water storage tank pumps, the refueling water storage tank, diesel generator fuel oil line, and the service water pump electrical leads. These components should be evaluated for damage.

Escalation to a higher emergency classification, if appropriate, is based on further missile damage from any source, system malfunctions, fission product barrier degradation, abnormal radiation levels, or emergency management judgment.

References:

WCAP 7525-L, Likelihood and Consequences of Turbine Overspeed at the Point Beach Nuclear Plant.

Reg Guide 1.115, Protection Against Low-Trajectory Turbine Missiles FSAR 14.1.12, Likelihood of Turbine-Generator Unit Overspeed NUREG 0654, Appendix 1 Initiating Condition: Unusual Event 14e and Alert 1Be Page 54 of 94 REFERENCE USE

POINT BEACH NUCLEAR PLANT EPIP 1.2 EMERGENCY PLAN IMPLEMENTING PROCEDURES NNSR Revision 37 EMERGENCY CLASSIFICATION November 15, 2002 ATTACHMENT B EMERGENCY ACTION LEVELS (EALs)

Category: External Events EAL 6.1.1.1 Sub-Category: Natural Destructive Phenomena Emergency Classification: UNUSUAL EVENT Emergency Action Level:

Any earthquakefelt by Control Room Operators.

OR An indicatorlight on two or more of the following Seismic Event Monitors SEI-6210 #3 Warehouse SEI-6211 Unit I Facade SEI-6212 Drum Prep Room SEI-6213 El. 8 "between vital switchgear room and autxfeedwater tunnel Basis:

As defined in the EPRI sponsored "Guidelines for Nuclear Plant Response to an Earthquake," dated October 1989, a "felt earthquake" is:

An earthquake of sufficient intensity such that: (a) the ground motion is felt at the nuclear plant site and recognized as an earthquake based on a consensus of control room operators on duty at the time, and (b) for plants with operable seismic instrumentation, the seismic detectors of the plant are activated.

The seismic event monitors are set to alarm at 0.01g. Minor damage to some portions of the site may occur at these levels but should not affect the ability to safely operate the plant. Additional inspections may be desired to determine the extent of any damage. Therefore an Unusual Event classification is warranted.

This EAL requires two valid seismic alarms to eliminate classification due to plant operations or maintenance activities, such as heavy equipment moving near the monitor or an accidental impact to a monitor. Further validation may be accomplished by contacting the University of Wisconsin - Milwaukee Seismic Center.

Page 55 of 94 REFERENCE USE

POINT BEACH NUCLEAR PLANT EPIP 1.2 EMERGENCY PLAN IMPLEMENTING PROCEDURES NNSR Revision 37 EMERGENCY CLASSIFICATION November 15, 2002 ATTACHMENT B EMERGENCY ACTION LEVELS (EALs)

References:

PBNP FSAR, Appendix A Setpoint Document STPT 22. 1, Seismic Event Monitoring EPRI Document, "Guidelines for Nuclear Plant Response to an Earthquake," dated October, 1989 NUREG 0654, Appendix 1 Initiating Condition: Unusual Event 13a Page 56 of 94 REFERENCE USE

POINT BEACH NUCLEAR PLANT EPIP 1.2 EMERGENCY PLAN IMPLEMENTING PROCEDURES NNSR Revision 37 EMERGENCY CLASSIFICATION November 15, 2002 ATTACHMENT B EMERGENCY ACTION LEVELS (EALs)

Category: External Events EAL 6.1.1.2 Sub-Category: Natural Destructive Phenomena Emergency Classification: ALERT Emergency Action Level:

I Valid Seismic Event Monitor readings of an intensity greaterthan O.04g vertical or O.06g horizontal.-j Basis:

This EAL addresses events that may have resulted in the plant's vital equipment being subjected to forces beyond operational limits. Therefore an Alert classification is warranted. Classification should occur prior to a detailed damage assessment.

Values in this EAL are based on the Operating Basis Earthquake (OBE) limits (ground accelerations of

.04g vertical and .06g horizontal) as defined by the FSAR.

Validation of seismic activity would be by severe ground shaking or by contacting University of Wisconsin - Milwaukee Seismic Center (Emergency Telephone Directory).

References:

PBNP FSAR, Appendix A Setpoint Document STPT 22.1, Seismic Event Monitoring EPRI Document, "Guidelines for Nuclear Plant Response to an Earthquake," dated October 1989 NUREG 0654, Appendix 1 Initiating Condition: Alert 17a Page 57 of 94 REFERENCE USE

POINT BEACH NUCLEAR PLANT EPIP 1.2 EMERGENCY PLAN IMPLEMENTING PROCEDURES NNSR Revision 37 EMERGENCY CLASSIFICATION November 15, 2002 ATTACHMENT B EMERGENCY ACTION LEVELS (EALs)

Category: External Events EAL 6.1.1.3 Sub-Category: Natural Destructive Phenomena Emergency Classification: SITE EMERGENCY Emergency Action Level:

I Valid Seismic Event Monitor readingsof an intensity greaterthan O.08g vertical or O.12g horizontal. I Basis:

This EAL addresses events that may have resulted in the plant's vital equipment being subject to forces that may prevent safe shutdown and cooldown of the plant. Therefore a Site Emergency classification is warranted. Classification should occur prior to a detailed damage assessment.

Values in this EAL are based on the Safe Shutdown Earthquake (SSE) limits (ground accelerations of

.08g vertical and .12g horizontal) as defined by the FSAR.

Validation of seismic activity would be by severe ground shaking or by contacting University of Wisconsin - Milwaukee Seismic Center (Emergency Telephone Directory).

References:

PBNP FSAR, Appendix A Setpoint Document STPT 22.1, Seismic Event Monitoring EPRI Document, "Guidelines for Nuclear Plant Response to an Earthquake," dated October 1989 NUREG 0654, Appendix 1 Initiating Condition: Site Emergency 15a Pagre 58 of 94 REFERENCE USE

POINT BEACH NUCLEAR PLANT EPIP 1.2 EMERGENCY PLAN IMPLEMENTING PROCEDURES NNSR Revision 37 EMERGENCY CLASSIFICATION November 15, 2002 (EALs)

B LEVELS ATTACHMENT BACTION ATT'ACHMENTEMERGENCY EMERGENCY ACTION LEVELS (EALs)

Category: External Events EAL 6.1.2.1 Sub-Category: Natural Destructive Phenomena Emergency Classification: UNUSUAL EVENT Emergency Action Level:

I Report by plant personnel that a tornado is within, or has been in, the ProtectedArea or Switchyard. I~

Basis:

This EAL is based on the assumption that a tornado within the Protected Area may potentially damage plant structures containing functions or systems required for the safe shutdown of the plant. Due to the rapid transient nature of tornadoes only those which actually occur within the Protected Area are considered, as these have the potential of actually damaging plant structures. An Unusual Event classification is warranted.

If damage to safety-related equipment is confirmed (either by observation or plant instrumentation) the event may be escalated to an Alert. Other EALs should also be considered such as loss of electrical power.

The Protected Area Boundary is defined in the Security Plan. Note: The Independent Spent Fuel Storage Installation (ISFSI) is part of the site Protected Area.

References:

AOP- 13C, Severe Weather Conditions Probabilistic Safety Assessment -- High Winds, and Others Sec 9, Rev 0, Dated July 1995 NUREG 0654, Appendix 1 Initiating Condition: Unusual Event 13c SSCP - Security and Safeguards Contingency Plan Page 59 of 94 REFERENCE USE

POINT BEACH NUCLEAR PLANT EPIP 1.2 EMERGENCY PLAN IMPLEMENTING PROCEDURES NNSR Revision 37 EMERGENCY CLASSIFICATION November 15, 2002 ATTACHMENT B EMERGENCY ACTION LEVELS (EALs)

Category: External Events EAL 6.1.2.2 Sub-Category: Natural Destructive Phenomena Emergency Classification: ALERT Emergency Action Level:

Indications or observations that a tornado has damaged a vital structure OR Sustained winds greaterthan 90 MPH.

Basis:

This EAL addresses events that may have resulted in a plant vital area being subjected to forces approaching or beyond design limits. It is assumed that damage may have occurred to plant safety systems. Therefore an Alert classification is warranted. Classification should occur prior to a detailed damage assessment.

The 90 MPH sustained wind speed was chosen as a value approaching the design basis for non-Class 1 metal structures at the plant. Although no damage to permanent plant structures should occur at this level, non-permanent structures (trailers, work shacks, temporary storage, etc.) could have significant damage and impact plant operations. Winds at this level would also impact personnel movement within and to the plant.

References:

AOP-13C, Severe Weather Conditions FSAR 5.1, Containment System Structure Probabilistic Safety Assessment -- High Winds, and Others Sec 9, Rev 0, Dated July 1995 Bechtel Corporation, "Westinghouse Electric Corporation--Wisconsin Michigan Power Company--Point Beach Atomic Power Station--Design Criteria for Nuclear Power Plants Against Tornadoes,"

March 12, 1970, B-TOP-3.

NUREG 0654, Appendix 1 Initiating Condition: Alert 17c Page 60 of 94 REFERENCE USE

POINT BEACH NUCLEAR PLANT EPIP 1.2 EMERGENCY PLAN IMPLEMENTING PROCEDURES NNSR Revision 37 EMERGENCY CLASSIFICATION November 15, 2002 (EALs)

B LEVELS ATTACHMENT BACTION ATTACHMENT EMERGENCY EMERGENCY ACTION LEVELS (EALs)

Category: External Events EAL 6.1.2.3 Sub-Category: Natural Destructive Phenomena Emergency Classification: SITE EMERGENCY Emergency Action Level:

Sustained winds greater than 100 MPH AND Reports or indications of damage to vital equipment or structures.

Basis:

This EAL addresses events that have resulted in plant areas being subjected to forces beyond design limits. It is assumed that substantial damage has occurred to plant structures with probable damage to safety systems.

It is inferred from Section 5.1 in the FSAR that the design straight wind speed of 108 mph was used in the design of the non-Class 1 metal structures. This is consistent with the Bechtel topical report.

100 mph was used in this EAL due to limitations of available instrumentation.

Therefore, this condition is indicative of serious plant system conditions with possible adverse consequences on the public health and safety. A Site Emergency is warranted.

Emergency classifications under other EALs may also be appropriate due to offsite effects caused by high winds, particularly status of offsite power lines.

References:

AOP-13C, Severe Weather Conditions FSAR 5.1, Containment System Structure Probabilistic Safety Assessment -- High Winds, and Others Sec 9, Rev 0, Dated July 1995 Bechtel Corporation, "Westinghouse Electric Corporation--Wisconsin Michigan Power Company--Point Beach Atomic Power Station--Design Criteria for Nuclear Power Plants Against Tornadoes,"

March 12, 1970, B-TOP-3.

NUREG 0654, Appendix I Initiating Condition: Site Emergency 15c Page 61 of 94 REFERENCE USE

POINT BEACH NUCLEAR PLANT EPIP 1.2 EMERGENCY PLAN IMPLEMENTING PROCEDURES NNSR Revision 37 EMERGENCY CLASSIFICATION November 15, 2002 (EALs)

B LEVELS ATTACHMENT ATTACHMENTEMERGENCYBACTION EMERGENCY ACTION LEVELS (EALs)

Category: External Events EAL 6.2.1.1 Sub-Category: High Lake/Low Forebay Water Level Emergency Classification: UNUSUAL EVENT Emergency Action Level:

I Less than -]1 "forebayor pump bay level with one unit's CWpurmps off. II Basis:

This condition is considered a potential degradation in the level of safety of the plant due to Circulating Water Pumps and/or Service Water losing suction. Water levels at or below these levels impairs the ability of these pumps to provide water to their loads, and may result in subsequent loss of the safety function of the ultimate heat sink. Therefore, an Unusual Event classification is warranted.

References:

AOP-5A, Loss of Condenser Vacuum AOP-13A, Circulating Water System Malfunction NUREG 0654, Appendix I Initiating Condition: Unusual Event 13b Page 62 of 94 REFERENCE USE

POINT BEACH NUCLEAR PLANT EPIP 1.2 EMERGENCY PLAN IMPLEMENTING PROCEDURES NNSR Revision 37 EMERGENCY CLASSIFICATION November 15, 2002 ATTACHMENT B EMERGENCY ACTION LEVELS (EALs)

Category: External Events EAL 6.2.2.1 Sub-Category: High Lake/Low Forebay Water Level Emergency Classification: UNUSUAL EVENT Emergency Action Level:

Anyflooding which precludes access to the site or areas of the plant.

Basis:

This condition is considered to be a potential degradation in the level of safety of the plant due to limited access to the site or potential safety concerns for onsite personnel. Therefore an Unusual Event classification is warranted.

References:

NUREG 0654, Appendix 1 Initiating Condition: Unusual Event 13b Page 63 of 94 REFERENCE USE

POINT BEACH NUCLEAR PLANT EPIP 1.2 EMERGENCY PLAN IMPLEMENTING PROCEDURES NNSR Revision 37 EMERGENCY CLASSIFICATION November 15, 2002 ATTACHMENT B EMERGENCY ACTION LEVELS (EALs)

Category: External Events EAL 6.2.2.2 Sub-Category: High Lake/Low Forebay Water Level Emergency Classification: ALERT Emergency Action Level:

I Floodingas indicated by greaterthan 6" of water in the 8 foot elevation of the Turbine Bldg. I Basis:

This EAL addresses an event that may result in a plant vital area being subjected to conditions beyond design limits adversely affecting plant safety systems. Therefore, this condition is indicative of abnormal plant conditions with possible adverse consequences on plant safety and is classified as an Alert.

This condition is entered when there is greater than six inches of water in the turbine hall. Although this EAL is in the category High Lake/Low Forebay, the cause of the flooding is not a factor. A broken Service Water or Circulating Water pipe could also create this condition.

The Turbine Building would flood before other plant areas, therefore it provides a representative indication of other possible problem areas. The feedwater pumps each sit on a base that is eight inches above the floor. The turbine seal oil pumps are approximately ten inches above the floor.

Escalation to a higher emergency class, if appropriate, will be based on Flooding in Vital Equipment Areas.

References:

NUREG 0654, Appendix I Initiating Condition: Alert 17b Page 64 of 94 REFERENCE USE

POINT BEACH NUCLEAR PLANT EPIP 1.2 EMERGENCY PLAN IMPLEMENTING PROCEDURES NNSR Revision 37 EMERGENCY CLASSIFICATION November 15, 2002 ATTACHMENT B EMERGENCY ACTION LEVELS (EALs)

Category: External Events EAL 6.2.2.3 Sub-Category: High Lake/Low Forebay Water Level Emergency Classification: SITE EMERGENCY Emergency Action Level:

Greaterthan 2 "ofwater in vital switchgearroom OR Greaterthan 2 *ofwater in auxiliaryfeedwater pump room.

Basis:

This EAL addresses conditions where plant vital equipment may be subjected to conditions beyond design limits, and damage may be assumed to have occurred to plant safety systems. Therefore, this condition is indicative of serious plant system conditions with possible adverse consequences on the public health and safety. A Site Emergency is warranted.

Plant vital area designations are contained in the PBNP Security Plan.

Water levels in excess of two feet in the vital switchgear room severely threaten safe plant operations.

Several 125-volt DC station batteries are installed in the vital switchgear room. The bottom and top of these batteries are 6 and 36 inches above the floor, respectively. Numerous electrical cabinets containing electrical components for the safety injection pumps, the station service transformers, and the 4.16 kV electrical system are also located in the room.

Water levels in excess of two feet in the auxiliary feedwater pump room threatens operation of the feedwater system and ultimately the ability to cool the reactor core. The turbine-operated auxiliary feedwater pumps are located approximately 18 inches above the floor and the motor operated auxiliary feedwater pumps are located approximately two feet above the floor. Additionally, the Source Range Output Expansion Control Panel is approximately two feet above the floor.

This EAL used to also contain criteria of greater than three feet of water in both EDG rooms, however this was before G03 and G04 were installed, hence spoke of G01 and G02 only. Due to the electrical arrangement of G03 and G04 as backups to G01 and G02 and the fact that G03 and G04 are at a significantly higher elevation, they have been removed from this EAL.

Emergency classifications under other EALs may be appropriate due to offsite effects caused by severe weather, particularly the status of offsite power lines.

Page 65 of 94 REFERENCE USE

POINT BEACH NUCLEAR PLANT EPIP 1.2 EMERGENCY PLAN IMPLEMENTING PROCEDURES NNSR Revision 37 EMERGENCY CLASSIFICATION November 15, 2002 ATTACHMENT B EMERGENCY ACTION LEVELS (EALs)

References:

SOER 85-5, Internal Flooding of Power Plant Buildings NUREG 0654, Appendix 1 Initiating Condition: Site Emergency 15b Page 66 of 94 REFERENCE USE

POINT BEACH NUCLEAR PLANT EPIP 1.2 EMERGENCY PLAN IMPLEMENTING PROCEDURES NNSR Revision 37 EMERGENCY CLASSIFICATION November 15, 2002 ATTACHMENT B EMERGENCY ACTION LEVELS (EALs)

Category: External Events EAL 6.3.1.1 Sub-Category: Toxic/Flammable Gas Intrusion Emergency Classification: UNUSUAL EVENT Emergency Action Level:

A toxic orflammable gas release in or near the ProtectedArea.

Basis:

The release of toxic or flammable gas in or near the Protected Area may pose a potential threat to reactor plant and personnel safety. It is the potential threat to normal operation or hazard to personnel which must be evaluated. If no such threat exists, the EAL is not met. If, however, personnel safety or plant operation is threatened, an Unusual Event is warranted.

For the purpose of this EAL, 'in or near' is considered to include those owner-controlled areas immediately surrounding the Protected Area and all areas within the Protected Area.

Flammable gases are typically more limiting than toxic gases. Although an SCBA could protect from toxicity, detonation of a flammable gas could be immediately hazardous to personnel.

The Plant Protected Area Boundary is defined in the Security Plan. Note: The Independent Spent Fuel Storage Installation (ISFSI) is part of the site Protected Area.

References:

NUREG 0654, Appendix 1 Initiating Condition: Unusual Event 14d SSCP Security and Safeguards Contingency Plan Page 67 of 94 REFERENCE USE

POINT BEACH NUCLEAR PLANT EPIP 1.2 EMERGENCY PLAN IMPLEMENTING PROCEDURES NNSR Revision 37 EMERGENCY CLASSIFICATION November 15, 2002 (EALs)

B LEVELS ATI'ACHMENT BACTION ATTACHMENT EMERGENCY EMERGENCY ACTION LEVELS (EALs)

Category: External Events EAL 6.3.1.2 Sub-Category: Toxic/Flammable Gas Intrusion Emergency Classification: ALERT Emergency Action Level:

Entry of toxic orflammable gas into a plant building atmosphere affecting operation or access.

Basis:

The release of toxic or flammable gas significant enough to affect plant operation (i.e., initiate a plant transient or preclude access to plant equipment) warrants declaration of an Alert:

Flammable gases are typically more limiting than toxic gases. Although an SCBA could protect from toxicity, detonation of a flammable gas could be immediately hazardous to personnel. An area where access is not required for plant operation, which could be evacuated, does not warrant an Alert, but may warrant an Unusual Event if the potential exists to affect operation or personnel.

Any affected area normally accessed for plant operation (PAB, Turbine hall, etc.) meets the Alert level.

If vital areas are affected, see EAL 6.3.1.3.

The primary flammable gases considered are acetylene, propane and the hydrogen.

References:

NUREG 0654, Appendix 1 Initiating Condition: Alert 18d Page 68 of 94 REFERENCE USE

POINT BEACH NUCLEAR PLANT EPIP 1.2 EMERGENCY PLAN IMPLEMENTING PROCEDURES NNSR Revision 37 EMERGENCY CLASSIFICATION November 15, 2002 ATTACHMENT B EMERGENCY ACTION LEVELS (EALs)

Category: External Events EAL 6.3.1.3 Sub-Category: Toxic/Flammable Gas Intrusion Emergency Classification: SITE EMERGENCY Emergency Action Level:

Entry of toxic orflammable gas into a plant vital area affecting operation or personnel safety AND Reactor coolant temperature greater than 200 OF.

Basis:

The release of toxic or flammable gas into a plant vital area poses a significant threat to plant safety by precluding access to plant vital equipment which may be needed for Safe Shutdown. Therefore this condition warrants declaration of a Site Emergency.

Flammable gases are typically more limiting than toxic gases. Although an SCBA could protect from toxicity, detonation of a flammable gas could be immediately hazardous to personnel.

This EAL does not apply in cold shutdown or refueling modes due to the significantly reduced probability that the loss of access would result in fuel failure and/or a release.

References:

NUREG 0654, Appendix 1 Initiating Condition: Site Emergency 16c Page 69 of 94 REFERENCE USE

POINT BEACH NUCLEAR PLANT EPIP 1.2 EMERGENCY PLAN IMPLEMENTING PROCEDURES NNSR Revision 37 EMERGENCY CLASSIFICATION November 15, 2002 ATTIACHMENT B EMERGENCY ACTION LEVELS (EALs)

Category: External Events EAL 6.4.1.1 Sub-Category: Vehicle/Missile Impacts Emergency Classification: UNUSUAL EVENT Emergency Action Level:

IAircraft crash within the ProtectedArea or Switchyard.

Basis:

This event may warrant the prompt notification of state and local authorities and perhaps a precautionary notification of Emergency Response Organization personnel. This event could pose a potential threat to plant operation or personnel safety and therefore warrants declaration of an Unusual Event.

The Protected Area Physical Barrier is defined in the PBNP Security Plan. Note: The Independent Spent Fuel Storage Installation (ISFSI) is a separate Protected Area.

References:

NUREG 0654, Appendix 1 Initiating Condition: Unusual Event 14a SSCP - Security and Safeguards Contingency Plan Page 70 of 94 REFERENCE USE

POINT BEACH NUCLEAR PLANT EPIP 1.2 EMERGENCY PLAN IMPLEMENTING PROCEDURES NNSR Revision 37 EMERGENCY CLASSIFICATION November 15, 2002 ATTACHMENT B EMERGENCY ACTION LEVELS (EALs)

Category: External Events EAL 6.4.1.2 Sub-Category: Vehicle/Missile Impacts Emergency Classification: ALERT Emergency Action Level:

I Aircraft crash affecting operabilityof one(]) train of a safety system. I Basis:

Safety systems as used here designates systems with safety-related functions. Attachment D lists safety systems and systems with safety-related functions.

This condition is entered when Control Room Operators become aware of impact on a safety function after an aircraft crash has been reported.

Only a crash that actually causes damage to equipment required for safe operation AND only damage that renders a single train of a safety system unable to perform its intended safety function meet the threshold of this EAL. A lengthy damage assessment should not be performed. The occurrence of a crash with evidence of damage likely to prevent one train from performing its intended safety function is sufficient for declaration.

This condition is indicative of abnormal plant system conditions with possible adverse consequences on the public health and safety is classified as an ALERT.

Escalation to a higher emergency class, if appropriate, will be based on further system malfunctions, fission product barrier degradation, abnormal radiation levels, or Emergency Director judgment.

References:

NUREG 0654, Appendix 1 Initiating Condition: Alert 18a Page 71 of 94 REFERENCE USE

POINT BEACH NUCLEAR PLANT EPIP 1.2 EMERGENCY PLAN IMPLEMENTING PROCEDURES NNSR Revision 37 EMERGENCY CLASSIFICATION November 15, 2002 ATTACHMENT B EMERGENCY ACTION LEVELS (EALs)

Category: External Events EAL 6.4.1.3 Sub-Category: Vehicle/Missile Impacts Emergency Classification: SITE EMERGENCY Emergency Action Level:

Aircraft crash affecting operabilityof tvo (2) trains of safety systems.

Basis:

This condition is indicative of severe degradation of the level of safety at the plant and with possible adverse consequences on the public health and safety is classified as a Site Emergency.

Only crashes that actually cause damage to equipment required for safe operation of more than one safety system train AND only damage that affects the systems' ability to perform intended functions meet the threshold of this EAL. A lengthy damage assessment should not be performed. The occurrence of a crash with evidence of damage likely to prevent the equipment in more than one train of a safety system from performing intended safety functions is sufficient for declaration.

Safety systems as used here designates systems with safety-related functions. Attachment D lists safety systems and systems with safety-related functions.

References:

NUREG 0654, Appendix 1 Initiating Condition: Site Emergency 16a Page 72 of 94 REFERENCE USE

POINT BEACH NUCLEAR PLANT EPIP 1.2 EMERGENCY PLAN IMPLEMENTING PROCEDURES NNSR Revision 37 EMERGENCY CLASSIFICATION November 15, 2002 ATTACHMENT B EMERGENCY ACTION LEVELS (EALs)

Category: External Events EAL 6.4.2.2 Sub-Category: Vehicle/Missile Impacts Emergency Classification: ALERT Emergency Action Level:

Any missile impact affecting operabilityof one(]) train of a safety system.

Basis:

Safety systems as used here designates systems with safety-related functions. Attachment D lists safety systems and systems with safety-related functions.

This condition is entered when Control Room Operators become aware of impact on a safety functions after a missile impact has been reported.

Only a missile that actually causes damage to equipment required for safe operation AND only damage that renders a single train of a safety system unable to perform its intended safety function meet the threshold of this EAL. A lengthy damage assessment should not be performed. The occurrence of a missile impact with evidence of damage likely to prevent one train from performing its intended safety function is sufficient for declaration.

This condition is indicative of abnormal plant system conditions with possible adverse consequences on the public health and safety is classified as an ALERT.

Escalation to a higher emergency class, if appropriate, will be based on further system malfunctions, fission product barrier degradation, abnormal radiation levels, or emergency management judgment.

References:

NUREG 0654, Appendix 1 Initiating Condition: Alert 18b Page 73 of 94 REFERENCE USE

POINT BEACH NUCLEAR PLANT EPIP 1.2 EMERGENCY PLAN IMPLEMENTING PROCEDURES NNSR Revision 37 EMERGENCY CLASSIFICATION November 15, 2002 (EALs)

B LEVELS ATTACHMENT BACTION ATTACHMENT EMERGENCY EMERGENCY ACTION LEVELS (EALs)

Category: External Events EAL 6.4.2.3 Sub-Category: Vehicle/Missile Impacts Emergency Classification: SITE EMERGENCY Emergency Action Level:

Missile impact affecting operability of two (2) trains of safety systems.

Basis:

Safety systems as used here designates systems with safety-related functions. Attachment D lists safety systems and systems with safety-related functions.

Only missile impacts that actually cause damage to equipment required for safe operation of more than one safety system train AND only damage that affects the systems' ability to perform intended functions meet the threshold of this EAL. A lengthy damage assessment should not be performed. An immediate assessment of the probability of damage making multiple trains incapable of performing their safety function is all that is required. The occurrence of a missile impact with evidence of damage likely to prevent the equipment in more than one train of a safety system from performing intended safety functions is sufficient for declaration.

Major losses of plant safety systems, as defined by failure of the ability of two or more of the safety systems to perform their intended function, warrants declaration of a Site Emergency.

References:

NUREG 0654, Appendix 1 Initiating Condition: Site Emergency 16b Page 74 of 94 REFERENCE USE

POINT BEACH NUCLEAR PLANT EPIP 1.2 EMERGENCY PLAN IMPLEMENTING PROCEDURES NNSR Revision 37 EMERGENCY CLASSIFICATION November 15, 2002 ATTACHMENT B EMERGENCY ACTION LEVELS (EALs)

Category: Fuel Handling/ISFSI Events EAL 7.1.1.2 Sub-Category: Fuel Handling Events Emergency Classification: ALERT Emergency Action Level:

Report of possible damage to irradiatedfitel combined with an alarm on any of the following radiation monitors RE-211, Containment airparticulatemonitor RE-212 Containmentnoble gas monitor RE-221 Drumming Area Vent ManipulatorArea Monitor Spent Fuel Bridge Area Monitor.

Basis:

A report of possible damage to irradiated fuel, combined with an alarm on any of the radiation monitors indicates the probable damage to spent fuel.

NUREG/CR-4982 states that even if no corrective actions are taken, no prompt fatalities are predicted and the risk of injury is low. In addition, NRC Information Notice No. 90-08 presents the following clarifications:

"In the event of a serious accident involving decayed spent fuel, protective actions would be needed for personnel on site, while offsite doses (assuming an exclusion area radius of one mile from the plant site) would be well below the Environmental Protection Agency's Protective Action Guides. Accordingly, it is important to be able to properly survey and monitor for Kr-85 in the event of an accident with decayed spent fuel."

An Alert classification is appropriate for this event. Escalation would be based on actual radiological releases and/or SM judgment.

Page 75 of 94 REFERENCE USE

POINT BEACH NUCLEAR PLANT EPIP 1,.2 EMERGENCY PLAN IMPLEMENTING PROCEDURES NNSR Revisic 9n 37 EMERGENCY CLASSIFICATION Novem ber 15, 2002 ATTACHMENT B EMERGENCY ACTION LEVELS (EALs)

References:

NUREG/CR-4982, "Severe Accident in Spent Fuel Pools in Support of Generic Safety Issue 82" NRC Information Notice No. 90-08, "Kr-85 Hazards from Decayed Fuel" AOP-8B, Irradiated Fuel Handling Accident in Containment AOP-8C, Fuel Handling Accident in Primary Auxiliary Building NUREG 0654, Appendix 1 Initiating Condition: Alert 12 Page 76 of 94 REFERENCE USE

POINT BEACH NUCLEAR PLANT EPIP 1.2 EMERGENCY PLAN IMPLEMENTING PROCEDURES NNSR Revision 37 EMERGENCY CLASSIFICATION November 15, 2002 (EALs)

B LEVELS ATTACHMENT BACTION ATTACHMENTEMERGENCY EMERGENCY ACTION LEVELS (EALs)

Category: Fuel Handling/ISFSI Events EAL 7.2.1.2 Sub-Category: Irradiated Fuel Events Emergency Classification: ALERT Emergency Action Level:

I I !ndictintlon nf irradiatedfelm uncovered.

Basis:

This EAL applies to any area where irradiated fuel is located; reactor cavity, reactor vessel, or the spent fuel pool.

Any releases caused by uncovering the fuel are not generally the primary concern. The primary concern of this EAL is two-fold. First, is the evident loss of control of inventory. The second is the immediate, life threatening dose which could be present in the area due to loss of shielding.

An Alert classification is appropriate for this event. Escalation, if required, would be based on actual radiological releases or Emergency Director judgment.

This EAL applies to spent fuel requiring water coverage and is not intended to address spent fuel which is licensed for dry storage.

References:

NUREG/CR-4982, "Severe Accident in Spent Fuel Pools in Support of Generic Safety Issue 82" NRC Information Notice No. 90-08, "Kr-85 Hazards from Decayed Fuel" AOP-8F, Loss of Spent Fuel Pool Cooling NUREG 0654, Appendix 1 Initiating Condition: Alert 12 Page 77 of 94 REFERENCE USE

POINT BEACH NUCLEAR PLANT EPIP 1.2 EMERGENCY PLAN IMPLEMENTING PROCEDURES NNSR Revision 37 EMERGENCY CLASSIFICATION November 15, 2002 ATTACHMENT B EMERGENCY ACTION LEVELS (EALs)

Category: Fuel Handling/ISFSI Events EAL 7.3.1.1 Sub-Category: ISFSI Events Emergency Classification: UNUSUAL EVENT Emergency Action Level:

A loaded spent fuel cask dropped or tipped.

Basis:

The Independent Spent Fuel Storage Installation (ISFSI) stores spent fuel in vertical casks outside the main Protected Area. Engineering safeguards and procedures insure these casks are not dropped or tipped for the duration of their expected lifetimes. If they should be dropped or tipped it is appropriate to declare an Unusual Event until the situation is analyzed and corrected.

References:

NUREG/CR-4982, "Severe Accident in Spent Fuel Pools in Support of Generic Safety Issue 82" NRC Information Notice No. 90-08, "Kr-85 Hazards from Decayed Fuel" AOP-8G, Ventilated Storage Cask (VSC) Drop or Tipover NUREG 0654, Appendix 1 Initiating Condition: Alert 12 Page 78 of 94 REFERENCE USE

POINT BEACH NUCLEAR PLANT EPIP 1.2 EMERGENCY PLAN IMPLEMENTING PROCEDURES NNSR Revision 37 EMERGENCY CLASSIFICATION November 15, 2002 ATTACHMENT B EMERGENCY ACTION LEVELS (EALs)

Category: Fuel Handling/ISFSI Events EAL 7.3.1.2 Sub-Category: ISFSI Events Emergency Classification: ALERT Emergency Action Level:

Breach of a loaded spentfitel cask as indicatedby a reading of greater than 1000 mRemlhr at 1 meter.

Basis:

The Independent Spent Fuel Storage Installation (ISFSI) stores spent fuel bundles in vertical cask in an area outside the main Protected Area. Engineering safeguards and procedures are in place to insure these casks are not subjected to forces that could breach their integrity. If a cask is breached it is appropriate to declare an Alert due to the potential threat to site personnel.

References:

NUREG/CR-4982, "Severe Accident in Spent Fuel Pools in Support of Generic Safety Issue 82" NRC Information Notice No. 90-08, "Kr-85 Hazards from Decayed Fuel" NUREG 0654, Appendix 1 Initiating Condition: Alert 12 Page 79 of 94 REFERENCE USE

POINT BEACH NUCLEAR PLANT EPIP 1.2 EMERGENCY PLAN IMPLEMENTING PROCEDURES NNSR Revision 37 EMERGENCY CLASSIFICATION November 15, 2002 ATT~ACHMENT B EMERGENCY ACTION LEVELS (EALs)

Category: Emergency Management Judgment EAL 8.1.1.1 Sub-Category: None Emergency Classification: UNUSUAL EVENT Emergency Action Level:

Any event which in the judgment of the Shift Manageror the Emergency Directorcould lead to, or has led to, a potential degradationof the level of safety of the plant.

Basis:

This EAL would pertain to conditions not explicitly addressed elsewhere in the EALs, but which warrant the declaration of an emergency due to the potential degradation of the level of safety of the plant. The Shift Manager or Emergency Director makes this determination.

References:

NUREG 0654, Appendix 1 Initiating Condition: Unusual Event 15 Page 80 of 94 REFERENCE USE

POINT BEACH NUCLEAR PLANT EPIP 1.2 EMERGENCY PLAN IMPLEMENTING PROCEDURES NNSR Revision 37 EMERGENCY CLASSIFICATION November 15, 2002 ATTACHMENT B EMERGENCY ACTION LEVELS (EALs)

Category: Emergency Management Judgment EAL 8.1.1.2 Sub-Category: None Emergency Classification: ALERT Emergency Action Level:

Any event which in the judgment of the Shift Manager or the Emergency Directorcould cause or has caused actual or potentialsubstantialdegradationof the level of safety of the plant.

Basis:

This EAL would pertain to conditions not explicitly addressed elsewhere in the EALs, but which warrant the declaration of an emergency due to the actual or substantial potential degradation of the level of safety of the plant. The Shift Manager or Emergency Director makes this determination.

In keeping with other EALs, generally events which challenge single (RCS or Fuel Cladding) barriers, or affect only single safety systems or functions fall in this category.

References:

NUREG 0654, Appendix 1 Initiating Condition: Alert 19 Page 81 of 94 REFERENCE USE

POINT BEACH NUCLEAR PLANT EPIP 1.2 EMERGENCY PLAN IMPLEMENTING PROCEDURES NNSR Revision 37 EMERGENCY CLASSIFICATION November 15, 2002 ATTACHMENT B EMERGENCY ACTION LEVELS (EALs)

Category: Emergency Management Judgment EAL 8.1.1.3 Sub-Category: None Emergency Classification: SITE EMERGENCY Emergency Action Level:

Any event which in the judgment of the Shift Manager or the Emergency Directorcould indicate actual or likely majorfailures of plantfinctions needed to protect the public. Any releases are not expected to result in exposures in excess of EPA PAGs.

Basis:

This EAL would pertain to conditions not explicitly addressed elsewhere in the EALs, but which warrant the declaration of an emergency due to the actual or likely failure of major plant functions needed for the protection of the public. The Shift Manager or Emergency Director makes this determination.

In keeping with other EALs, generally events which challenge two barriers (but not three), or affect more than one safety system or safety function, fall into this category.

References:

NUREG 0654, Appendix 1 Initiating Condition: Site Emergency 17 Page 82 of 94 REFERENCE USE

POINT BEACH NUCLEAR PLANT EPIP 1.2 EMERGENCY PLAN IMPLEMENTING PROCEDURES NNSR Revision 37 EMERGENCY CLASSIFICATION November 15, 2002 (EALs)

B LEVELS ATTACHMENT BACTION ATT~ACHMENTEMERGENCY EMERGENCY ACTION LEVELS (EALs)

Category: Emergency Management Judgment EAL 8.1.1.4 Sub-Category: None Emergency Classification: GENERAL EMERGENCY Emergency Action Level:

Any event which in the judgment of the Shift Manageror the Emergency Directorcould lead to actualor imminent core damage and the potentialfora large release of radioactivematerial(in excess of EPA PAGs) outside the site boundary.

Basis:

This EAL pertains to conditions not explicitly addressed elsewhere in the EALs but which warrant declaration of an emergency due to actual or imminent core damage and the potential exists for a release of large amounts of radioactive material. The Shift Manager or Emergency Director makes this determination.

In keeping with other EALs, generally events which challenge all three barriers, indicate the potential for core damage, or which reflect possible large releases fall into this category.

References:

NUREG 0654, Appendix 1 Initiating Condition: General Emergency 4 and 7 Page 83 of 94 REFERENCE USE

POINT BEACH NUCLEAR PLANT EPIP 1.2 EMERGENCY PLAN IMPLEMENTING PROCEDURES NNSR Revision 37 EMERGENCY CLASSIFICATION November 15, 2002 ATTACHMENT C FISSION PRODUCT BARRIER (FPB) MATRIX This attachment is used to determine the status of the three primary Fission Product Barriers as they relate to classification. Wherever possible existing well-known parameters have been selected as thresholds for determining the status of the barriers. This is to' integrate setpoints and thresholds already in existence in EOPs and Critical Safety Status Trees into the classification process. The intended purpose is to minimize the number of separate limits and values that must be memorized.

NOTE: Do not "anticipate" challenge or loss of a barrierunless the trend is rapid,and the values are close to the threshold/criteria.

The table on the following page may be used to 'check off the status of the three Fission Product Barriers. Next to each code (FC-1, RL-2, etc.) is an empty box. If the plant conditions meet the conditions in the box, the associated box may be checked, either in the Challenged or Loss column.

The number and status of Fission Product Barriers may then be compared to the EALs that specifically address Fission Product Barrier status (category 1 of Attachment A). Generally, one barrier LOST is an Alert (unless the barrier is Containment alone), two barriers LOST is a Site Emergency, and two barriers LOST, with a CHALLENGE or LOSS of the third barrier is a General Emergency.

The codes (FC-1, RL-2, etc.) may be used to obtain further explanation as to the basis of their development. Each initial code letter; 'F' for Fuel Cladding, 'R' for Reactor Coolant System, or 'C' for Containment is followed by either 'C' for Challenge or 'L' for Loss. (For example FL-# indicates a parameter for Fuel Cladding LOSS, RC-# indicates a parameter for Reactor Coolant System Challenge.)

The bases are on the pages following the Table, arranged by barrier, Challenge then Loss.

Page 84 of 94 REFERENCE USE

POINT BEACH NUCLEAR PLANT EPIP 1.2 EMERGENCY PLAN IMPLEMENTING PROCEDURES NNSR Revision 37 EMERGENCY CLASSIFICATION November 15, 2002 ATTACHMENT C FISSION PRODUCT BARRIER (FPB) MATRIX FUELCL)A(1W .DCHALLENGE FUEL CLAD *LOSS FC-1 ST-2 (Core Cooling) Orange Path. FL-i ST-2 (Core Cooling) Red Path.

Degraded core cooling as indicated by Inadequate core cooling as indicated ANY of the following: by EITHER:

"* CET <700'F AND reactor <25' NR

• CETs >1200°F Also SAMG entry.

"* CET >700°F AND reactor >25' NR

• CETs > 700'F and reactor vessel

"* Reactor vessel <[120] 110' WR with level <25' NR 2 RCPs OR <[60150' with 1 RCP.

FC-2 Failed fuel monitor (RE-109) reading FL-2 Failed fuel monitor (RE-109) reading greater than 120 mRem/hr. greater than 600 mRem/hr.

Unusual Event (1.1.2.1)

FC-3 2 of 3 containment high range monitors FL-3 2 of 3 containment high range monitors reading greater than 1000 Rem/hr. reading greater than 6000 Rem/hr.

Unusual Event (1.1.2.1)

FC-4 Coolant activity greater than Technical FL-4 Coolant activity greater than Specification TS 3.4.16 250 gCi/gram equivalent of 1-131 Unusual Event (1.1.1.1)

FC-5 Any condition which in the judgment of FL-5 Any condition which in the judgment of the Emergency Director is indicative of the Emergency Director is indicative of a challenge to the Fuel Cladding barrier, a loss of the Fuel Cladding barrier.

RCS ,CHALLENGE RCS- 'iLOSS RC-1 RCS leak greater than 10 gpm or RL-1 RCS leak greater than 50 gpm.

500 gallons per day in either steam generator (Technical Specifications). If greater than 400 gpm, see also Unusual Event (1.1.3.1, 1.1.4.1) EAL 2.4.1.3 RC-2 ST-4 (Integrity) Orange Path. RL-2 ST-4 (Integrity) Red Path.

Excess RCS cooldown or cold Temperature in either cold leg <285°F overpressurization of the RCS and cooldown >100'F in the last Unusual Event (1.1.5.1) 60 minutes.

RC-3 Any condition which in the judgment of RL-3 Any condition which in the judgment of the Emergency Director is indicative of the Emergency Director is indicative of a challenge to the Reactor Coolant a loss of the Reactor Coolant System System barrier, barrier.

CONTAINMENT. CHALLENGE CONTAINMENTL LOSS CC-1 ST-5 (Containment) Orange Path. CL-i ST-5 (Containment) Red Path.

Containment pressure >25 psig and Pressure >60 psig increasing following actuation of containment spray OR Sump 'B' >74" CC-2 Hydrogen concentration greater than CL-2 Hydrogen concentration greater than 2%. 4%.

CC-3 Atmospheric dump(s) or reliefs open CL-3 Unisolable steam line break outside and greater than 10 gpm Primary to containment. If primary to secondary Secondary leakage exists leakage >10 gpm exists, see also EAL 1.1.2.2. If>50 gpm, see RL-1.

CC-4 Any condition which in the judgment of CL-4 Inability to isolate Containment.

the Emergency Director is indicative of a challenge to Containment barrier.

CL-5 Any condition which in the judgment of the Emergency Director is indicative of a loss of the Containment barrier.

k Page 85 of 94 REFERENCE USE

POINT BEACH NUCLEAR PLANT EPIP 1.2 EMERGENCY PLAN IMPLEMENTING PROCEDURES NNSR Revision 37 EMERGENCY CLASSIFICATION November 15, 2002 ATTACHMENT C FISSION PRODUCT BARRIER (FPB) MATRIX Fuel Cladding - CHALLENGE FC-1 ST-2 (Core Cooling) Orange Path indicates that RCS subcooling has been lost as well as loss of RCS inventory. RCS subcooling and reactor vessel level are fundamental indications of the assurance of adequate core cooling. These conditions indicate a challenge to the fuel cladding barrier due to degraded core cooling.

For the purposes of emergency classification, the barrier is to be considered CHALLENGED.

FC-2 The function of the failed fuel monitor is to monitor coolant activity. As the fuel cladding barrier degrades increasing amounts of activity are present in the coolant, and seen by this monitor. The value selected is approximately Technical Specifications, hence an Unusual Event must also be declared at this value, if this is the only fission product barrier affected. (1.1.2.1)

For the purposes of emergency classification, the barrier is to be considered CHALLENGED.

FC-3 In-containment high radiation monitors monitor activity in the coolant. As the fuel cladding barrier degrades increasing amounts of activity are present in the coolant, and seen by these monitors. This parameter may be the first indication of cladding degradation due to the location of the failed fuel monitor and possible containment isolation. The value is not correlated to a specific percentage of clad damage.

For the purposes of emergency classification, the barrier is to be considered CHALLENGED.

FC-4 Coolant activity greater than Technical Specifications is considered a precursor to loss of the fuel cladding barrier.

For purposes of emergency classification, the barrier is to be considered CHALLENGED.

FC-5 It is unlikely that any classification scheme can anticipate every circumstance. Therefore this

'threshold' criteria is based on an ad hoc judgment call. If the Emergency Director has reason to believe the integrity of this barrier is being challenged, he may declare it so.

He should have objective reason to believe the barrier is challenged. Simply not knowing (for example loss of indications) should not be used as a basis for declaring a barrier challenged or lost.

If the barrier is subsequently determined not to have been challenged, it may be declared intact.

If it is determined that the barrier was challenged, but is no longer challenged, the barrier must remain as challenged, until the Recovery phase of the emergency.

For the purposes of emergency classification, the barrier is to be considered CHALLENGED.

Page 86 of 94 REFERENCE USE

POINT BEACH NUCLEAR PLANT EPIP 1.2 EMERGENCY PLAN IMPLEMENTING PROCEDURES NNSR Revision 37 EMERGENCY CLASSIFICATION November 15, 2002 ATTACHMENT C FISSION PRODUCT BARRIER (FPB) MATRIX Fuel Cladding - LOSS FL-1 ST-2 (Core Cooling) Red Path indicates that RCS subcooling has been lost as well as significant loss of RCS inventory. RCS subcooling and reactor vessel level are fundamental indications of the assurance of adequate core cooling. These conditions indicate the fuel cladding barrier has been subjected to conditions which may cause its failure due to inadequate core cooling.. For the purposes of emergency classification, the barrier is to be considered LOST.

Core exit thermocouple reading in excess of 1200'F is also an entry condition for Severe Accident Management Guidelines (SAMGs).

FL-2 The function of the failed fuel monitor is to monitor coolant activity. As the fuel cladding barrier degrades increasing amounts of activity are present in the coolant, and seen by this monitor. The value is not correlated to a specific percentage of clad damage, but is beyond Technical Specifications.

For the purposes of emergency classification, the barrier is to be considered LOST.

FL-3 In-containment high radiation monitors monitor activity in the coolant. As the fuel cladding barrier degrades increasing amounts of activity are present in the coolant, and seen by these monitors. This parameter may be the first indication of cladding degradation due to the location of the failed fuel monitor and possible containment isolation. The value is not correlated to a specific percentage of clad damage, but is beyond Technical Specifications.

For the purposes of emergency classification, the barrier is to be considered LOST.

FL-4 Coolant activity greater than this level is not correlated to a specific percentage of clad damage, but is beyond Technical Specifications.

For the purposes of emergency classification, the barrier is to be considered LOST.

FL-5 It is unlikely that any classification scheme can anticipate every circumstance. Therefore this

'threshold' criteria is based on an ad hoc judgment call. If the Emergency Director has reason to believe the integrity of this barrier is lost, he may declare it so.

He should have objective reason to believe the barrier is lost. Simply not knowing (for example loss of indications) should not be used as a basis for declaring a barrier challenged or lost.

If the barrier is subsequently determined not to have been lost, it may be declared intact, or challenged, as appropriate.

If it is determined that the barrier was lost, but is no longer lost, the barrier must remain as lost, until the Recovery phase of the emergency.

For the purposes of emergency classification, the barrier is to be considered LOST.

Page 87 of 94 REFERENCE USE

POINT BEACH NUCLEAR PLANT EPIP 1.2 EMERGENCY PLAN IMPLEMENTING PROCEDURES NNSR Revision 37 EMERGENCY CLASSIFICATION November 15, 2002 I

ATTACHMENT C FISSION PRODUCT BARRIER (FPB) MATRIX Reactor Coolant System - CHALLENGE RC-1 These conditions represent minor leakage from the RCS. Because the source of the leak may not be known, and leaks can become worse, these conditions are considered precursors to more serious events. As such, an Unusual Event must be declared on these conditions, if the RCS is the only barrier affected. (1.1.3.1, 1.1.4.1)

For the purposes of emergency classification, the barrier is to be considered CHALLENGED.

RC-2 Conditions of ST-4 (Integrity) Orange Path reflect an excessive cooldown of the vessel or cold overpressurization of the RCS. These conditions represent a challenge to the RCS barrier. An Unusual Event must be declared. (1.1.5.1)

For the purposes of emergency classification, the barrier is to be considered CHALLENGED.

RC-3 It is unlikely that any classification scheme can anticipate every circumstance. Therefore this

'threshold' criteria is based on an ad hoc judgment call. If the Emergency Director has reason to believe the integrity of this barrier is being challenged, he may declare it so.

He should have objective reason to believe the barrier is challenged. Simply not knowing (for example loss of indications) should not be used as a basis for declaring a barrier challenged or lost.

If the barrier is subsequently determined not to have been challenged, it may be declared intact.

If it is determined that the barrier was challenged, but is no longer challenged, the barrier must remain as challenged, until the Recovery phase of the emergency.

For the purposes of emergency classification, the barrier is to be considered CHALLENGED.

Page 88 of 94 REFERENCE USE

POINT BEACH NUCLEAR PLANT EPIP 1.2 EMERGENCY PLAN IMPLEMENTING PROCEDURES NNSR Revision 37 EMERGENCY CLASSIFICATION November 15, 2002 ATTACHMENT C FISSION PRODUCT BARRIER (FPB) MATRIX Reactor Coolant System - LOSS RL-1 This value is derived from NUREG-0654, Appendix 1. Although 50 gpm is well within the capacity of available pumps, this leak can be either into Containment or from Primary to Secondary systems. Thus, the RCS barrier is no longer serving its function of preventing the transport of fission products.

For the purposes of emergency classification, the barrier is to be considered LOST.

RL-2 Conditions of ST-4 (Integrity) Red Path reflect an excessive cooldown of the vessel. These conditions indicate the RCS barrier has been subjected to conditions which may cause its failure For the purposes of emergency classification, the barrier is to be considered LOST.

RL-3 It is unlikely that any classification scheme can anticipate every circumstance. Therefore this

'threshold' criteria is based on an ad hoc judgment call. If the Emergency Director has reason to believe the integrity of this barrier is lost, he may declare it so.

He should have objective reason to believe the barrier is lost. Simply not knowing (for example loss of indications) should not be used as a basis for declaring a barrier challenged or lost.

If the barrier is subsequently determined not to have been lost, it may be declared intact, or challenged, as appropriate.

If it is determined that the barrier was lost, but is no longer lost, the barrier must remain as lost, until the Recovery phase of the emergency.

For the purposes of emergency classification, the barrier is to be considered LOST.

Page 89 of 94 REFERENCE USE

POINT BEACH NUCLEAR PLANT EPIP 1.2 EMERGENCY PLAN IMPLEMENTING PROCEDURES NNSR Revision 37 EMERGENCY CLASSIFICATION November 15, 2002 ATTACHMENT C FISSION PRODUCT BARRIER (FPB) MATRIX Containment - CHALLENGE CC-1 ST-5 (Containment) Orange Path represent conditions beyond normal operating parameters due to either pressure or sump "B" level.

For the purposes of emergency classification, the barrier is to be considered CHALLENGED CC-2 Existence of hydrogen at these concentrations does not yet represent an explosive mixture, however, there are limited means to reduce hydrogen in containment, especially during an emergency.

For the purposes of emergency classification, the barrier is to be considered CHALLENGED.

CC-3 This challenge threshold is designed to ensure that if Fuel Cladding AND RCS barriers are LOST, a General Emergency would be declared if the atmospheric dump valves or relief valves on the affected steam generator open (or are opened) and greater than 10 gpm Primary to Secondary leakage exists.

If the Primary to Secondary leakage is less than 10 gpm the RCS barrier may be considered intact.

This threshold is included to address NUREG-0654, Appendix 1 Initiating Condition A4.

For the purposes of emergency classification, the barrier is to be considered CHALLENGED.

CC-4 It is unlikely that any classification scheme can anticipate every circumstance. Therefore this

'threshold' criteria is based on an ad hoc judgment call. If the Emergency Director has reason to believe the integrity of this barrier is being challenged, he may declare it so.

He should have objective reason to believe the barrier is challenged. Simply not knowing (for example loss of indications) should not be used as a basis for declaring a barrier challenged or lost.

If the barrier is subsequently determined not to have been challenged, it may be declared intact.

If it is determined that the barrier was challenged, but is no longer challenged, the barrier must remain as challenged, until the Recovery phase of the emergency.

For the purposes of emergency classification, the barrier is to be considered CHALLENGED.

Page 90 of 94 REFERENCE USE

POINT BEACH NUCLEAR PLANT EPIP 1.2 EMERGENCY PLAN IMPLEMENTING PROCEDURES NNSR Revision 37 EMERGENCY CLASSIFICATION November 15, 2002 ATTACHMENT C FISSION PRODUCT BARRIER (FPB) MATRIX Containment - LOSS CL-1 ST-5 (Containment) Red Path represent conditions indicate the containment barrier has been subjected to conditions which may cause its failure.

For the purposes of emergency classification, the barrier is to be considered LOST.

CL-2 Hydrogen at these concentrations may detonate. This would create an explosion in Containment.

For the purposes of emergency classification, the barrier is to be considered LOST.

CL-3 Main steam line piping outside containment, up to and including the isolation valves may be considered a part of the Containment barrier.

The inability to isolate assumes it is desired and has been attempted. This attempt includes only actions which may be taken from the Control Room. If actions must be taken outside the Control Room to isolate, the barrier must be considered lost.

For the purposes of emergency classification, the barrier is to be considered LOST.

CL-4 This criteria includes all isolation paths, including access hatches. Only one valve or door in a given path need be closed.

A physical loss of integrity (crack or hole) also meets this criteria.

For the purposes of emergency classification, the barrier is to be considered LOST.

CL-5 It is unlikely that any classification scheme can anticipate every circumstance. Therefore this

'threshold' criteria is based on an ad hoc judgment call. If the Emergency Director has reason to believe the integrity of this barrier is lost, he may declare it so.

He should have objective reason to believe the barrier is lost. Simply not knowing (for example loss of indications) should not be used as a basis for declaring a barrier challenged or lost.

If the barrier is subsequently determined not to have been lost, it may be declared intact, or challenged, as appropriate.

If it is determined that the barrier was lost, but is no longer lost, the barrier must remain as lost, until the Recovery phase of the emergency.

For the purposes of emergency classification, the barrier is to be considered LOST.

Page 91 of 94 REFERENCE USE

POINT BEACH NUCLEAR PLANT EPIP 1.2 EMERGENCY PLAN IMPLEMENTING PROCEDURES NNSR Revision 37 EMERGENCY CLASSIFICATION November 15, 2002 D

ATTACHMENT ATTACHMENT D SAFETY AND SAFETY-RELATED SYSTEMS Designator System Safety-Related Functions AF Auxiliary Feedwater Feedwater supply AMSAC ATWS M itigation System Non-safety-related isolation Actuation Circuitry BS Boron Recycle Supports Safety Injection CCW Component Cooling Water Cools safety-related equipment CI Containment Integrity Containment integrity COMP Computers Monitoring CONT Containment Structures Containment integrity CP Containment Penetrations Containment integrity CS Condensate and Feedwater Core Cooling CV Chemical and Volume Control Emergency cooling, containment integrity DA Diesel Starting Air Diesel start DG Diesel Generator Emergency power supply ESF Engineered Safety Features Core Cooling and Integrity (Safeguards)

FH Fuel Handling Fuel integrity FM In core Flux Mapping Fuel integrity FO Fuel Oil Power supply FP Fire Protection Fire Protection FW Feedwater (I&C only) Non-safety-related isolation HV Auxiliary Steam, Heating Steam & Containment cooling and integrity Condensate, Chilled & Hot Water Page 92 of 94 REFERENCE USE

POINT BEACH NUCLEAR PLANT EPIP 1.2 EMERGENCY PLAN IMPLEMENTING PROCEDURES NNSR Revision 37 EMERGENCY CLASSIFICATION November 15, 2002 ATTACHMENT D SAFETY AND SAFETY-RELATED SYSTEMS Designator System Safety-Related Functions IA Instrument Air Containment isolation and integrity IST Inservice Test Equipment Reactor coolant system integrity (i.e., steam generator nozzle dams)

MRR Metering, Relaying, & Regulation Monitoring MS Main, Extraction, Gland Seal & Containment integrity, heat removal Reheat Steam NG Nitrogen Gas Monitoring NI Nuclear Instrumentation Reactor protection PACV Post-Accident Vent, Drains, etc. Containment integrity, containment hydrogen control PPCS Plant Process Computer System Monitoring RC Reactor Coolant Reactor coolant system integrity, reactor protection, containment integrity RDC Rod Drive Control Reactor coolant system integrity, reactor protection RH Residual Heat Removal (LPSI) Containment integrity, emergency cooling RM Radiation Monitoring Monitoring, RCS and containment integrity RP Reactor Protection Reactor protection, monitoring, RS Radwaste Steam Non-safety-related isolation S Structures Safety-related equipment safety SA Service Air Containment integrity SF Spent Fuel Cooling and Filtration Heat removal and containment integrity SC Primary Sampling Containment and RCS integrity SI Safety Injection (HPSI) Emergency cooling, heat removal, containment integrity Page 93 of 94 REFERENCE USE

POINT BEACH NUCLEAR PLANT EPIP 1.2 EMERGENCY PLAN IMPLEMENTING PROCEDURES NNSR Revision 37 EMERGENCY CLASSIFICATION November 15, 2002 ATTACHMENT D SAFETY AND SAFETY-RELATED SYSTEMS Designator System Safety-Related Functions Service Wrater Feedwater supply, heat removal, SW containment integrity VNBI PAB Batt¢cry & Inverter Room Heat removal, battery room hydrogen H&V control VNCC Containm ent Accident Fans H&V Heat removal VNDG Diesel Ge nerator Room H&V Support Diesel operation VNPSE Containm ent Purge Supply & Containment integrity Exhaust H&V VNRC Reactor Cavity Cooling H&V Containment integrity WG Waste Gas Containment integrity WL Waste Liquid Containment integrity Y Vital Instrument Bus 120 VAC Power supply 4.16KV 4160V Electrical Power supply 480V 480V Electrical Power supply 125V 125VDC Electrical Power supply Page 94 of 94 REFERENCE USE