Text

Public Version of Emergency Plan Implementing Procedures, Including Rev 12 to EPIP-OSC-2-1, Emergency Plan Activation & Notifications & Rev 5 to EPIP-OSC-15-1, Control Room Communication
	ML20215E699
Person / Time
Site:	Fort Calhoun
Issue date:	06/04/1987
From:	; OMAHA PUBLIC POWER DISTRICT
To:	;
Shared Package
ML20215E689	List: ML20215E686; ML20215E699;
References
	PROC-870604-01, NUDOCS 8706220062
	Download: ML20215E699 (114)
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p .c i OMAHA PUBLIC' POWER DISTRICT ^ CONFIRMATION OF TRANSMITTAL EMERGENCY PLAN IMPLEMENTING PROCEDURES l (EPIP) NAME: >>md bnd/ >A DATE: June 4, 1987 Holds and Maintains Copy No. //o[ The following documents are provided.for your use: 1 Remove Insert Procedure No. Pace No.(s)- Procedure No. Paae No.(s) Table of 1 (4/27/87) Table of i (5/16/87) Contents ii (3/24/87) Content's ii (5/16/87) v (3/26/87) v (5/20/87)' xi (3/17/87) xi (5/23/87) OSC OSC-2-1 thru 05C-2-18 OSC-2 OSC-2-1 thru 05C-2-20 (4/25/87) (5/15/87) OSC-15 05C-15-1 thru 05C-15-5 OSC-15 OSC-15-1 thru OSC-15-5 l (8/13/86) (5/6/87) 3 OSC-16 OSC-16-1 thru 05C-16-22 OSC-16 05C-16-1 thru OSC-16-22 (2/26/87) ~ (5/15/87) (ContinuedonBack) NOTE: Procedure OSC-15 contains h' ^ ^- d. Proprietary Information. Manager - Radiological Health and Emergency Planning I hereby acknowledge receipt of the above copy or numbered pages. The additional or revised pages have been included in my assigned copy of the EPIP and/or super-seded pages have been removed as required. 4,. i Signed-Date (Please sian and return this form within 5 davs to Rhonda Hankins. Jones Street Station. Omaha Public Power District.1623 Harney Street. Omaha. NE 68102). NOTE: If your copy of the Emergency Plan Implementing Procedures has been trans- { ferred to another person or address, please fill out the spaces below. Name of Holder Address 4 Title / Department i 8706220062 870604 PDR ADOCK 05000285 F ppg i 1. ,,.m. a.,... r~...rw--.

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t t OMAHA PUBLIC POWER DISTRICT CONFIRMATION OF TRANSMITTAL EMERGENCY PLAN IMPLEMENTING PROCEDURES (EPIP) (Continued) hhfhhureNo, Pace No.(s) Pr edure No. Pace No.(s) OSC-18 OSC-18-1 thru OSC-18-3 OSC-18 05C-18-1 thru OSC-18-4' (10/16/86) (5/15/87) TSC-8 TSC-8-1 thru TSC-8-D-4 TSC-8 TSC-8-1 thru TSC-D-4 (2/23/87) j/ (5/15/87)- RR-79 RR-79-1 thru RR-79-2 (New Procedure) (5/20/87) .j 1 i 1 1 j o 'y 1 1 ~ )

l 1 c VOLUME III ~ OMAHA PUBLIC POWER DISTRICT - FORT CALHOUN STATION EMERGENCY PLAN IMPLEMENTING PROCEDURES TABLE OF CONTENTS 1. OPERATION SUPPORT CENTER Revision Procedure No. Title No./Date EPIP-OSC-1 Emergency Classification R11 02-26-87 EPIP-OSC-2 Emergency Plan Activation R12 05-15-87 EPIP-05C-3 Notification of Unusual Event Actions R4 05-23-86 EPIP-05C-4 Alert Event Actions R4 05-23-86 EPIP-05C-5 Site Area Emergency Actions. R4 05-23-66 EPIP-05C-6 General Emergency Actions R4 05-23-86 ' t EPIP-OSC-7 Personnel Rescue R5 04-30-86 + EPIP-OSC-8 Medical Assistance R3 12-22-86 EPIP-OSC-9 Emer.gency Repairs, Corrective Actions and Damage Control R1 06-20-85 EPIP-OSC-10 Initial Assessment of Plant Parameters and Effluent Monitors (R5 06-20-85) to Determine Source Term DELETED 1-23-86 EPIP-05C-11 Initial Dose Assessment Based (R3 10-11-84) on Plant Instrumentation DELETED 1-23-86 EPIP-OSC-12 Accidental Actuation of Early Warning Siren System R4 03-06 I EPIP-OSC-13 Onsite Radiological Monitoring R2 03-12-87 EPIP-OSC-14 Shift Supervisor / Site Director Actions R9 03-23-87 EPIP-OSC-15 Control Room Communicator R5 05-06-87 ~ E

4225$ ilsys h, 4

FC/INDEX/01 05-16-87 i

i k t l l i ( VOLUME III TABLE OF CONTENTS i (Continued) j 1. OPERATION SUPPORT CENTER (Continued) I ( Revision ) Procedure No. Title No./Date EPIP-05C-16 Emergency Team R5 05-15-87 EPIP-OSC-17 Technical Augmentation Staff R1 06-20-87 EPIP-OSC-18 Initial Response Organization Notification R2 05-15-87 .l ( C ii FC/INDEX/01 05-16-87

r t VOLUME III mk, TABLE OF CONTENTS (continued) 3. TECHNICAL SUPPORT CENTER Revision Procedure No. Title No./Date EPIP-TSC-1 Activation of Technical S,upport Center R7 03-21-87 i EPIP-TSC.-2 Control Room & Technical Support Center Communication R3 05-05-86 EPIP-TSC-3 Plant and Reactor Operation Support - Alert Classification (R2 10-04-84) DELETED 05-28-86 EPIP-TSC 4 Plant and Reactor Operation Support - Site Area Emergency Classification (R3 10-04-84) DELETED 05-28-86 EPIP-TSC-5 _ Plant and Reactor Operation Support - General Emergency Classification (R2 10-04-84) DELETED 05-28-86 EPIP-TSC-6 Plant Engineering and Repair (R2 09-24-84) DELETED 05-28-86 EPIP-TSC-7 Emergency Response Assistance Combustion Engineering R1 04-30-86 EPIP-TSC-8 Estimate of Core Damage R4 05-15-87 C' y G.- FC/INDEX/01 05-20-87

a VOLUME III l.'^ TABLE OF CONTENTS (Continuea) 5. PUBLIC INFORMATION Revision ) Procedure No. Title No./Date EPIP-RR-67 Emergency Recovery Organization's Clerical Assistant R2 12-22-86 EPIP-RR-68 Emergency Recovery Organization's Recovery Manager Secretary R2 12-22-86 EPIP-RR-69 Recovery Organization's Dose Assessment Operator R3 11-12-86 EPIP-RR-70 Emergency Recovery Organization's Recovery Operations Coordinator R1 06-20-85 EPIP-RR-71 Emergency Recovery Organization's Recovery Manager Communicator R2 12-22-86 EPIP-RR-72 Emergency Recovery Organization's Dose Assessment Specialist R3 02-26-87 EPIP-RR-73 Emergency Recovery Organization's Dose Assessment Data Processor R2 10-16-86 EPIP-RR-74 Emergency Recovery Organization's Chemistry and Environmental Survey Coordinator R2 04-30-86 EPIP-RR-75 Emergency Recovery Organization's Site Representative (Des Moines, Iowa) R1 04-30-86 EPIP-RR-76 Emergency Recovery Organization's EOF Aaministration Supervisor R2 10-22-86 EPIP-RR-77 Emergency Recovery Organization's Call list Caller R0 12-22-86 EPIP-RR-78 Emergency Recovery Organization's Corporate Spokesperson R0 03-10-87 EPIP-RR-79 Emergency Recovery Organization's Computer Specialist R0 05-20-87 Xi FC/INDEX/01 05-23-87.

x. 1 EPIP-OSC-2-1 .l C Fort Calhoun Station Uriit No.1 'i EMERGENCY PLAN IMPLEMENTING PROCEDURE i ~ EPIP-OSC-2 Emeroency Plan Activation and Notifications i I. PURPOSE This procedure provides the instruction to be followed by the Shift l Supervisor or his designee when an emergency occurs that requires initiation of the emergency plan. I II. PRERE0VISITE ] 1. The emergency has been classified in accordance with EPIP-05C-1. 2. Initial actions to ensure the plant-is in a safe condition have .been'taken. 3. A completed " Initial Response Organization Call List" (Attachment'3), is available in the N.O. Station's Shift Supervisors Office, Systems Operations dispatcher 43rd Street Dispatch, Fort Calhoun Station Control Room and all Emergency Team TAG No.1 personnel. II. PRECAUTIONS ~ 1. All significant events and actions shall be logged in the Operations Log Book. 2. The person relieving the Shift Supervisor as Site Director shall be fully briefed prior to assuming responsibility. 3. The " Initial Response Organization Call List" is latest revision. I

V. PROCEDURE

1. Notification of Unusual Event Classification A. The Shift Supervisor or his designee shall perform the following: 1. Notify plant personnel by' announcing the classification condition twice'over the Gaitronic system. I 1SSUED FC/EPIP/03 R12 05-15-87 P 1$W& 45 9 g

~ i EPIP-OSC-2-2 I

V. PROCEDURE

(Continued) e 1.

A. 2. Initiate emergency response by utilizing the Initial Response Organization Call List. j 3. Notify State / Agencies: a. Provide initial notification within 15 minutes to the ] Nebraska State Civil Defense Agency and Iowa Office of -{ Oisaster Services via the Conference Operations (COP) i Network. Complete the form shown in Attachment 1 of this procedure for any and all emergency classifications j This form is titled Nuclear Power Plant Incident Initial y Report to Offsite Government Acencies. b. The telephone and NAWAS may be used as alternate metnods l of notification. See Attachment 3 of EPIP-OSC-15 for additional requirements when these are used.- 'l 4 Notify the NRC Operation Center via the ENS (Red Telephone) immediately after State Notification, NOT.to exceed I hour. after classification. (Refer to OPPD Standing Order R-11). Backup telephone numbers are on the red telephones. 5. Use Gai-tronics to contact Shift HP'and Chemist if they are not already in the Control Room. 6. Notify the Security Force (Ext. 6657) to confirm that the 5 Security Emergency Procedures are in effect. d 7. Consider evacuation of compartments within the. containment and auxiliary buildings. j i 2. Alert Classification ) A. The Shift Supervisor or his designee shall perform the following: 1. Notify plant personnel by sounding _the Nuclear. Emergency Alarm followed by an announcement on the Gaitronic system with special evacuation instructions. Repeat the instructions, j 2. Perform actions listed for " Notification of Unusual Event" if not completed. 3. Complete accountability of personnel. 4 Initiate personnel search and rescue as needed. i ISSUED g a MAY 151987 FC/EPIP/03 R12 05-15-87 E as +-g, ,a mee.e i. =A=4 ,w-L

+ 4 4 J t EPIP-OSC-2-3 h I

V. PROCEDURE

(Continued) 2.

A. 5. Direct in-plant radiological surveys and determine control room habitability. 6. Direct offsite surveys as additional emergency monitors arrive. 7. Provide update reports to the States whenever emergency ] classifications are changed and/or the emergency is' terminated, or every 15 minutes during a radioactive release. In the Control Room, fill out FC-195. (Attachment 2) and transmit using the CDP Network, regular phone, or NAWAS. In the TSC, the terminetts will be used; If unavailable, transmit FC-195 data to the States using the CDP Network or the Facisimile Network. 8. Transfer emergency responsibility with complete briefing to relie/ing Site Director or Recovery Manager when they i are available, in accordance with either EPIP-EOF-13 or EPIP-EOF-14'. 3. Site Area Emeroency and General Emeroency Classifications A. The Shift Supervisor or his designee shall perform the following-1 1. Perform actions listed for the " Notification of Unusual Eve.tt" and " Alert" classifications if not completed 1 2. Within 15 minutes of classification, contact counties i and recommend Early Warning Siren Activation 1.A.W. EPIP-EOF-17, 3. Transfer emergency responsibility with complete briefing ) to the Site Director or Recovery Manager when those persons become available. The transfer shall be performed in accordance with either EPIP-EOF-13 or EPIP-EOF-14 4 The Shift Supervisor should now direct all efforts to operation of the reactor and plant. B. The Site Director or his designee shall perform the following: i 1. Continue update reports to State and County Agencies l until responsibility is transferred to the Recovery Manager. c. 2. Provide :ttitial notification to Counties at Site Area 4 { and General Emergencies unless EOF personnel agree to j make such notifications. i 1SSUED i FC/EPIP/03 R12 05-15-87

y t p EPIP-0SC,2-4 ' 6 .j ATTACHMENT 1 NUCLEAR POWER PLANT INCIDENT INITIAL REPORT TO 0FFSITE GOVERNMENT AGENCIES 1. This is , at the (Name') (Title) ] Fort Calhoun plant. ' Telephone call-back number (USE ONLY WHEN C0P i NETWORK IS NOT USED) is Time is 2. NOTIFICATION OF UNUSUAL EVENT / ALERT / SITE AREA / GENERAL EMERGENCY was declared at on -(Time) (Date) 3. Airborne / Liquid / No Release of radioactive material occurred. (ITEMS 4 THRU 6 WILL NOT BE COMPLETED WHEN NO RELEASE HAS OCCURRED) 4. The estimated duration of the release is minutes. The release is Terminated / In Progress / Potential. 5. Wind speed is MPH; wind direction is from 6. Sector (s)affectedare for

miles, for miles.

7. Recommended. protective actions are: (See EPIP-EOF-7, Table EOF-7.1) [] Advisory - for information only - no public notification

i (Notification of Unusual Event).

1 Standby - be prepared for possible public action (Alert). Public Reponse - notify public (Site Area and General). 1 [] 15-minute siren notification - standby for more information (Site Area and General) [] Limit use of potentially affected water (liquid release). [] Inhouse shelter may be necessary (Site Area). [] Inhouse shelter (General Emergency - 2 mile radius /5 mile downwind). [] Evacuation may be necessary (Site Area - General Emergency). 8. Remarks; (This IS/IS NOT an exercise. Repeat. This IS/IS NOT an exercise) 9. Report was received by: (NAML) (AGENCY). (TIML) (DAit) 7 FC/EPIP/03 ISSUED R12 05-15-87 MAY 151987 x

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A TO t

PE AE S Z TAP TI AHS r "f OT NH T ') 7 NN TE e" E T UT S' EA D R t L SG I L ['S. MG ED E f s B A SE 1 HR REY NRR as A LT IT CO ATC OEE e R L - .E L A LSN HGB ,r E I 2 RL LT TE CEE PAM dp P W EP AR-TS EUG PU e O E LM CO AN DQR E N t n N RH LO P O EE M* se I ET AC EE P NRM O* ih L C' HR' S E E H lT 5 LO L T E ES 1 AT TL ,D R BIR s C SA N M Y IH EA U a. L STO l E TT LS S A AI Y E rl T SI A2 H N ea S I L LR C3 I T O bc Y LI LE 6-H m S B AL R6-I N P u 'o LI CL E/ I nt R LS A H3 K E AN EC T2 R em G CO H I 6 O h e A P TT E6 T W th P ES S N t HE FI NT E l E TR IL I A V at H E i n T da L w F A t I 1 2 3 U su S ro 7 U Y i y D 8 N C f 9 U NY r E 1 EC E ,e F GN M sb E U 5 O RE A rm T 1 EG N eu O S N MR bn N Y O .EE m S A T AE nn t M ue I .I h A E o C RL N rh I TAA O ep FR R I 3 g I EEE ) T 2) RR1) ae TLTN E D A D EE D ph OAI E L RE MT#E DKfE t N SG T YOR RS R RL R e N I RSI OI GI OAGI hl AAAA T STI U FLAU CTAU t a d / HCSVQ NATQ E TQ i .E RE .E rd Y TII RE I I R N R o T LSMER C U AYEP FET.1 COT.2 l F D EiHU1 OP SH HPCS( ESE( TPE( 4 1 ' i:' 1 6 l' 3

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A TO Z

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A TO Z

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A TO Z

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4 o i EPlP-OSC-15-1 I 4 -j i Fort Calhoun Station Unit No. 1 ) . (~j.. EMERGENCY PLAN IMPLEMENTING PROCEDURE 3 EPIP-OSC-15 Control Room Comunicator I. PURPOSE To provide a procedure which delineates the duties, responsibilities and actions of the Control Room Communicator (6th Operator on duty). II. PRERE0VISITES The Cortrol Room Communicator has been trained. III. PRECAUTIONS The " Site Director" is that person directing the Emergency Response effort. For this procedure " Site Director" indicates the Shift Supervisor or his designee. I

V. PROCEDURE

1. Report to the Control Room once an emergency has been classified or upon request of the Shift Supervisor, i 2. Obtain key to Control Room Emergency Gear Locker from the Shift Supervisor or break key box on locker door. Open locker and obtain Control Room Communicator Equipment Box. 3. At work station, set up log book and log all notifications calls made by Control Room and the time these calls were made. 4 If directed by the Site Director, make notification to North Omaha Station or System Operations or 43rd St. Dispatch to start emergency call list. See Shift Supervisor Checklist (frnm EPIP-OSC-14, entitled " Shift Supervisor / Site Director Actions) l for phone numbers. Inform contact who you are, what classification has been declared, and if event is a drill or actual emergency. 5. If directed by Site Director, make initial notification to States of Iowa and Nebraska using available communications. See Shift. Supervisor Checklist for call requirements / phone numbers, etc. If the regular phone must be used, see Attachment 2 of this procedure for verification procedure. Read off information on form (Attachment 1 of EPIP-0SC-2) once contact has been made. I 6. If directed by Site Director, make initial notification to Washington. Harrison, and Pottawattamie' Counties using available connunications. See Shift Supervisor Checklist for call requirements / phone numbers. etc. Read off information on form (Attachment 1 of EPIP-EOF-17) once-t contact har been made. skueu g FC/EPIP/01 MAY 0 6 Igg 7 R5 05-06-87

EPIP-0SC-15-2 I

V. PROCEDURE

(Continued) 7. If directed by Site Director, make notifi a on t KFAB s available communications. See EPIP E0 - phone numbers, etc. Read off information on form (from EPIP-EOF-17) once contact has been made. 8. If directed by Site Director make notification to NRC Operations Center using available commun,ications. See Shift Supervisor i Checklist for call requirements / phone numbers, etc. Read off information on form (Appendix B and C from Standing Order R-11), if applicable, once contact has been made. 9. If directed by the Site Director, make update notifications to l l States of Iowa and Nebraska using available communications. See Shift Supervisor Checklist for call requirements / phone numbers, etc. I Read off information on form (Attachment 2 of EPIP-OSC-2) once contact has been made. l

10. Establish a conference network with the following parties as they arrive at their stations:

i l j i 1 l To accomplish this, perform the following: ~ a. Use th one in the Control Room. b. If desired, connect headset unit (from Control Room Communicator i kit) to phone for long-term use. c. When the first position arrives and call u will be in conference with them, upon answering. d. When the next position arrives and calls 6685, he will re - sy signal, which he should hold for 10 seconds. You ill' then hear a " beep". Refer to Attachment 1 for in ructions on how to connect to that caller, and how to place them in conference if desired (and how to place a call while in conference), e. Repeat part d. above for each position until all Jositions are on the conference line. FC/EPIP/01 MAY 0 6 05-06-87

m 4 d EPIP-OSC-15-3 [~ '^ I

V. PROCEDURE

(Continued) .e

11. Once any part of this conference is established, pass any data requested by the parties on the line. Instruct Control Room Data Collector to obtain requested data on FC-194 and/or FC-197 forms, then read this data over conference line. Update data and repeat data transmission every 15 minutes or as directed by personnel on conference line.

12. You will be relieved of making certain notification calls as.

soon as required positions are manned, or facilities are activated. This is summarized as follows: a. State Updates: Once Dose Assessment is performed in'the TSC or EOF, update will be handled via computer / phone modem system, FAX Network, or C0P Network from those facilities. b. NRC Updates: Once TSC Communicator position is manned in.the may take over communication on the ENS (Red) phone. TSC, they !"Yontrol Room personnel must man that phone as Until then requested by NRC. l i S $ U J ?. FC/EPIP/01 MAY 0 61987 Jus 05-06-87 /

EPIP-OSC-15-4 PROCEDURES FOR ROLM PHONE OPERATION ESTABLISHING A CONFERENCE CALL l 1. Dial first number - tell individual to wait 2. HOOK - FLASH. 3. Dial next number - tell individual to wait 4 HOOK - FLASH j 'i 5. Press

4 - now everyone is in conference i

) 6. Repeat steps 2 thru 5 for all numbers s. g i RECEIVING CALL WHILE IN CONFERENCE-l , ~. 1. Beep - (beep is heard when calling party waits on busy signal for 10

)

seconds) 2. HOOK - FLASH 2 3. Press

1 - you are now connected to caller 4

HOOK - FLASH l i Sa. Press

1 - you are now back in conference 5b. Press
4 - caller and you are now back in conference call 1

PLACING CALLS WHILE IN CONFERENCE 1. HOOK - FLASH 2. Dial Number - you may call wherever you wish-3. HOOK - FLASH 4 Press

1 - you are reconnected to conference

'v lSSUED FC/EPIP/01 RS 05-06-87 MAY 0 61987

\\,_ n q I EPIP-OSC-15-5 (*, ATTACHMENT 2 VERIFICATION PROCEDURE i i IF NOTIFICATION OF THE STATES IS' PERFORMED USING THE REGULAR PHONE SYSTEM, A VERIFICATION PROCEDURE MUST BE PERFORMED AS FOLLOWS: ] a. Obtain the Nebraska State Patrol letter from the. Shift Supervisor's key locker. i J b. Contact the Nebraska State Patrol us.ing phone numbers from the Shift Supervisor Checklist, c. Once dispatcher answers, identify yourself.' explain purpose of call, and then ask his/her last name. Then locate name on letter, find corresponding badge number of that dispatcher, and state that badge number to him/her. d. If dispatcher is satisfied that call is genuine, then_go ahead with required message. If not, you may also provide the initials of the dispatcher, which are on the letter. If that is not sufficient, you'should then ask for another dispatcher, and start over at step c. ISSUED FC/EPIP/01 MAY 0 61987 R5 05-06-87 .,+ ,,. ~. .+,

q v: ...s m j EPIP-OSC-16-1 i ($ Fort Calhoun Station Unit No. 1 EMERGENCY PLAN IMPLEMENTING PROCEDURE EPIP-OSC-16 l ^ EMERGENCY TEAM. I I I. PURPOSE The purpose of this procedure is to provide instructions for personnel assigned to Emergency Team positions. II. FRERE 0VISITE Both the primary and alternate individuals filling a particular Emergency. leam position, TAG No.1 through TAG No. 24,.have been fully trainec and are aware of their duties ano responsibilities. I III. PRECAUTIONS i None j j 1 I

V. PROCEDURE

(?:;. i Upon activation of the Initial Response Organization, those individuals assigned to a position on the Emergency Team will carry out their assignment as detailed in Appendix 1 of this Implementing Procedure. g !55UED MAY 151987 FC/EPIP/03 R5 05-15-87 7_ 7m . _ _....,,,y,.,,., ., ~

l I I EPIP-05C-16-2 Fort Calhoun Station Unit No. 1 EMERGENCY PLAN IMPLEMENTING PROCEDURE EPIP-OSC-16 APPENDIX-1 ( EMERGENCY TEAM A. TAG 1 -- RECORDER / PHONE TALKER Recor. ting Location: Technical Support Center Reports To: Site Director Basic Resoonsibilities: After receiving pager signal or a phone call to report to their emergency station, members will report to the TSC, pick up the proper tags fecm the TSC Tag Board and inform the Site Director of their presence. l If the pager system is inoperable, after receiving a call from the Call List Laller, the Recorder / Phone Talkers will contact each other to determine which one will immediately report to the TSC. The remaining gne will continue the Call List from the location where he/she was contacted until the call outs have been completed for the Emergency - 'y Action Level, or until notified tha' t no further calls are required. He/she will then also report to the TSC. When the Recorder / Phone Talkers report to the TSC, they will pick up the proper tag from the TSC Tag Board and inform the Site Director of their presence. Sets up, operates and maintains the tape recorder. Briefs the Site Director on 4ccide' t status and present conditions when n Site Director arrives. l Maintains emergency 109 book. Clerical assistance is available through the Security and Administrative Supervisor. i Performs telephone communications. Receives Operational Data from the Control Room and posts this data on the "FCS Emergency Status Board." Assumes responsibility for the Conference Operations (COP) Network from the Control Room when the Site D'irector assumes emergency response actions from the Shift Supervisor. Mans the Conference Operations (COP) Network and provides uodated in-formation to State and County officials as directed by the Site Director until transferring CDP Network responsibility to the E.O.F. Commt.nicator. ISSUED FC/EPIP/03 MAY 151987 RS 05-15-87 l i Lt_

.t. 3 ~.. i .i .c EPIP-OSC-16-3 'l ( [i APPENDIX-1 (Continued) { a i i 1 B. TAG 2, 3 - EMERGENCY RE-ENTRY TEAM j 1 Recorting Location: ) Operation Support Center Extension in Technical Support Center Building. Report To: Health Physics / Chemistry Supervisor when re-entry team required. / Basic Resoonsibilities: Re-entry Team personnel will report and receive instructions from the HP/ Chemistry Supervisor thru the Monitor Coordinator. Procures emergency kit, monitor kits, air samplers, and breathing ecuipment from storage area. { (' ' Obtains and battery checks high range survey instruments. -{ Obtains a set of protective clothing. Dons shoe covers and. coveralls, checks. .I out_and puts _on a TLD and high range dosimeter. Has other protective i clothing ready to don.on instruction from Monitor Coordinator. .j Checks out a self-contained breathing apparatus'for readiness to use. Checks j the mask for proper fit. I Prepares. for entry to the Auxiliary Building, verifies proper' dress _with the Monitor Coordinater. Enters the Auxiliary Building as directed by the Health Physics / Chemistry Supervisor and instructed by the Monitor Coordinator. Performs assigned tasks such as (a) search and rescue of injured person (s) - (b) emergency repair to equipment.and (c) assistance to Operations in 1 performing corrective actions. ISSUED (m, MAY 151987 FC/EPIP/03

RS. 05-15-87

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j EPIP-OSC-16-4 ) APPENDIX-l' [j (Continued) C. TAG 4 - DOSE ASSESSMENT OPERATOR-Reporting Location: Technical Support Center Building, Room 107 Recorts Tc: Health Physics / Chemistry Supervisor Coordinates With: Site Director until the Health Physics / Chemistry Supervisor arrives at._ l the TSC. i Basic Resoonsibilities: ?) 1. Sets up and establishes maps and overlays pertinent to the emergency conditions. 2. Establishes direct-line communications with the Control Room if l information is not already available in the TSC.to obtain meteorological and radiological data needed to perform Dose Assessment. FC-197 " Meteorological and Radiological Data Worksheet" will be used to record a this data. 3. Obtain manual Dose Assessment Data from the Control Room and enter the information into the computer prior to performing normal computer operations. 4 The Meteorological and Radiological Data Worksheet (FC-197) revision number and date will be verified by referring-to EPIP-EOF-6 Section H in the official set of Operating Manuals maintained in the TSC prior to their use. 5. Calculates airborne activity, dose rate and integrated dose for locations outside the plant structures and enters this data on FC-195. 6. Ensures the Health Physics / Chemistry Supervisor is receiving calculated data. i Assists the Site Director /HP/ Chemistry Supervisor on evaluation of radiological data as required. j ISSUED j FC/EPIP/03 MAY 151987 RS 05-15-87

~. 4 EPIP-OSC-16-5 f: s-' APPENDIX-1 (Continued) C. TAG 4 - DOSE ASSESSMENT OPERATOR (Continued), Maintains data in a'cu'rrent status. Refers to EPIP-EOF-6 Section H "Onsite and 0ffsite Dose Assessment using i the computerized program" when performing dose assessment duties with the computer. ~ Refers to EPIP-EOF-6 "Onsite and' 0ffsite Dose Assessment, for step-by-step Procedures using plant parameters and effluent monitors -to determine source term. Contact the National Weather Service in Omaha, telephone number-9-1-402-571-8351, and request projected meteorological weather-information if necessary to make long term dose-exposure projections. C 1 ISSUED MAY 15 G87 FC/EPIP/03 R5 05-15-87.' f + .u,w a =., yew ig gt yd = Tap'* mMf 8 *W M* 4 y e p .(, k= M+ p. 3 +* Vv # 4

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L EPIP-05C-16-6

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APPENDIX-1 I (Continued) 'v D. TAG 5, 6, 7'and 8 - 0FFSITE MONITOR R cortina Location: i Operation Support Center Extension in Technical Support Center Building Reports To: ' Monitor Coordinator Basic Responsibilities: -) Obtains monitoring kit, air sampler, water sampling bottles and vehicle. NOTE: . A set of keys for the vehicles are. located'inside each of the 3 offsite monitor team kits.. Xits are numbered to correspond with the OPPD vehicle identification number. ,,g -] %7 - t Informs the Monitor Coordinator / Dose Assessment Special.ist prior to any. departures. As directed, proceeds.to designated location and takes samples as assigned and analyzes air samples. Labels each sample and saves separately in plastic bags.in accordance with- ) EPIP-EOF-3,'" Emergency Instrumentation and Equipment". This procedure j des::ribes in detail how the samples are to be collected 'and analyzed. Reports results to the Monitor Coordinator by radio thru the radio operator, Tag 16. When the Recovery Organization has been activated..the offsite monitor. team will be under the control of the Dose Assessment Specialist I at the EOF.: Communicates with the Emergency Response Facilities on Channel No. 1, the j dedicated radio line for emergency field communication. Other channels can be used if problems develop. ' Review plant conditions and projected or known release-information with 1 Monitor Coordinator or Dose Assessment Specialist-periodically while.in plume sectors. ISSUED ) j FC/EPIP/03 MAY16l907 R5 05-15..

m. . 4 EPIP-05C-16-7 C APPENDIX-1 (Continuec) E. TAG 9 and GATE MONITOR Reporting Location: General Services Building Reoorts To: . Monitor Coordinator Basic Resoonsibilities: Notifies the Monitor Coordinator by telephone when he arrives at his work location to obtain any special instructions. Obtains available friskers from the-storage area, checks batteries and makes preparations for monitoring personnel and equipment. ~, l Monitors all personnel exiting the plant area: Paying particular attention to hands, feet and head area. Sends contaminated pe,rsonnel to the West entrance of the General Services Building for entry to the personnel decon station. Monitors all emergency team members'-returning from the plant closely for f contamination; properly be.gs anti-contamination clothing, if contaminated. Monitors all vehicles leaving the plant; paying particular attention to vehicle tires and top. Vehicles returning to the site will not be routinely monitored unless specified by Monitor Coordinator. r l e I i !SSUED ( MAY.151987 FC/EPIP/03 RS 05-15-87 J i $'I 4 ....s

lp -) EPIP-OSC-16-8 APPENDIX-1 'Dj -(Continued) F. TAG 11, 12, 13 and 14 - ONSITE MONITOR Reoortino location: f Operation Support Center Extension in Technical Support Center Building Reports To: Moaitor Coordinator Basic Resoonsibilities: .{ Obtains a full set of protective clothing including a full face respirator, . TLD and high range dosimeter. Obtains a survey instrument, clipboard, pencil and survey maps from the emergency locker. Reports to the monitor assembly room and informs the monitor Coordinator of S), his arrival. Informs the Monitor Coordinator of any departures from the TSC. ~ Maintains survey of on site affected areas. One monitor team will use a vehicle for survey and inspection, if required. May perform in-plant surveys as directed by the Site Director or Monitor Coordinator. Visually inspect the owner controlled area for personnel occupancy. Owner controlled area consists of the property within the site boundary and the strip of exclusion land directly across the Missouri River which can be 4 viewed from the screen house. Performs habitability check of the Guard Building, Storeroom and a survey of the General Employee Training Building. If the early warning strens have sounded (site area and general emergencies) instruct all non-emergency workers to-vacate the owner controlled area. If a site or public evacuation has been declared, report all refusals to i vacate by non-emergency workers immediately to the Monitor Coordinator who will notify the appropriate county sheriff. ISSUED .y FC/EPIP/03 MAY 151987 R5 05-15-87 z...

s 9 EPIP-0SC-16-9 r. I APPENDIX-1 (Continued) F. TAG 11, 12, 13 and 14 - ONSITE MONITOR (Continued) After initial surveys and inspection have been completed, report back to the monitor assembly. area. to receive more directions. from the Monitor Coordinator. for continuing surveys. Return all sample and survey results to the Sample' Counter / Dosimetry. Issuance ET TAG 21. Use dose rate sample log (Figure EPIP-OSC-13.1) to record all dose rate l survey data which includes time of. survey, location; type of survey. i.e. Beta or Gamma; dose rate in mrem per hour; and name of surveyor. Also at the top'of the form, fill in the date'and the instruments used and serial ~ numbers. ! S S U'E D 15 W FC/EPIP/03 - R5 05-15-87 +

6 EPIP-OSC-16 APPENDIX-1 h (Continued)~ G.- TAG 15 - MONIT,0R COORDINATOR Reportino Location: Technical Support Center Building. Room 107 Reports To: Health Physics / Chemistry Supervisor Directs / Coordinates: Emergency Re-entry Team Offsite Monitor Gate Monitor-Onsite Monitor Radio Operator Rescue Squad Monitor p:i-Personnel Decontamination tu Outside Coordinator Basic Resconsibilities: Is responsible initially to the Site Director for all monitor team. activities. When the HP/ Chemistry Supervisor position.is manned, Tag 15 shall report directly to this individual. Ensures that the TSC Emergency Status Board indicates current conditions and-assessments. Ensures that all surveys and data are documented 'in' the field by the monitor teams and delivered to the TSC. Coordinates in plant first aid assistance as needed.- Ensures that all onsite personnel have been' checked for contamination. Ensures that a radiation survey of the Control Room, OSC AND TSC has been conducted. When the Recovery Organization ha's been formed and. the EOF. activated,-- the-Dose Assessment Specialist will assume responsibility for the Offsite - Monitor Teams. ISSUED '. l ~ FC/EPIP/03 MAY 151987 R5 05-15-87

s EPIP-OSC-16-11 [', APPENDIX-1 (Continted) 3 G. TAG 15 - MONITOR COORDINATOR (Continued). In coordination with the HP/ Chemistry Supervisor, directs onsite and offsite teams when to. don protective clothing. I Ensures habitability checks are performed in the Guard Building, Storeroom, General Employees Training Building and the helipad as necessary. Ensures all radio transmissions during drills and/or exercises start and end with "THIS IS A DRILL MESSAGE" or "THIS IS AN EXERCISE MESSAGE". Ensures off-site and on-site monitor teams are briefed on changing plant conditions and known or projected release information, periodically when teams are in plume sectors. Maintains dosimetry log on each of the off-site monitor team members. Records, on the. monitor locations map, monitor team location, survey results and time surveys were taken for each of the field monitoring teams (OPPD teams and state teams). C t 1 I ISSUED ('. FC/EPIP/03 MAY 151987 a5 05 15.s7 i

a,.. a EPIP-0SC-16-12 APPENDIX-1 "l-N (Continued) H. TAG RADIO OPERATOR Recortino Location: Operation Support Center Extension in Technical Support Center Building. Reports To: Monitor Coordinator Basic Responsibilities: Establishes and maintains radio communication with Offsite Monitor Teams, and the access road Security Guard from the TSC Building. NOTE: The Offsite Monitor Teams will use dedicated radio Channel-No.,1 for communication. Records in writing, all messages received so that the information c( ) is distributed and/or transmitted by FAX. Transmits radio messages as directed to field Emergency Team Members. i Authorizes onsite entry of the Recovery Organization _ and emergency response personnel from prepared list. Relays entry authorization. of other personnel as directed by the Monitor Coordinator. Ouring drills and/or exercises ensures all radio transmissions start and end with "THIS IS A ORILL MESSAGE" or "THIS IS.AN j EXERCISE MESSAGE" as appropriate. o IssuEo d MAY 151987 FC/EPIP/03 Rs 05-15-87

t EPI'P-05C-16-13 t APPENDIX-1 (Continued) I. TAG 17 - MESSAGE DISTRIBUTION / CLERICAL SUPPORT Reporting Location: Operation Support Center Extension in Technical Support Center Building. Reports To: Security and Administrative Supervisor Basic Responsibilities: Reports to the TSC and checks in with the Security and Administrative Supervisor. Responsible for the collection and distribution of message traffic within the TSC. Performs other administrative duties.as required. 1 l 4 e h ISSUED . FC/EPIP/03 MAY 151987 )R5. 05-15-87 l

.w 4 - 1 x a i EPIP-OSC-16-14 w APPENDIX-1 'd, (Continued): I 1 I 'J. TAG 18 - RESCUE SQUAD MONITOR a Reoorting Location: Operation Support Center Extension in Technical Support Center Building 3 l Recorts To: .j Monitor Coordinator. J Basic Resonnsibilities: 1 Receives instructions from the Monitor. Coordinator i.e. name and location of injured personnel and'1f.the injured person is : contaminated. Obtains a radiation survey instrument and.' performs operational check. Also, 1 obtains a personnel air sampler. .jg j My' ] Obtains four (4) high-range, zerced pencil dosimeters. ] Meets the Rescue Squad and issues a dosimeter to each member'and ensures they ) are dressed in protective. clothing if the injured person is contaminated. i Briefs them on the location and probable condition of any casualties. ) Briefs the. Rescue Squad on radiation hazards 'and other precautions.to be taken. NOTE: The Rescue Squad personnel do not nonnally enter the. auxiliary building unless personnel injuries dictate this' entry is'necessary. The squad will normally be met with the. injured personnel at -the north emergency exit. Accompanies the Rescue Squad to pick up casualties and provide, radiological coverage during the trip to; the hospital. Furnishes hospital personnel with the following information,. if. known: (a) Types and extent of. radiation exposure. I (b) Levels of external contamination. lSSUED '7 -.W FC/EPIP/03 -MAY 151987 ,Rs 05-15-87 m .....-4 3--- ra ~- 4 - -

EPIP-0SC-16 15 APPENDIX-1 (Continued) l i 1 J. TAG 18 - RESCUE SOUAD MONITOR (Continued) (c) Probability of interr.al contamination. Collects, reads and records pencil dosimeters from Rescue.5 quad personnel. Ensures that the Scuad members, vehicle and eouipment are free of { contamination prior to release. { I 1 i l b J al 1 1SSUED k MAYIUIbb7 FC/EPIP/03 R5 05-15-87 )

e- - m y' 3, e o 4 EPIP-OSC-16-16 '/ APPENDIX-l' (Continued) ~ t K. TAG 19 - PERSONNEL DECONTAMINATION Reporting location: .0peration. Support Center Extension in Technical Support Center Building Alternate Location: General Services Building Reports To: Monitor Coordinator Basic 'Resconsibilities: Reports to the Monitor Coordinator for a briefing.

gd Will aid in the decontamination of personnel using facilities specified.by the Monitor Coordinator.

The following equipment w.ill be available for use: I (a) Frisker (RM-14/15/19) { (b) Step-off pad with undress area - (c) Containers for contaminated clothing.. NOTE: Each individual who is contaminated or who has. contaminated clothing must have this clothing bagged individually with the-person's name and the time placed on each bag. (d) Cleaning materials (soap, brushes, towels, etc.)- ) (e) Clean clothing (i.e., paper coveralis and shoe-covers). Uses procedure EPIP-EOF-10. " Personnel Decontamination", to _brief personnel j on decontamination methods to use'(i.e., complete shower, wash hands. etc.). ll i I-lSSUED 0.3/ FC/EPIP/03. MAY_151987 RS 05-15-87; l J 1;.

EPIP-05C-16-17 / APPENDIX-1 (Continued) K. TAG 19 - PERSONNEL DECONTAMINATION (Continued) NOTE: Complete showers should be avoided unless absolutely necessary to prevent spread of contamination to other parts of individual's body. Keeps Monitor Coordinator briefed on personnel decontamination status. Records names and survey results, initial and final, of personnel admitted to decontamination station. ( lSSUED FC/EPIP/03 MAY 151987 as _05 33 37

EPIP-OSC-16-18 .~. APPENDIX-1 (Continued) L. TAG 20 - CONTROL ROOM DATA COLLECTOR Reporting location: FCS Control Room Reoorts To: Operations Support Manager Basic Resconsibilities: 1. Collect operational, meteorological and radiological data from instrumentation and computer equipment in the Control Room. 1 2. Information collected is recorded on appropriate Fort Calhoun forms. (a) Complete all items, as appropriateLfor accident, on FC-194 If a .-s .particular item does not apply enter "N/A". "3 ., a (b) Complete all items in Sections I and II, as appropriate for accident, on FC-197. For items that do not' apply enter "N/A". 3. Deliver completed forms to; the Control. Room communicator'(phone talker) who will in-turn relay the information to phone talkers in the OSC,' TSC and EOF. 1 I 4. Data must be collected and provided to the C. R. communicator every 15 i minutes. i .) 1SSUED ANI5100I FC/EPIP/03 R5 05-15.,., -,_

-d- 't EPIP-05C-16-19 m K f. APPENDIX-1 (Continued) ) l M. TAG 21 - SAMPLE COUNTER / DOSIMETRY ISSUANCE Reoortino Location: Operat_ ion Support Center Extension in Technical Support. Center Building. 3 Reports To: 1 Monitor Coordinator 1 1 Basic Responsibilities: 1 Ensures that all team members needing TLD"s and dosimeters have them. -l NOTE: All dosimeters will be zeroed before'being issued. i Maintains the TLD/desimeter log. I Sets up a counting station and counts all samples brought into the TSC and c reports results to the Monitor Coordinator. Ensures that all samples-are saved and labeled for future counting if needed. Collects radiation monitoring devices from team members 'as they. return from assigned tasks-. i Keeps the Monitor Coordinator informed on counting results/ personnel exposure. 1 1 i [-- lSSUED j a MAY-151987 FC/EPIP/03 -RS 05-15-871 l i y

l i EPIP-OSC-16-20 APPENDIX-1 (Continued) N. TAG 22 - OUTSIDE-COORDINATOR .) Reportint; Location-I General Services Building i Reoorts To: 1 Monitor Coordinator l Supervises: All personnel that have evacuated the plant Basic.Resoonsibilities: Contacts the Monitor Coordinator in the.TSC by telephone for instructions. Coordinates outside activities in the vicinity of the entrance gate and general assembly area during NOE and Alert Emergenties. Reports to the EOF for re-assignment during " Site Area Emergency" and " General" Emergencies requiring site evacuation. Ensures that all personnel that have evacuated the plant are in two groups: 4 i (a) Personnel exiting from the auxiliary building area j or (b) Personnel exiting the uncontrolled areas of the plant. 1 Ensures that all contaminated personnel are sent to the emergency decon station. Ensures that all vehicles leaving the plant area 'are monitored (except emergency vehicles). 3 1 Reports names of any contaminated / injured personnel to; the Monitor I Coordinator. ? * : '.'.~ D FC/EPIP/03 MAYI51987-R5 05-15-87 r-.. - n-

~ _.., j EPIP-OSC-16-21 -' ~ . APPENDIX-1 W (Continued) O. TAG 23 - EOF INFORMATION SPECIALIST q j Reporting Location: Emergency Operations Facility j Reports To: Site Director until the Recovery Manager takes full authority i Coordinates: EOF Technical Liaison Media Release Center 4 Site Director { Recovery Manager when Recovery Organization is activated j Responsibilities: j j Monitors status of emergency and relays timely and accurate information to j the Media Release Center (MRC). Maintains information time log for post emergency reference. Refer to M.2.6.2 for primary respons.ibility as related to the Recovery. Organization. i l Refer to EPIP-RR-40 for reporting assignment and basic duties, i i i i i 1SSUED i MAY 151987 FC/EPIP/03 R5 05-15-87 L

@@'e 'tI 4 ,f vs ee (* q EPIP-OSC-16-22 APPENDIX-1 (Continued) ) I P. TAG 24 - EOF TECHNICAL LIAISON Reporting Location: ) Emergency Operations Facility -l i Reports To: EOF Information Specialist j i Responsibilities: Monitors status of emergency and assists EOF Information Specialist in collecting and interpreting nuclear-related data. I Serves as the EOF contact for technical liaison assigned to assist official spokesperson at the Media Release Center'(MRC). .g Refer to M.2.6.3 for primary responsibility as related to the secondary j organi:ation. Refer to EPIP-RR-41 for reporting assignment and basic duties. I I .i I !SSUED. FC/EPIP/03 RS 15-87 m.

..a = o EPIP-OSC-18-1 i Fort Calhoun Station Unit No. 1 ~ EMERGENCY PLAN IMPLEMENTING PROCEDURE EPIP-OSC-18 Initial Resoonse Orcanization Notifications I. PURPOSE The purpose of this procedure is to detail-assignment and responsibilities and provide instructions to be followed by the individuals-filling the position of Call List Caller. This orocedure also provides instruction to the Initial Response Organization as to their response to a notification. II. PREREOUISITE-1. The Emergency has been classified ir, accordance with EPIP-OSC-1. 2. Notification has been received from the Fort Calhoun Station Shift Supervisor, or his designee, to initiate the " Initial Response Organization Call List". 3. The " Initial Response Organization Call List" is available in the North Omaha Station Shift Supervisor's office, at the Fort Calhoun Station, or with the E.T. Tag # 1 members / _4 The " Initial Response Organization Call List" utilized shall be the latest revision or issue date available at that location. III. PRECAUTIONS 1. No member shall have more than one pager assigned to-the Initial Response Organization in their possession at one time nor shall a member be simultaneously responding for more than one position. I

V. PROCEDURE

1. Upon activation of the Emergency Recovery Organization, those' individuals assigned to the position of Call List Caller shall carry out thei.r assignment as detailed-in Appendix 1 of'this i implementing procedure. 2. Upon notification,' the members of the Initial Response Organ-ization shall respond as detailed in Appendix 2. i i h 1SSUED FC/EPIP/01 R2 05-15-87 4 i

r.... 1 EPIP-OSC-18-2 i Fort Calhoun Station Unit No. 1 EMERGENCY PLAN IMPLEMENTING PROCEDURE EPIP-OSC-18 ] APPENDIX l' Call List Caller J A. Personnel Assignment -(Job Title)' North Omaha Station Shift Supervisor North Omaha Station Operators 1 S. Reporting Location Normal Work Station C. Reports To. Site Director-(Initially the Fort Calhoun Station Shift Supervisor)'. D. Primary Responsibility Makes calls to members of the Initial Response Organization when an emergency is declared at Fort Calhoun Station. E. Basic Duties 1. Upon notification from the Fort Calhoun Station Site Director or ~ designee, the Call List Caller will initiate calling the Initial Response Organization. 2. The Call List Caller will follow the order and special instructions on the Call List when making outgoing calls. 3. Outgo;ng calls made to Initial Response Organization members will I begin with the normal work phone or home phone number, depending' ~ on the time of day, then will proceed 'to individual's pager number, if applicable. 4 The Call List Caller will inform the Initial Response Organization member called of the Emergency Classification and reouest the member report immediately, s 5. The Call List Caller will note the time each is contacted by telephone. l 6. The Call List Caller will call ~ the' Fort Calhoun Station Site-Director and report the status of completion of the Call List.' i i S S U 'E D 'l ~ "- u FC/EP!P/01 MAY 151987 R2 05-15. --,.x s

.~ i EPIP-OSC-18-3 ( i APPENDIX 1 ~ Call List Caller F. Substitute Duties i 1. If the Call List Caller has been previously informed that the pager I system is inoperable, the Call List Caller shall continue with the remaining steps under this section. 2. Upon notification from the Fort Calhoun Station Site Director or his designee, the Call List Caller will initiate calling the Initial Response Organization. 3. The Call List Caller will follow the order and special instructions on the Call List when making outgoing calls. 4'. Outgoing calls made to Initial Response Organization members will begin with the normal work phone or home phone number, depending on the time of day, and then will proceed to the individual's pager number, if applicable. 5. The Call List Caller will inform the Initial Response Organization J member of the Emergency Classification and request that the member j report immediately to their Emergency Response Station. 6. The Call List Calker will note the time the individual is contacted. 7. The Call List Caller will call all the individuals listed on the Call List for each position. 8. The Call List Caller shall continue the Call List until two (2) of the individuals listed for the TSC-Recorder Phone Talker (E.T. Tag No. 1) have been contacted who can respond to the emergency. Both of the TSC-Recorder Phone Talkers will be informed by the Call List Caller as to the status of the call list. The second TSC-Recorder Phone Talker shall also be informed as to the name of the other responding TSC-Recorder Phone Talker. 9. If the Call List Caller is unable to contact two (2) of the individuals listed for the TSC-Recorder Phone Talker (E.T. Tag No. 1) who can respond, then he/she shall continue the Call List until the STOP point is reached for that Emergency Classification.

10. The Call List Caller will call the Fort Calhoun Station Site Director and inform him when the calls have been completed.

l FC/EPIP/01 MAY 15 G87 R2 05-15-87

1 EPIP-OSC-18-4 Fort Calhoun Station Unit No.1 EMERGENCY PLAN IMPLEMENTING PROCEDURE EPIP-05C-18 APPENDIX 2 Initial Resoonse Orcanization A. Personnel Assignment Initial Response Organization members B. Primary Resoonsibility Respond when paged or called by the Cail List Caller, TSC-Recorder Phone Talker (E.T. Tag Team No. 1), or Emergency Planning Group personnel. 1 C. Basic Duties 1. When contacted by pager, members of the Initial Response Organization will carry out the response as applicable for the numeric code as received on their pager. CODE CONDITION REQUIRED RESPONSE 2-2-2 EMERGENCY AS CLASSIFIED REPORT IMMEDIATELY PER EPIP-05C-1 TO ASSIGNED EMERGENCY ) RESPONSE STATION i 8-8-8 NOTIFICATION TEST TELEPHONE THE EOF 1 2. If an individual, is contacted by telephone, he/she will report immediately to the designated Emergency Response Station. 3. If the page is due to a Notification Test, the member shall call the number listed. If the line is busy, the member shall continue attempting to contact the number. Once the phone is answered, the j member will give the person answering the phone his name, Initial Response Organization position, and other information which may be required for completion of the test. 4. The members shall maintain the pagers in an operational-condition and be familiar with their use. The members are responsible to keep the pagers on their persons or within audible range [one hundred (100) feet] of themselves. 5. Pagers are assigned based upon the number of individuals needed to fill each position of the Initial Response Organization in the event of an emergency. The actual carrying of pagers will be done on a rotating basis between all members of an assigned position. The rotation schedule is to.be completed by the members themselves and should include the current month along with the two subsequent months. Vacations.and any time when a member would be outside of the pager receiving area (45 mile radius of Omaha) should be taken into account when completing this schedule. iS3UEO FC/EPIP/01 R2 05-15-87 MAY 151987 -~

4 6 1 l 3, j.' 7 ~; .a /_;- j, L. - g. Q ~ x ':.~.. x ' N.: 'x 3 u.;._2.= a I, F EM PURP eor rt GEN OSE C q C alho u A. Y Sq P LuAn S Th TE N r is t CHNICAL TS IM at EPIP PL m edu pr i A s EoMn U oc E ay nda ENT r.i that havt edur gate n SUP C-8 INt N "'~ e P da e e G ORT P sti co a wi fm o locc m R.1 lo ge atentai of l OCEDURE IP T N CENTER or w f Co EP ir yi r s ns chart.st,el ed r F of f e AST d cu o D S t S as he ri the ur am T C81 yste RANSI detequing a me t se e m anype cti as f e (Pm A )E N: minckeac ol ur r ons l T st ci nd Co a SS ow e r nd de to ed a nt (EP te e ed Ob o /o nt degr pr E~ a f mpe da b tai fr nd may a r o

m. th m si ee vi IP T sti sam e

st ati cor r e st atu ge co n o tu of de ntai a SCmat pl gui acc ing,c i. r SLOW ct an indepe es i bnm 8 C)e e on e (EP tr but e e s da TRANSIEN e nt cor u u. (EP IP-T da e nce ate Us r m nde r a e sin r stim ad nd an 2P T SC theage nt e d si e co s i iamg b nts.ul ated ati s ge the el assesse nd S 8 D)T: a shand w sho ts iti ow otopi C 8 A) ich on b P 5 7e ul a u Us a o nd e e si st nd me ct y o r n do asdb a s c u datasin Ac th nt io n se 'P u n g r e pe f cor ot Co n !P T e g ci e of s em u r con e e or r ed mi e at S nt (s r co de at t an tai exi xpe Ee an al nta nt ta C 8 B)EP to,'b eheaP rl ent ther ctedxi an t nm t Sa ched al ysis (in t IP pr o T ysis (E e mpl m to T tai oy to he oc r m EP nt S n st e i m b edu ocou IP P T dr [(' N adioco e r stim ati upl v 'et e nhaC 8 C)s P I h ng r e a Ac y ci ate mpl e i is nce a y pl TS C 8 ge S o d on e r To es e e nt of do te a n d7I A d nd c (C )C 8 A),8)n ivi ea c f se mpe Sa cor r ur ET d rl is or ty oE - mpl e a ate i ed i r otopi ntai g da te atu pe co er r i nd ad Co i e n a nto m r f da h S age an es or 'B degringsntain the ge data c ma nm al e yste. y nt fr me f i W sis T-ee om nt oll hm al ydr (P e [' 1 h a ndin ot M0 s wi atiysis geo A )n p Do o c ge. opi Is of R SS cor dataotopi 9e c Rate ng f on n 1A [L e s. c da a our to Data age.ndRnal se i,. m A sf Co dete An M 0 ysi ctio 918s o ydrh ntai min ysi to r al n s u s: nme ses e o e F s nd ge a n nt the uses im e {( st ge' Hn ydr ty ,r degr te Exi ofatedoge pe po ate er e ee n a st p in Me nd ac t mper T i degrci t'. cor asur ee nt he co de r G e co r e tu mo da nt em r da e co age.cha upl age.ai emt m n of m n nt dete e e (C ) T nge r i ET a s nd min \\ n r e th@ Varr ei s e se

.s .8 e* + EPIP-TSC-8-2 .ng ). TYPE OF TRANSIENT I 4 FAST SLOW TAANSIENT, TRANSIENT .i i 1 j ~ RACIATION ISOTCPIC CCAE EXIT-AADIATION DOSE RATES ANALYSIS THERMCCOUPLES CCSE.AATES EPIP-TSC A EPIP-TSC-8-B EPIP-TSC-8-0 EPIP-TSC-8-A HYOROGEN. ISOTCPIC 1 MEASUREMENT ANALYSIS EPIP-T SC C EPIP-TSC-8-S l CCAE EXIT 'HYOROGEN THERMOCOUPLES MEASUREMENT EPIP-TSC-8-0 EPIP-TSC C t-EPIP-TSC-8 g CORE DAMAGE ASSESSMENT PROCEDURE. FLOW CHART-1SSUED MAY 151987 R4.5-15-87 'N*. 1 4 ) ) (

4 4 EPIP-TSC-8-A-1 SECTION A Estimate of Core' Damage Using Containment Radiat_ ion Dose Rates I. PREREQUISITES A. A plant accident with the potential for core damage has occurred. B. Wide Range Radiation Dose Monitors RM-091A and PM-0910 are orcrable-and able to measure arec dose rnes in concainment resulting from fi:skn products dispursed 1.n the building atmosphere and plated out on the building. surfaces. II. PRECAUTIONS AND LIMITATIONS A. The total quantity of fission products measured at monitor locations in containment may be changing rapidly due to transient plant i conditions. Therefore, multiple measurements.should be obtained within -l a minimum time period and under stabilized plant conditions when possible. Samples obtained during rapidly changing plant conditions i should not be weighed heavily into the assessment of core damage. RM-091A and RM-0918 have an upper radiation reading limit of 1E7 R/HR B. This procedure only provides an upper limit estimate of the progres-sive core damage and cannot accurately distinguish between the condi-tions of fuel cladding failure and fuel overheat when the'resulting (, dose rates are the same. This procedure does not attempt to identify the' extent of any potential fuel melting. j i C. This procedure is intended for.use when the fission product inventory in the core has had. sufficient time to reach equilibrium. Based upon' the fission products of concern, equilibrium conditions are achieved after 30 days of operations at constant power. Reactor power is. considered to be constant if it has not changed by more than 210%. This procedure can be applied following non-constant periods of operation and when the reactor has produced. power for less than 30 days. The assessment of core damage for-less than 30 days of opera-tion will underpredict the actual conditions. ( lSSUTU MAY 151387 R4 05-15-87 FC/EPIP/02

.~ 4. EPIP-TSC-8-A-2 II

I. PROCEDURE

's t -1 A. Record the following plant indications. Additional or fewer radiation dose rate measurements may be recorded as a function of data available. CONTAINMENT BUILDING Radiation Dose Rate Rads /hr. Time o'f Measurement Date' Time i J Radiation Oose Rate' Rads /hr. Time of Measurement Date Time-Radiation 00se Rate _ Rads /hr. Time of Measurement Date Time Radiation Dose Rate Rads /hr. Time of Measurement - Date Time .[Y a Prior 30 days power history: Power, Percent. Duration,' Days W B. Time of reactor shutdown Date' Time i 'i i ISSUED MAY 151987 g4 os.15 87-FC/EPIP/02 J e

8 r.

? gn -#es6g p e a ,4- + w r , /> p

l.. EPIP-TSC-8-A-3 L II

I. PROCEDURE

C. Plant' Power Correction The measured radiation dose rate inside' the containment bu'ilding -is to be corrected for the plant power history. To correct the measured dose rate-to the corresponding value had the plant been operating at 100 percent power, follow the guidelines below. For operation at constant power for more than'30 days, apply a simple ratio of' power.. Reactor power. is. considered to -be -constant if it has ' not changed by more.than :10%.- To corres.t the radiation dose rate for the case in which reactor power level has not remained constant.during the 30 days prior to the' . reactor shutdown,. engineering judgement is used.to determine the most representative power level. The' guidelines-below should be used in. making this determination. The average power during the 30 day period;is Lnot necessarily the most . representative value to be used'for the correction'to equilibrium conditions. The power levels at which the reactor last gperated should weigh more-heavily in the determination than the earlier power levels. l ("s' Continued operation for an extended period should weigh more' heavily X than brief transient periods. For operation less than 30 days this procedure can still be employed, however, the estimate of core damage is expected to.be an under- ) prediction of actual co~nditions. 'l 1 i 2 i e,. [Q f S S U E 0. FC/EPIP/02 MAY 151981 R4.05-15-87 w, - n ~.. ~.

4 EPIP-TSC-8-A-4 II

I. PROCEDURE

(Continued) D. Apply the following equation to determine tne radiation dose rate corresponding to equilibrium full power source inventory conditions. Equilibrium, Measured X 100 Dose Rate 00se Rate % Reactor Power Level {

dose rate correction factor E.

To determine the decay correction for the radiation dose rate, record \\ the time between measurement of the dose rate and reactor shutdown, j recorded in Step 8. I Time duration for this measurement = l hours F. Estimate the extent of core damage using the equilibrium dose rate (step III.0), the duration of reactor shutdown, and the analytically determined dose rates provided in Enclosure 2. judgement to determine which category of core damage shown onUse engine is more representative of the plotted value. following criteria when making this determination-Consider the periods of transient conditions within the plant.Some do 1 Measurements made during stable plant conditions should weigh more heavily in ..s { the assessment of core damage. { major fuel overheat may indicate concurrent fue This procedure pellet melting..can.not be employed to estimate the degree of fuel ] 1 Fuel cladding failure should be anticipated for dose rates within any category of fuel overheating. This procedure can not dis-tinguish the relative contributions of the two categories to the total dose rate, but does give an estimate of the highest cate-gory of damage. Dose rates corresponding to the two categories of major cladding failure and initial fuel overheat are observed to overlap on. required to distinguish between them.The evaluation of other plant p However, concurrent conditions may be anticipated. ISSUED FC/EPIP/02 MAY 16 I307 R4 05-15-87

1 4)il) 1 0 0 5 5 s 1 0 0 5 fd 1 n 1 n oo n 0 a a A R a n 5 h n h 7 t h a T a T 8 8 ne T h o h eg T t r T r 5 C ca s e e 1 S rm s s 0 t s t 0 T ea e s 1 a s a 5 3 PD L e e e e 0 P L r L r p I G G u P 4 n E e rR r u t o B n n C n e F o o d me F F i ti n eg 0 t nt 1 1 a rn 0 0 0 a ea18 5 e ua 5 0 3%d l d3l 6 a g sR 5 7 22i ai<< 7 i 77 x vx a a l l s m e 0 iO p a M u D q 0 E 0 l 2 e l u c o F i t tt e e s e f sn r l e a t n r o ie o pr G no u rm C uu f ei s s ee ot ondl t) s e t r m ca eeaae e i cu u or t gcvdr r r as m me nouiio P o ra it rp urduxC g ae xi em od o q0 r e hM ax he myrE f o t C ME TT AHP(1o f a i C n n l e E C o a v R R i ro i U N t ut G 0 S o a t O f me t z ce s L o sg i uu e C ia e r rD u N s nm u u t l E c aa D s S yd n a i hD s t ao V t c ep e fili s ef ra r orCt Mo uG p g a ir e t r seld p e n ps e st ei o t o ua v onux n c M RG O lIF0 r ar u a e B h r) 0 0 C uF 0 5 e l t* 8 3 u e a( 0 1 3 f g r 5 a ee 7 0 0 d m pg s 0 0 n a mn 2 8 a D ea 1 1 TR e d h ) a ru l s C t s e e l t r e g e l e P yg n r e l ra e i u g P l r om g d l n e o ga a d' ei i l e P t eD m a t a d eg tt g g c t a l aF d un aen l n a al D C i a Fi ili ei e Ce dg l t dl t ut R u l l e en Ce l a eea fa CF e ar ni r ae nPe e n i i R u u rd ru ih r h rh o i Nf F tl ed ol t r el r or o i i t a ji ie t ee je s o na nl aa nv nuv av d N I F I C MF O IFO MO n I ep 0:. e l 2 3 4 5 6 7 'D . ""' u C m U s mN g c. gN n 2

EPIP-TSC-8-N-6 ENCLOSLfE 2 FORT CALHOUN IN-CONTAINMENT POST ACCIDENT DOSE RATE 1 0.9X106, i 1x10 0 f 4 t, 'O o k /, 4 g 'o,, 5~ b W. 4 e4 4 0 1X10 /y, ,#'t c< o,, 0, 0 '4g 's+4

Ci

'4sg O l f 3 1X10 i 1 l 1 10 100 100 TIME POST ACCIDENT, HOURS i' t, ISSUED !f l R4 05-15-87 3 MAY 151987

m y I ' + 3,ms,1 4 4 EPIP-TSC-8-B-1 . SECTION B Estimate of core Damage Using Isotopic Data I. PREREOUISITE! -] A. A plant accident with the potential for core damage has occurred. B. Isotopic activities are available per 01-SL-2 or 01-PAP-2. C. Computer program UTYPASS 1.s available for execution. D. Tha Post Accident Sainpling System (PASS) is operable and has the capbility to obtain and analyze the identity and concentration of fission product'isoteres.in' fluid samples which have the potential to be highly radioactive. II. PRECAUTIONS AND LIMITATIONS The assessments of core damage obtained by using this procedure are only ~ estimates and represent lower limit +1 mates.of clad damage. The total quantity of fission c' odur's 4 easured at monitor locations in containment may be' changing rig h 4 4 s to transient plant conditions. Therefore, multiple measurements sold be obtaineo within a minimum time period and under stabilized plant oonditions when possible. Samples 4 obtained during rapidly inanging plant conditions should not be weighed heavily into the assessment of core damage. II

I. PROCEDURE

Estimation of Core Damage.Using Isotopic Analysis A. Perform an'isotcpic specific activity analysis by obtaining _ samples' j based on the following criteria. 3 TYPE TRANSIENT-SAMPLE LOCATIONS a) Rapid Depressurization Reactor Coolant, Containment of Primary System.................. Atmosphere, Containment Sump b) Slow Depressurization of Primary System j EARLY......................... Reactor Coolant LATER......................... Reactor Coolant,. Containment Atmosphere, Containment Sump-1 ISSUED a FC/EPIP/02 R4--05-15-87' i j

4 EPIP-TSC-8-B-2 1 2 'II

I. PROCEDURE

(Continued) ] ~ NOTE: If the core damage estimate is performed after purging containment, the noble gas concentrations should be adjusted to reflect the initial concentrations prior to the purge operation. 3 1 B. Using 01-51.-2 or 01-PAP-2, obtain and analyze the selected samples for 1 fission product specific activity and record in Table _1. -All of the isotopes listed in the table may not be observed in the sample. C. Record the time of' reactor trip. Date / Time O. If the Safety injection Tanks (SIT's) and/or Safety Injection Refueling Water Tank-(SIRWT') have been used during the transient, complete Table 2. E. Determine the elapsed time between reactor trip and sample measurement: hours F. Complete Table 3 on Power History. G. This step uses the computer _ program UTYPASS on.the IBM PC to provide an estimate of' core damage using isotopic analysis. If access to the IBM PC is not available, proceed to Step I to access-the program on the termtnet. 1. Insert the UTYPASS floppy disk, found in the Core Physics Supervisor d). emergency packet, into the "A" floppy drive, 2. Enter the following statements, in order: CD\\EPIP CR (Carriage Return) COPY A: CR PE CR CR EDIT USERI.PRG CR 3. Input the information from Tables 1, 2 and 3 by using the arrow, P90n and PgUp keys on the right side of the-keyboard to modify _the input data file as outlined in Appendix A. 4 Enter the following statements, in order: FILE USERI.PRG NOTABS CR QUIT CR PRGPASS CR l '.I ISSUED y MAY 151987 l FC/EPIP/02 R4 0s-15-87 ) l 1 n f

y;' EPIP-TSC-8-B-3 . c. II

I. PROCEDURE

(Continued)- 'Y G. 5. When "Stop - Program Terminated." appears on the screen, enter the 'following statements: PE CR EDIT USERO.PRG CR 6. Use the arrow, PgDn and Pgup keys to view the output, or use the F7 function key on the left side of the keyboard to pr. int the output file. 7. When finished, return the floppy disk to the Core Physics Supervisor emergency packet and type the following: 0U11 CR QUIT CR CD\\ 'R H. CORE DAMAGE ASSESSMENT The conclusion on core damage'is made using the three parameters developed above. These are: 1. Identification of the fission product isotopes which most characterize a given samp1', step 8. Table 1. e 2. Identification of the source of the release from program output. 1 or step L if manual calculation was used. 3. Quantity of the fission product available for release to the l environment expressed as a percent of source: inventory from program output or step 0 if manual calculation.was.used. Compare the three parameters above to the definitions of the 10 NRC 1 categories of fuel damage found in Enclosure 1.- Core damage is not anticipated to take place uniformly. Therefore when evaluating the three parameters listed above, the procedure is' anticipated to yield a combination of one or more of the 10 categories defined ir. Enclosure 1. These categories will exist simultaneously. CONCLUSIONS: 1 iSSUEB j FC/EPIP/02 MAY 15 ISS7 R4 c5 15 87

.+ EP!P-TSC-8-B-4 II

I. PROCEDURE

(Continued) .~ I. This step uses the computer program UTYPASS on the-terminet to provide 'J an' estimate of core damage using isotopic. analysis. If' access to UTYPASS or the terminet is not available.. continue with procedure Step J for manual calculation toward an isotopic core damage estimate. 1. To use the Combustion Engineering computer system upon which UTYPASS resides, SET TERMINET TO HALF.0UPLEX Dial 1-800-243-3202 or 1-203-683-0411' or 1-203-683-2734 2. Once the computer has been accessed, use the following sequence to' l log on and access UTYPASS. USER NAME, PASSWORD: OMAHA, 0 CR I XEDIT,UTYPASS,P CR Perform input changes as outlined in' Appendix A Q,,RL CR j 3. To submit,the computer job, type: SUBMIT, UTYPASS.T CR 4. When the job ts complete, use the following command 'to identify the job name'. ENQUIRE,UJN CR S. To print the results, input the following sequence: TRMDEF,PW=136. CR QGET,LfN CR COPY,LfN CR 6. Summarize the core damage results as directed in procedure Step H. 1 issueo i d 1 FC/EPIP/02 MAY 151987 R4- 05-15-87 4 = e a 9

l y ) l EPIP-TSC-8-8-5 { ~s l III PROCEDURE (Continued) . b-U J. Complete the Enclosure 3 worksheet to correct the measured sample specific activity to standard temperature and pressure as follows.. NOTE: This portion of the procedure-is only to be used when' access: to UTYPASS is not available. i 1. Reactor coolant liquid samples are corrected for system temperature and pressure using the factor for water density provided in Enclosure 2. The correction factor obtained'from the enclosure is divided into the measured value.to obtain the density corrected value. I 2. Containment building sump samples do not require correction for temperature 'and pressure within the accuracy of this procedure. .3. Containment building atmosphere gas samples are corrected using the following equation. I* Specific Activity (STP) = Specific Activity x-( )x(t + 460) P + 14.7 492 j Where: 1 l; T,P = Measured Sample temperature and' pressure recorded in step 8.- K. Correct the sample specific activity at STP for decay back to the time .S of reactor shutdown, as recorded in step C.:using the.following 4 .) equation. Enclosure 4 is provided as a worksheet. A = A 0 Where: , 37 A = the specific activity of the sample corrected back to the I o time of reactor shutdown, ci/cc. i A = the measured specific activity, yci/cc. the radioactive decay constant, 1/sec. = z the time period from reactor shutdown to sample analysis, sec. = i l ISSUE 9 U' FC/EPIP/02 MAY 151987 R4 05-15-87 f S

EPIP-TSC'-8-8-6 III. PROCEDURE-(Continued) L. ' Identification of the Fission Product Release Source. 1. Calculate the following ratios for each noble gas and iodine. isotope only using the specific activities obtained in step L. is provided as a worksheet. Noble Gas Ratio = Noble Gas Isotooe Soecific Activity Xel33 Specific Activity Iodine Ratio = Iodine Isotope Soecific Activity I-131 Specific Activity 2. Determine the source of release by comparing the results obtained to the predicted ratios provided in Enclosure 5. An accurate comparison is not anticipated. Within the accuracy of this procedure it is appropriate to select as the source that ratio which is, closest to the value obtained in step L.1. M. Calculate the total quantity of fission products available for release to the environment. Enclosure 6 is provided as a worksheet. 1. The quantity of fission products found in the reactor coolant is calculated as follows.- a. If the water level in the reactor. vessel recorded in step 8 indicates that the vessel is full, the quantity of fission ] products found in the reactor coolant is calculated by the following equation. TotalActivity(Ci)=Ao(uci/cc)xRCSVolume l Where: Ao = the specific activity of the reactor coolant sample ] corrected to time of reactor shutdown obtained in step K, uci/cc. RCS Volume = the full reactor coolant system water volume-corrected to standard temperature and pressure using. b. If the water levels in the reactor vessel'and pressurizer-recorded in step B, Table 1. indicates that a steam void is present in the reactor vessel, the quantity of fission products found in the reactor coolant is calculated'using j the equation from step M.1.a. 'l c. If the water level in the reactor vessel recorded in step B i is below the low end capability of the indicator, it is not j possible to determine the quantity of fission products from { this sample because the volume of water in the reactor coolant system is unknown. Under this condition, assessment-of core damage is obtained by using'the containment sump ~y i sample. FC/EPIP/02 R4.05-15, MAY 151987

s e-A.

a

c

w a

) EPIP-TSC-8-B-7 . <~s . II

I. PROCEDURE

(Continued) .t^ M. 2. The quantity of fission products found in the containment-building sump is determined as follows. 'a. The' water volume in the containment building sump is determined from the sump level recorded in step,B, Table 1 and the curve provided in Enclosure 7. b. -The quantity of fission products in the sump is calculated by the following equation. . Total Activity, Ci = Ao (uci/cc) x Sump, Volume Where: Ao = the specific activity of the containment sump sample corrected to the time of reactor shutdown ~obtained in' step K,uci/cc. 3. The quantity of fission products found in the containtaent building atmosphere is determined as follows. a. The volume of gas in the containment building at the time of the accident, is corrected to standard temperature and' pressure using the following equation. Ca" . Gas' Volume (STP) = Gas Volume x (14.7 + P ) x492 1 14.7 (T1 + 460) Where: 'T, P1 = Containment Atmosphere temperature and pressure 1 ~ recorded in step B. Table 1. 1 4 The total quantity of fission products available for. release to the environment is equal to the sum of the values obtained from each sample location.

j

(( ISSUED FC/EPIP/02 MAY.151987 g4 os.13 37 .l

e i

s

e u.' 1 .g 4 EPIP-TSC-8-8-8 II

I. PROCEDURE

(Continued) '} N. PLANT POWER CORRECTION \\ The quantitative release of the fission products is expressed as the percent of the source inventory et the time of the accident..The equilibrium source inventories are to be corrected for plant' power history. Use information from step F Table 3. 1. To correct the source inventory for the case in which plant power level nas remained constant for a period greater than four. radioactive half lives the following procedure is employed. is provided as a worksheet, a. The fission products.are divided into two groups based upon the radioactive half lives.. Group'1 isotopes are to be employed in the case where core power-had not changed greater than : 10 percent within the last 30 days prior to the reactor shutdown. Group 2 isotopes are to be employed in the case where core power had not changed greater than : 10 percent within the last a days prior to the reactor shutdown. 1 b. The following equation may be applied to the fission product Group which meets the criteria stated in.N.l.a only. Group 1 Power Correction Factor = n Steady State Power Le' vel for Prior 30 days-

)

i 100 Group 2 Power Correction Factor = Steady State Power Level for Prior 4 Days 100 2. To correct the source inventory for the case in which plant power-level has not remained constant prior to reactor' shutdown, the 1 following equation is employed. The entire 30 days power history should be employed. Enclosure 9 is provided as a worksheet. OT Power Correction Factor = I,P, (1-e j) e~ j 100 Where: P) = steady reactor power in period j t = duration of period j t) = time from end of period j to reactor shutdown ISSUED MAY 151987 -93_13,37 FC/EPIP/02 g. ..A w

-EPIP-TSC-8-B-9. ) l II

I. PROCEDURE

(Continued) n 0. Comparison of. Measured Data with Source I'nventory .The total quantity of fission products available for release'to the environment obtained in step M.4 is compared to the source inventory corrected for plant power history'obtained in step N.2. This comparison is made by dividing the two values for each. isotope and calculating the percent of the corrected source inventory that is now in the sampled fluid and therefore available for release to the environment. Enclosure 10 is provided as worksheet, l P. CORE DAMAGE ASSESSMENT ] Sunnarize the core damage results as directed in procedure step H. l s-l i (., 1SSUta FC/EPIP/02 MAY 151987 _R4. 05 15 87

.y EP IP-TSC-8 10 TABLE 1 Temperature Pressure 'F ps i a -- Containment Gas Loop RCS (Tave) Reactor Vessel Level Pressurizer Level Containment Sump Level Specific Activity (microcuries/cc) RC CA CS: Isotope Xe-131M Xe-133 .r. Kr-88

j -

Kr 85 Kr 87 I-132 1-133 I-135 I-131 Cs-134 Te-132 Ba-140 Ru-103 Time of Sample Measurement ISSUED s FC/EPIP/02 R4-5-15-87 =

.-x-s EPIP-TSC.3-B, ('. '~ ' TABLE 2 Safety Injection Tank Level, % Before After SI-6A SI-6B SI-6C SI-6D SIRWT Level (Inches)' Before 'After SIRWT g ' (.7 ISSUED FC/EPIP/02 MAY 151987 R4 05-15-B7 M t 1

i EPIP-TSC-8 12 TABLE 3 30 DAY POWER HISTORY O P) T T 4 Up to 8 Intervals May be Revised j! i Pj = Steady reactor power operated in period j % full power NOTE: 'In each period, the variation of steady power should be limited to t10%. Tj = Duration of operating period j (days) l Toj = Time between the end of operating period j and time I of reactor shutdown (days) ,i FC/EPIP/02 l 1SSUED ur15:M R4 5-15-87 i q I

EPIP-TSC-8-B-13 0 APPENDIX A l j t : User input guide for core damage estimate computer program PRGPASS. I l CARD 1 I. 1 ~ 20 - 25 30 35 I 40 5 10 15 l HOURS ..TIIfMtE1= H10!UI 15 l 111' tti t I( f1 tt, i t t i_1 iit i R Hours is in F5.1 format starting in Card Column 12 Cf? Hours is the time from reactor trip until the primary sample is read. CARDS 2 through 5 - SI Tank Levels j q l 5 10 15' 20 25 30 35l 40 SlIl-161X f111 101.1010 0tEf+10f0 (01.1010 01E l+ s 010 III I f f1 i Cards 2 through 5 are for SI, 6A, 6B, GC, 6D tank levels before use and af ter use. The levels are in percent. Format is E9.3,1X,' E9.3 starting in Card-Column 12. ,y I.- 1SSUED FC/EPIP/02

. i i;?

~R4 5-15-87 f

1 EPI P-TSC-8 14 CARD 6 - S!WRT Level 1 -..,3 ? i 5 10 15 20 25 30 35 40 INITIAL FINAL j l SlT lRlWIT 11II IIII ffi 1 IIII I' ' ' I ' o 4 Level is input in Card 6 is for SIWRT level before and af ter use, if any. Format is E9.3,1X, E9.3 starting in Card Column '12. inches. CARD 7 - Containment Temperature and Pressure t l 36 l '40 5 10 15 20 25 30 TEMP. PRESS. 1 ,h C l 0l Nl T1 A.IlNIMITl lll1 fit 1 lf\\\\ lt I 1 111 1 II t t .:? Card 7 is for input of containment temperature (*F) and containment pressure (PSIA). These values should correspond to the time the containment atmospheric sample was taken. Format starting in Card Column 12 is E9.3,1X, E9.3. l CARD 8 - Gas Sample Temperature and Pressure 5 10-15 20 25l 30 35 40 TEMP PRESS. GlAISI TL LI C1 0t t illI t,ig lgll P Card 8 is for input of the gas sample temperature (*F) and pressure (PSIA). The femat starting in Card Column 12 is E9.3,1X, E9.3.

z. )

FC/EPIP/02 ISSUED R4 5-15-87 [F 15 = a s. A-

EPIP-TSC-8-B 15 CARD 9 - Core Average Reactor Coolant System Temperature g( 5 10 15 20 25 30 35 40 TEMP. RICtsi 1T _ EIM1 Pt t lIII fff f IfII If f f III1 Card 9 is for input of the core average RCS temperatures ('F) cc.rresponding to the time when the primary sample was taken. Format starting in Card Column 12 i s E9,.3. CARDS 10 through 21 - Isotopic Information 5' 10l 15 20 25 30 35l 40 RC l' CA' CS ([ I ISI 0ITIO P I E l !.1 1III fI1 1 IIII ff f I f I1 i f1f I { 1 i Cards 10 through 21 are for input of isotopic information. { The activities input are in microcuries/cc. The reactor coolant values start in Card Column 12. The containment atmospheric values start in Card Column 22. The containment sump values start in Card Column 32. ] I If a sample is not read for a particular isotope or a region is not sampled, input zero's. The format for each card is A10,1X, E9.3,1X, E9.3,1X, E9.3.. The isotope cards must be in the following order. CARD # ISOTOPE 10 XE-133M 11 XE-133 12 KR-88 j 13 KR-85 14 KR 87 15 I-132-16 I-133 17 I-131 h^.; ISSUED 18 CS-134 19 TE-132 GY 15 357 20 BA-140 21 RU-103 R4 5-15 87 FC/EPIP/02 i e q

? r EPIP-TSC-8 16 CARD 22 - Choice to use Noble Gases or Cs-134/I-131 for Overheating Calcula-tions. .\\ 'c !' 40-SI 10 15' '20 25 30 N1 j C lH l0fi t C EI f f f I I 'l I ff1 I ItI1 11 1 1 1 111 t t 't t W1 = 0 Use I-131 and Cs-134 O = 1* Use Nobel Gases 1

Use the data,for Iodine or Cesium only when the data for noble gases is not available.

CdRD 23 - Number of Values in History File (Cards 24 through 31) <3 'c! do 5 10 15 20 25' in i N PlWlRl$1 {ttq I(f( ffq f (fll tt t t fit t t tt t N = Number of power changes for 30 Days prior to trip, N = 1'to 8. Integer format in Card Column 12. ] FC/EPIP/02 ISSUED ..) R4 5-15-87

L s

e

EP!P-TSC-8 17 ] 1 p CARDS 24 through 31 - Power History ( 5 101 15 20 25' 30l 35 40 I l POWER DAYAT DAYTil lfi1 1Ii1 1Iii f f I 't ItlI -l 1 1 1 111 1 I t t 't l The power history is for the 30 days prior to shutdown. Up to 8 power '] changes may be modeled. POWER =. Power level in percent of full power. Note 10% levels are sufficient modeling changes. DAYAT = Duration of operating period (Days). DAYTIL = Time between the end of the operating period and the time of reactor shutdown (Days). FORMAT = 15X, F5.1, 5X, F5.1, SX, F5.1. ISSUED ma, FC/EPIP/02 R4 5-15-87 9

)l g' ${,? M y ) c r i o tat .+ s n ise rav e n 0 0 0-tdI 5 5 .5' ce ase 0 n 0 n 0 n rsc I 1 a 1 a 1 a' aer h h h hru n n t n t n t Cpo a a 0 a 0 a 0 xS h h 5 r h 5 r h 5 r fE t t e t e t e o f o t o t o t eo s s t a 's t a s t a ep s s e s e s e sot e e 0 r e 0 r e 0 r at n L l 1 G L 1 G l 1 G e eoe g l sc a eI r m R e a P D 3 le c 3 ,2 04 u i 7 1 83 44 F t 3 3 81 11 s 1 e3 f - i e X1 be ar o rp s RT LP eo C .Im. s tt e co 49 02 1 i as 18 31 32 44 r ri 38 13 11 11 o a 1 1 g h b e se aa e C IR XI CT BL t a C t t t t t t 1 f e e e e e e C oe l l l l l l e R s l l l l l l r N ea p p p p e e e e e' e u ce a a a a P P P P P P s f rl G G G G o o ue l l l l l l ,) l z oR s s s s e e e e e e c s S a a a a u u u u u~ u n c G G G G F f F F f F ,t E i t s e i s n r n a e~ e o e t t f g .i l 5 l c o n ds y e" o a e im nu r RI M r ms k u af a a 0' m a sa ii f d l h i e pn ti n o C nl Sa sD un n r ae r r oo o 2 f l hR nU vpe Bi i0 a c e 8as s sU e c e gp Ga nu u. al p i M om d e .if fm g l a asl af f o ce o ar l ae ri ir su I l IT CGR GD DF EF o d ) i a R ~ te f l t l t o g e l e t e e e n r e l l u l yg e i u g P l e f l ra g d l n e H e om a d ei i l e P e P' ga m a t a d eg tt g g t tt eD a l aF d un aen l n e al .l t D C i a fi ili ei l ie e al dg l t dlt ut l dM u' Ce l l e en Ce l a eea fa e e F' u e ar mi r ae mPe e P mt CF u iu rd ru ih r h rh re r R F tl ed ol t r el r or l el ot N ii t a ji ie t ee je e tl jl o na nl aa nv nuv av u ne ae N IF IC MF Io IFO Mo f IP MM "cmO L ,+~i" 'n' 0 E % < S E p~I N

5 e EPIP-TSC-8-B-19 . ( ENCLOSURE 2 RATID 0F H O DENSITY TO H O DENSITY AT 2 2 STP vs TEMPERATURE 700 I s00 - 500 C;; 400 _ 300 - 200 - 100 t t t 9 0 0 0.25 0.50 0.75 L0 / S PACTA TP h FC/EPIP/02 lSSUED ...-r i 5 r47 R4 5-15-87

b EPIP-TSC-8-B-20 ) l h ENCLOSURE 3 1 i i FECORD OF SAMPLE TEMPERATURE CDRRECTION i Sample Number: Location. 1 . Time of Analysis: Temperature, 'F: Pressure, PSIG: 1 Measured Specific Activity Correction Specific Activity Isetece '( Table 1 ),UCi/cc Factor 9 STP, "CI/ce Kr 87 Xe 131m b Xe 133 .hi? I 131 1 132 j I 133 I 135 Cs 134 Rb 88 Te 129 Te 132 Sr 89 Ba 140 La 140 i La 142 Pr 144 ..)

r' ISSUED FC/EPIP/02 s -

~ ~ ~ R4 5-15-87 p n

~. EPIP-TSC-8-Bqi ENCLOSURE 4 RECORD OF DECAY CORRECTION Time of Reactor Shutdown, Step IV.C. Sample Number: Location:. Time of Analysis: l Decay Specific Activity Decay Corrected Constant. 9 STP (Enclosure 3), Specific Activity. 'uci uci jeg fee Isoteee 1/sec Kr 87 1.5 (-4) Xe 131m 6.7(-7) Xe 133 1.5 (-6) I 131 9.9(-7) I 132 8.4 (-5) I 133 9.3'(-6) I 135 2.9(-5) Cs 134 1.1 (-8) j Rb 88 6.5(-4) Te 129 1.7(-4) Te 132 2.5(-6) Sr 89 1.6(-7) Ba 140 6.3 (-7) La 140 4.8(-6) La 142 1.2(-4) (, Pr 144 6.7 (-4) !SSUED FC/EPIP/02 g35g

^% U ~

yy, a,'

w u s, de i e f c i r t u n o e S d I o 3 i t p 0 5 0 0 a a R G 0 0 5 y s 0 N t a 1 1 1 0 O i G 0 0 1 I v 0 0 0 0 0 5 1 T i n A t i 0 0 1 1 0 0 0 C c I A FI TN ED I t y e y t E l r 3 i C l o 0 v 2 0 0 0 4 0 8 e t i R e 0 0 1 1 1 2 1 y t y P n t t y c t 0 l v i t A i S e n v i v u I i v c i 5 E F t i i t S c t f c E A R E A c i A A c U L c e c O E S i c pi _ l f i S f L i f i C T o c i e c h N C d i e c p e E E U e t p e o p D t a S p t S O a R S o R l s s 1 P u e a 3 .I 3 c p G 3 1 N l o 1 e O a t e n 1 I c o l e i S s b X d d S I o o e I N d I t F e c d t d e F e c e r 0 t e t r c r c o ) c e r e C l y c r o r lo d t / r C r y c e iI o o a E t v C C y C c P, a e R c i y c yD e t r c ) a e a r A 4 c D c o e e C c e D D i r y f u = = a i s c c o o o r n e e l i i e o D p c t t b i S n a a m t E R R u a ( N c s -e o a n e L G i l d p e o m m l I a e S p 1 3 h o 7 3 3 1 2 3 5 o t 8 1 1 3 3 3 3 H 1 1 1 1 os r e e I K X X I I I I s_._ mmcmo

f. " = tN mC"o &Ar a

u gu ,O i a e i r

. o: EPIP-TSC-8-B-33 _. s..:. ENCLOSURE B RECORD OF RELEASE QUANTITY Reactor Coolant Containment Sump Containment Total Sample Number. -Sample Number. Atmosphere Sample Quantity Isotooe C1 Ci Number , C1 Ci Kr 87 Xe 131m Xe 133 I 131 h I 132 I 133 I 135 -i l Cs 134 Rb 88 Te 129 l Te 132 Sr 89 Ba 140 La 140 La 142 Pr 144 I o* ISSUED FC/EPIP/02 u,,.,,.W, R4 5-15-87 1

s 1 4 EPIP.TSC'8-B-24 ' ENCLOSURE 7- ~ "y l CONTAINNENT BUILDING WATER LIVEL vs YDLUME 2 ) I 1 8 '/ 7 ,.j ~ 8 / I ~ gs ] .i 1 ,. ;;s. m 'p j. w . us. s g 4 1 3 a / 1 2 i / 1 i i i J e ' 0 i 0 10000 .20000 20000 40000 50000 3 i VOLUME; FT 4 6 ISSUED FC/EP:P/02 '~ -R4,5-15-87; ~ - a

-q .lrrl 3 3 1 .i EPIP-TSC-8-B - 1 ENCLOSUd"E 7A ,3 i (- 1 .CMWNMENT SUMP CURVE j i l LEVEL VS VOL.tME a 3 I GALS. FT l 110 14.71 1 y ~ 1903 I 102 '/ f in - 13. 3 / i 5 - 12.03 /- N - 10.70 e E 70 - B.2 l C'.. e nn . e.0a ~ W ~/ 5 2 - 8.2 E!! ~ I. / e - s.m / 2 - 4.01 -f E :W / 10 - 1.2 ~ '""""}"""'" ~, n n o,

nnnn, n o n n, en'onn nenn" g

4 2 4 6 6. 10 12 14 16 18 - 3 22 24 /' ' 50% 100% L: ii 'am m'2s 1SSUED . 33, SUMP LEVEL - INCHES R4 5-15-87, FC/EPIP/02 a.

-4 y, A EPIP-TSC 8-B 26

ENC 1.0SURE 8 J s RECORD OF STEADY STATE POWER CCRRECTION

.] 'I Sample Number: Location: Steady State ~ 30 Days Power. Level: ' Steady State.4 Day Power Level: Fuel Power 1 ' Equilibrium Corrected History Correction x ' Source Source = Isotope Grouping Factor. , Inventory -Inventory Gas Gap Inventory Kr 87 2 3.6(0)- Xe 131m 1 2.7(4)' t Xe 133 1 -1.3(7)L a3 i I~131 1 .4.4(6) 8-I 132 2 4.9(3) I'133 2 4.4(6). i I 135 2 7.0(5) Fuel Pellet Inventory Kr 87 2 1,8(7) Xe 131m 1 2.9(5) Xe 133 1 .1.5(8)' I 131 1 4.8(7) I 132 2 7.0(7)- I 133 2 1.5(8)- I 135 2 8.6(7) 1 Cs 134 1 6.1(6) Rb 88 2 2.9(7)- i Te 129 2 1.6(7) Te 132 1 7.0(7)' Sr 89 - 1 3.9(7) Ba 140 1 8.0(7). La 140 1 8.4(7)- La 142 2 1.0(8)- Pr 144 2 6.5(7) 4 i SSUED FC/EPIP/02s

' M "i

R4 5-15-87 _ =,.

~ - - ~ .a.. ) iPIP-TSC-8-B-27 ' 1 ENCLOSURE 9: 4 .b l W.CDE) 0F TRANSIENT POWER ConRECTION 'i .3 Sample Number: Location: Prior 30 Day Power History: . Power %' Du, ration. Days l u 1 .l ~ 1 Power Correction x Equilibrium Source-. Corrected-Source! Isotope Factor inventory Inventory Gas Gap inventory. 1 -i Kr 87 3.6(0) Xe 131m 2.7(4)- Xe 133 1.3(7) ( f. I 131 4.4(6) (f I 132 .4.9(3) I 133 4.4(6) 1 135 7.0(5. ) Fuel Pellet Inventory Kr 87 1.8(7)' I Xe 131m 2.9(5) i Xe 133

1.5(8)'

I 131 4.8(7)- I 132 7.0 (7 ).: 1 I 133 1.5(8)- 1 135 '8.6(7)- Cs 134 6.1(6) Rb 88 2.9(7). Te 129 1.6(7)' Te 132 7.0(7) Sr 89 3.9(7) Ba 140 8.0(7) La 140 8.4(7)' l La 142 l.0(B) Pr 144

6.5(7)

I i S S U E D 'r' \\ t/ EPIP/02 R4.5-15-87 'L L..

- 4 o EPIP-TSC-8-B-28 m ENCLOSURE 10 HECORD OF PERCENT HELEASE Total Quantity Power Corrected Available For Release + Source Inventory, x 100 = Isotece (Enclosure 6), ci C1 ( Enclosure 8 or 9 ) Percent Gas Gao Inventerv Kr 87 Xe 131 Xe 133 I 131' I 132 1 133 I 135 Fuel Pellet Inventerv Kr 87 c.~, Xe 131m ,S ) Xe 133 I 131 -l I 132 1 133 I 135 I ~ Cs 134 i Rb 88 Te 129 Te 132 j ' Sr 89 Ba 140 .La 140 La 142 Pr 144 j i I l FC/EPIP/02 ? Je) c ISSUED sf15 Gb i R4 5-15-87 i j +. ..s a -

~T ~ w. EPIP-TSC-8-C-1 'f SECTION C f ' Estimate of Core Damage Usino Hydrocen In Containment I. PRERE0VISITES l. i-A. A plant accident with the potential for core damage has occurred. B. The Post Accident Sampling System (PASS) is operable and'h's the.: a capability to obtain and analyze the identity and. concentration of fission product isotopes'in fluid samples which have the potential to be highly radioactive. II. PRECAUTIONS AND' LIMITATIONS The assessments of core damage obtained by using this procedure are only estimates and represent lower limit estimates of clad damage. This procedure only estimates the' percentage of rods which have. progressed to at least clad rupture or clad embrittlement, and.does not attempt to predict the physical configuration of those rods which have progressed beyond local clad fragmentation. This procedure assumes there are no voids measurable by the Reactor . Vessel Level Monitoring System. - However, if-the hydrogen samples are (1. taken under conditions.in which a measurable void does exist, the attached addendum contains guidelines to estimate the added contribution of hydrogen, by-that source, to the tota 1' hydrogen measured. This' procedure proyides an estimate of only the percentage'of. rods which have progressed. to at,least clad rupture or clad embrittlement, l and does not attempt to predict'the physical configuration of those rods which have progressed beyond local clad fragmentation. j i a s i ISSUEO i . MAY 151987-FC/EPIP/02 . R4 05-15-87

+ EPIP-TSC-8-C-2. - II

I. PROCEDURE

') Estimation of Core Damage Using Hydrogen Measurement a. Record the estimates for core uncovery and recovery times using-instrument records from the Reactor Vessel Level Monitoring System. Core Exit Thermocouple Temperature and Core Exit Thermocouple Saturation Margin. Estimated Estimated. Instrument Core Uncovery Time- . Core Recovery Time Reactor Vessel Level Lower Limit Elevation Lower Limit Elevation Monitoring System Uncovers.- Recovers. Time Time Core Exit Thermo-Start of Continuous Rapid 'Temperatu re ; couple Temperature Rise or Exceed of 600F. Orop to Saturation. Time Time Temperature - Temperature Core Exit Thermo-Start of Superheat. Return to Saturation couple Saturation Time - or Subcooling. Margin Time ~ b. Interpret the data from step a to determine the best estimate for.the time period of core uncovery and obtain the range of RCS pressure s.s (pressurizer pressure) indicated for this-time period. The.superheat .h derived from the thermocouple temperature and corresponding system - pressure is considered as the best indicator for coie.uncovery during core boiloff and should be used but should be. compared with other indicators to help identify possible anomolies.- R reord.these values below. 1 Core Uncovery Core Recovery Time Pressure c. Observe available instrument records to determine if there was some. reactor vessel inlet flow during the rising temperature portion of the core uncovery period up to approximately the time of peak core exit thermocouple temperature. Charging Flow Rate GPM . Letdown Flow Rate GPM HPSI Flow Rate' GPM

LPSI Flow Rate-GPM Other Flow Rate GPM NOTE: Net inlet flow indicates that this. procedure may have additional bias which underpredicts clad damage, ISSUE 9

. _.) FC/EPIP/02 MAY 151987 R4 05-15-87 4 4 e A

EPIP-TSC-8-C-3 II

I. PROCEDURE

(Continued) d. Obtain a liquid sample from the RCS' hot leg and a gas sample from the-j containment atmosphere and record the conditions in the containment at' q the time these samples are obtained, below. Analyze the samples for hydrogen concentration using the PASS procedure and record these-. results below. Estimate the total cubic feet of hydrogen'at standard temperature and pressure as outlined. Reactor Coolant' System Containment l Sampling Time LAtmosphere Pressure psig Pressure 'psig Atmosphere Temperature F. Temperature F Reactor Vessel ~ Coolant Level .Does Pressure or. Temper-es/No Y ature History Indicate a Pressurizer Level -Hydrogen Burn Cont.' Sample X Cont. Vol. X (32 + 460) + Normal Temp. = Ft3 Hydrogen l (Vol. %/100) (Ft3) + 460 at STP X 1.05 x 10 6 X 492 =- Hot Leg Sample X RCS Vol. X Density Ratio + 1000 = Ft3 Hydrogen (ce/ko la STP) (Ft3) (Enclosure 2) at STP X 6395 X + 1000 = l f Total = Record total hydrogen measured in step g, also. ~l j 1 i i h. ISSUED v FC/EPIP/02 MAY 151987 R4-.0s-15-87.

J e 's j EPIP-TSC-8-C - II

I. PROCEDURE

(Continued) e. Record the containment temperature.at selected time intervals and calculate the hydrogen generated by oxidation of containment materials ] utilizing the typical plant hydrogen production rate from Enclosure 3. [- -j 1 2 3 4' S l Avg. Containment H 2(Prod. Rate Time at Start Interval Temp.-During ft8/hr) Produced = of Intervals ' Duration (hr) Interval ('F) (Enclosure 3) H2X4 Accident Starts i

..)

Sampling Time Long Term Hydrogen Production in Containment, Total:= ft3 OSTP Typical Short term rapid hydrogen production by . containment aluminum 1000 -SCF'- Total Hydrogen Production In Containment SCF. ] Record total in step g also. 'l S S U E D ,;[ MAY 151987 g4. og.35 37 FC/EPIP/02

s. a n. EPIP-TSC'-8-C-5 II

I. PROCEDURE

(Continued) f. Record the data requested below and utilize the curve of Enclosure 4 to obtain the cubic feet of hydrogen at STP generated by radiolysis. Determine the approximate power to be used as.follows: For the case in which the operating. power is constant or has not changed by more than 210 percent for a period greater than 30 . days, that power is used. For the case in whic,h.the power has not remained constant during the 30 days. prior to the reactor shutdown, engineering judgement is required - to determine the most representative power level. The.last power levels at which the reactor operated should weigh more heavily.in the' judgement -l than.the earlier leveTs. Continued operation for an extended period should weigh more heavily in the judgement than brief transient levels. i I For the case in which the reactor has produced power for less than 30 days, the. procedure may still be employed. However, the estimate'of hydrogen from radiolysis will be too high and therefore the calculated-hydrogen by core oxidation will be too low. Hence an underprediction of core damage may result. 1 i Prior 30 day power history Power, cercent Duration, Days i Power to use in evaluating long term hydrogen production by radiolysis = ('1500', Mwt) X Reactor Trip Time hrs Sampling Time (see step d) hrs Decay Time (Sampling Time - Trip Time)- hrs Enter abscissa on Enclosure 4 with above decay time and read two values for hydrogen produced by radiolysis, one from each curve, in. cubic feet of hydrogen at STP per Mwt operating power. Multiply.by above power and record as follows: Hydrogen Produced. Operating Total Hydrogen Limit Curve (SCF/Mwt, Enclosure 4) Power Produced (SCF) Upper Lower Using results from Radiological Damage Assessment Procedure, estimate I which results should be used; upper limit for major fuel overheat, .m lower limit for initial fuel overheat or appropriate estimate between the two curves for intermediate fuel overhea't. Circle corresponding ~ value of hydrogen above and also record in step g. FC/EPIP/02 R4 05-15-87 MAY 151987

~' ) EPIP-TSC-8-C-6 II

I. PROCEDURE

(Continued) 1 g. Enter the amounts of hydrogen from steps d, e, and f below. Hydrogen Measured, Step d. SCF I J Hydrogen Produced in Containment, Step e. SCF Hydrogen Produced by Radiolysis, Step f. SCF i Subtract Step e and f from d to get j Hydfogen Produced by Core Clad Oxidation SCF j 1 ~ Divide b'y (2570'SCF/1% Clad 0xidized) =- l = % Core Oxidated 1 h. Enter the abscissa of the curve on Enclosure 5 with the "% Core. Oxidized" from step g. Use the curve labeled with the pressure closes to but greater than the RCS pressure during the core uncovery period as obtained rrd recorded in Step b. Read on the ordinate of Enclosure 1 5, the percent of fuel. rods with ruptured. clad and record below. Note that.the sensitivity of measurement of ' hydrogen is comparable to the s range of oxidation on Enclosure 5. Hence, small amounts of clad l rupture are not reliably predicted by this procedure, j % of Fuel Rods with Ruptured Clad =- i. Enter the abscissa of the curve on Enclosure 6 with th "% Core 0xidized" from step g. Read on the ordinate the lower and upper s' ) values of the range indicated by the curve for the percent of fuel rods which have embrittled clad and record below. Percent df Fu'el Rods with Local Oxidation Embrittlement Range - Upper Lower j. For a given percent oxidation of the core clad, the_' lower limit 1 estimate of embrittled clad in step i is, for most accident scenarios, the least amount of potential fuel structural failure.- Step g may be 1 interpreted as follows: When the pressure during uncovery from step b is less than about 100 psia, a rapid core uncovery by blowdown is concluded. Heatup-with minimum clad oxidation occurs. The extent of potential i clad structural failures by melting may be greater than the upper ' 1 limit of embrittlement from step i as determined by oxidation. Hence, use the upper limit from step 1. When there is inlet flow while the core is uncovering, the rate: I of uncovery is slower than assumed in the derivation of the curves on Enclosure 5 and 6. For a measured total amount of oxidation, the incal percentage oxidation is probably greater along a shorter length of the upper portion of the fuel. Hence, l favor the upper limit from step i. 1SSUED FC/EPIP/02 R4 05-15-87 l MAY 151987 9 4 t

~ EPIP-TSC-8-C. j r, . II

I. PROCEDURE

(Continued) ) N k. CORE DAMAGE ASSESSMENT I i The conclusion on core damage is made using the two 'results from ' i above. These are: 1. Percentage of fuel rods with ruptured clad, step h. 2. Percentage of fuel rods with embrittled or structurally damage clad, step i. 3 i . Knowledgeable judgement is used to. compare the above two results to { the definitions of the 10 NRC. categories of fuel damage found in 1. Core damage does not take place uniformly.. Therefore j when' evaluating damage using these results, Enclosure 1 may' yield a: { combination of categories of damage which exist simultaneously. ] Ol I l .f) , ' *j: i h ^ ISSUED FC/EPIP/02 - MAY 151987 R4 05 15_g7 0 4

v -- A 4 ' ~ e s a O' 9I ,n'* u x x 0 0 5 5 s 1 0 0 - fd 1 n 1 n oo n 0 a a R a n 5 h n h t h a T a T h ne T h o T r eg T t r ca s e e rm s s 0 t s t 0 ea e s 1 a s a 1 -PD L e e e e L r L r p G G on e r r u .t o B n n Cn e F c o d me F F i, ti n .eg 0 .t nt 1 1 a rn 0 0 0 a ea18 5 e ua 5 0 31d l d3! 6 a g sR 5 7 22i ai<< 1 is {<0 iO p x vx a a 1 1 m .e' .a M u 0 D q 0 E 2 l l e u c o t F i e tt e e s r f sn r l e a t n o ie o pr G no us rm C uu f ei s ee ot ondl t) s e t r m ca eeaae er i cu u or t gcvdr r as m me n o uii o P o ra it rp urduxC r g ae xi em odoqO e hM ax he myrE-f o f t C ME TT AlfP(1o a n x 1 .C e n l e C o a v r i r R i ro G,y u N t ut' s o a t o f me t z ce uu .o l o sg i ia e r rD' ~ c l n s nm u u t c aa D s S yd n a E V hD s t ao i t c ep e fili s ef ra r orCt Mo uG p g a ir e t 'r seld p e n ps e st ei u t o ua v onux n c N RG O LIF0 ru a B ra e h r) 0 0 l C uF 0 5 eu t* 8 3 f e a( 0 1 3 g r 5 a ee 7 0 0 d m pg 0 0 n a mn 2 8 a D ea 1 1 e TR r d ) 'u a o s m l t s C e e l t r l e P e g e yg n r e l r .ra e i u g P l om g d l n e o ga a d-ei i l e-P t eD m a t a d eg tt g g c m t a l aF d un aen l n_ a i a Fi ili ei e al D C Ce dg l t dlt ut R u l l e en Ce l a eea fa CF e ar mi r ae nPe e n R u i u rd ru ih r h rh o i Nf F tl ed ol t r el r or o ii t a ji i e t ee je s o na nl aa nv nuv av d c N IF IC MF IO IFO MO n. e pe D = n -(n (n C m O = ,.RO-Q" ~ Au~*g~ m4 '" g u

EPIP-TSC-8-C-9 ENCLOSURE 2 1 ,r RATIO OF H O TO H O DENSITY AT 2 2 STP vs TEMPERATURE ) i l 700 600 - 1 500 - t 400. w' 300 - 200 - 100 1 I t t 0 0 0.25 0.50 0.75 1.0 /

ACT 8STP

(- (SSUED . '. 5 ' r: ' R4 5-15-87 -i 1

O e EPI P-TSC-8-C-110 ENCLOSURE 3 i TYPICAL HYDROGEN PRODUCTION RATE 1100.- FROM ALUMINUM AND ZINC vs. TE'9ERATURE 1000 900 q a 800 2-E \\ 'g 700 ) i 9 0, . c 600 G i 7 3 8 s00 - ec>z m C, 400 tz 300 = 200 - 100 - ] e e t t ,~ )3.gog g o 150 200 250 300 g.3.. TEMPERAT1Jr.E,

F R4 5-15-87

+ e

llll l.

~ e 1 4 0STM eVT- ~ &gh9 s@9. $9 zf a8a2 ne0o3E$ N* I,E$ c ns 1 1 1 1 0 1 2 3 4 5 6 7 e e 0 1 2 3 O 10 0 SP E C I 2 F 0 I 0 C R M A A D J I O O L R Y 3 F T 03' U I E C t L N H L O Y L V D U T E R S s I R O U G H M H 4 E E ,E 0 E N 0 A H T P 4 O R U O R D I S N U T C T E I R O 6 M N 0 0 ED vs IA IN T T I I E T M F E IA U 6 L E 0 F L 0 U O E V L E O R V H E E R A 7 H T 0 U 0 AT 8 0 0 ~ mm6mo MEE2E a u " a~ t : n'7 3 7 j, Q !!f1 {ijj

7 EPIP-TSC-8-C-12 ENCLOSURE 5 PERCENT OF FUEL ROOS WITH RUPTURED CLAD vs CDRE CLAD OXIDATION 100 7 RUPTURE TEMPER AT1.lRE 80. c 5u -,1500*F-C ) g 60 3 soc.p t E = t:: l 40,. 8m u. WHEN THE PRESSURE USE CURVE LABELED. O IN STEP & IS WITH TEMPERATURE

I" ##

I E 20 - 41200 PSIA .1500*P { < 1650 PSIA 1800*F j gJ t t t i 0 0.5 1.0 1.5 2.0 1 % OX1DAT10N OF CORE CLAD VOLUME FC/EPIP/02 ISSUED ' ' ~ R4 5-15-87 _a = 4 + 3

e g

3

EPIP-TSC-8-C-13 ENCLOSURE 6 5 0F THE FUEL RODS WITH OXIDATION EMBRITTLEMENT vs ~. TOTAL CORE OXIDATION FOR 1% TO 3% DECAY HEAT AND 300 PSIA TO 2500 PSIA WHEN COOLANT LEVEL DROPS SY BOILOFF WITH NO INLET FLOW UNTIL CORE IS RAPIDLY QUENCHED { l 1 i 1 100 ) w g q ee 80 E i C i20 e 5 i i 8 i i e i 3 E 8 d 20 E s 8x t I 0 20 40 60 80 100 % OXIDATION OF CORE CLAD VOLUME - FC/EPIP/02 ISSUED w MAY I 5 bo# .R4 5-15-87

y. '] m' r EPIP-TSC-8-C-14 Fort Calhoun Station Unit No.1 m ADDENDUM TO EPIP-TSC-8-C. Estimation of hydrogen Volume in Reactor Vessel Head Void I. PURPOSE The purpose of this addendum is to provide a guideline for estimating the 'l amount of hydrogen gas contained in a void at the top.of the reactor ' 1 vessel. This volume of hydrogen is added to the measured hydrogen in 1 procedure Step d. of EPIP-TSC-8-C to yield the total hydrogen genera.tedl.by, 1 all sources-H LII.. PRECAUTIONS AND' LIMITATIONS This addendum should only be used to supplement the procedure when the primary system could not be filled prior to taking the liquid -sample. This - addendum can then be used to estimate the hydrogen in the vessel ;v'oid which is not evident from the hot leg liquid sample. This addendum only applies when the. coolant-level is above the hot leg.and- - the remainder of the primary system is filled.. Verification that the steam. y generator tubes are filled can be provided by the existence of natural convection flow in the primary system. The reactor vessel level monitoring system is required to provide the reactor coolant level. The volume of the voidLis obtained by relating the-volume in the vessel above the reactor coolant level to the value of level- /- for the Fort Calhoun reactor vessel' design. This addendum only provides.the analytical means to estimate the hydrogen contained in the void..The presence.of other gases including helium,- nitrogen and fission product gases will add. uncertainty to the result.- II

I. PROCEDURE

a. Determin' e the conditions of the void as follows:- V = Void volume (ft3),' derived from m'easurement of coolant level-Tg = Temperature of liquid at coolant surface _(*F): .} PSAT.= Water saturation pressure at temperature T. g PTOT = Reactor coolant system. pressure-(psic) b. .A first approximation.is made~ using the folloWing assumptions: ] The partial pressure of vapor in'the void is equal to the saturation-pressure at the liquid temperature T.. ' This implies no heating of--the - void gas by the reactor vessel walls or head.. They are normally at. 1 reactor' outlet temperature and could: remain aboy'e the temperature of- .j the void causing-the vapor to be superheated. ] 1 l fSSUEO-FC/EPIP/02 MAY 151987 R4; 05-15-87 i q 1 ,,s. o s e + j.4'e t. r - + +u a. -u o n

,r s EPIP-TS'C-8-C-15 r~ ( ADDENDUM (Continued) All the non-condensible gas in the void is assumed to be hydrogen there is no helium or fission product gas from ruptured fuel rods no nitrogen from the Safety Injection Tanks. and which eliminates this. assumption is given in Step d.A second approximation Calculate the amount of hydrogen as follows: c. P =P H TOT - SAT 2 PH, 492 FT3 H @ STP = (Y) 2 14.7 Tt + 460 Add this amount to the total hydrogen in Step d. of EPIP-TSC 8 C d. A second approximation can be made using the Post Accident S System and measuring the total gas and hydrogen dissolved in the h leg liquid sample. This approximation assumes the following: o the partial pressure of a gas to the amount of gas d a es liquid sample at equilibrium. na The dissolved gas is not in equilibrium with the gas in the void dissolved concentrations are still in the same relative proportion a , the if equilibrium did exist. s The partial pressure of hydrogen is calculated from: e. (cc/kg) PH, = (PTOT - PSAT I (cc/kg) Total { equation above in Step c. Calculate the amount of hydrogen in the ve g the 1 I i l i \\ {( issuse MAY 151987 FC/EPIP/02 R4 05-15-87 _ _ _ _ _ _ - - - - - - ~ ~ ~

.~ -~ T L. T ~ ~ i + ,+ EPIP-TSC-8-D-1 -O SECTION D k.. Estimate of Core Damage Usino Core Exit Thermocouple Temperatures .I. PREREQUISITES A. A plant accident with the potential.for core damage has occurred. B.. An operational core cooling instrumentation system which. includes core exit thermocouple ' temperatures, pressurizer pressure and core level: indications. It should be able to select-and permanently. record-the highest thermocoupleLtemperature for convenient later inspection. II. PRECAUTIONS AND LIMITATIONS .The assessmentsLof core damage obtained by using this procedure are only estimates and represent. lower limit estimates.of clad damage. This procedure provides'an estimate of damage up to the time the core temperature reaches about 2300'F...At that point the rods are expected to have. ruptured' clad but little other structural degradation.' More severe core damage cannot be quantified ~by this procedure. Although this procedure yields a morelimmediate assessment.of' core: damage, accuracy is limited to relatively less severe accidents such-as slow core uncovery by boiloff of the reactor coolant. For other more rapid uncovery scenarios this procedure could yield a very Jow h.. estimate-for the number of ruptured rods. In general, for core uncovery at pressures. below about 1200 psia' there is high confidence that at least the predicted estimate of rods are actually ruptured. II

I. PROCEDURE

Estimation of Core Damage Using CETzMeasurements-a. From the recordings of maximum core exit thermocouple temperature and reactory coolant system pressure.as a function.of time, obtain and. record the following data. Maximum Core Exit Thermocouple TemperatureJ F Time of Maximum Temperature Pressurizer Pressure at Above Time psia 1SSUED FC/EPIP/02 'R4 05-15-87.- .s. ...._...s

j ., - + i EPIP-TdC-8-0-2 II

I. PROCEDURE

(Continued) r] b. Select the curve on Enclosure 2 which is labeled with a pressure approximately equal to or greater than.the pressure recorded above. Enter the abscissa at the maximum temperature from above and read on. the ordinate the percent of the fuel rods which have ruptured clad and record below. Percant of ruptured rods. c. The result in step b is probably.a lower limit estimate of damage. Some judgement on the bias is available as follows. A smooth core exit thermocouple recording and an uncovery.- duration of 20 minutes or longer will yield a good prediction of clad ruptures. A large break is indicated when the pressure in-. step,a drops-below 100 psia in less than.two minutes after' accident initiation. This causes undetected core heatup followed by flashing during refill. Depending on the rate of refill...the thermocouple temperature may rise rapidly.then quench when the. core is recovered. Under these conditions, this procedure could-yield a very low estimate for the percent of rods-ruptured. If the pressure in step a is above about 1650 psia, it.could exceed the rod internal gas pressure, depending on rod burnup, and cause clad collapse onto the fuel pellet instead of. outward -T clad ballooning. The clad rupture criteria are less well defined t> for those conditions, but at temperatures above 1800*F where the-highest pressure curve applies in step b, clad failure sufficient to release fission gas is likely and this procedure may be used to obtain estimates of damage. d. CORE OAMAGE ASSESSMENT Use the percent of rods ruptured from step c and the clad damage characteristics of Enclosure 1 to determine the NRC category of cladding failure. This procedure yields damage estimates in categories 2, 3, or 4. NRC category of cladding failure from Enclosure 1 ISSUED FC/EPIP/02 MAYl51987 R4 05-15-87 e. M_,

m y o &6L q m.% i G-0 0 5 5 .s 1 0 0 fd 1 n 1 n oo n 0 a a R a n 5 h n h t h a T a T ne T h o h eg T t r T r ca s e e rm s s 0 t s t 0 ea e s 1 a s a PD L e e e e L r L r p G G unr e u r B t o n n Cn d e F o o n me F F i ti eg 0 t nt 1 1 a i rn 0 0 0 a ea%8 56 a e ua 5 0 31d id31 ' 7 i g sR 5 7 22i ai<< s a a l I 77x vx p m e 0 iO a M u 0 D q 0 E 2 l l e u c o t F i e tt e e s r f sn r l e a' t n~ u o ie o pr G no s rm C uu f ei s s ee ot ondlt) e e tr m ca eeaae r i cu u or t gcvdr P r as m me nouiio o ra it rp urduxC r g ae xi em odoq0 e hM ax he myrE f o f t C ME TT AHP(%o a n 1 C e n l v e C o a i ro r R i G u N t ut s o a t s o f me t z ce e uu l o sg i c ia e r rD u l n s nm u u t a c aa D s S yd n E V hD s t ao i t c ep e fili s ef ra r orCt g a Mo uG i p' g ir e t r seld u e n ps e st ei n t o ua v onux r c N RG O LIF0 u a B ra e l h r) 0 0 e C uF 0 5 t* 8 3 u e a( 0 1 3 F g r 5 a ee 7 0 0 d m pg 0 0 n a mn 2 8 a D ea 1 1 e TR d ) ru a s s l t C e e r t l e g e l e P yg n r e l r. ra e i u g P' l om g d l n e o. t ga a d-ei i l e P eD m a t a d eg tt g g c t a l aF d un aen l n a i a Fi ili ei e al D C t dlt ut R Ce dg l u l l e en Ce l a eea Fa Cf e ar mi r ae mPe e n R u i u rd ru ih r h rh o Nf F tl ed ol t r el r or o ii t a ji ie t ee je s o na nl aa nv nuv av d N IF IC MF IO IFO HO nep e. D ~ m m C m o~ mCm21o" E[

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,* MAT *n S

as + EP!P-TSC-8-0-4 ENCLOSURE 2 PERCENT OF FUEL RODS WITH RUPTURED CLAD vs MAXIMUM CORE EXIT THERMOCOUPLE TEMPERATURE i WHEN THE PAESSURE USE CURVE LABELED IN STEP a IS WITH TEMPERATURE < 100 PSIA 1200* F < 1200 PSIA 1500* F 41850 PSIA 1800* F. 100 c 1200*F e y CLAD RUPTURE .j) 3 TEMPERATURE B = so ER c 1500*F m 4 g0 E g 1800* F l 5 0 20 - 5 a. O' 1200 1400 1600 1800 2000 2200 MAXIMUM CORE EXIT THERMOCCUPLE TEMPERATURE 1 3UEO o FC/EPIP/02 .84 s-is I MY 151987

A.L %n ~ O EkIP-RR-79-1 ( Fort Calhoun Station Unit No.'1 \\- EMERGENCY PLAN IMPLEMENTING PROCEDURE EPIP-RR-79 Emergency Recoverv Organization's Comouter Soecialist I.- PURPOSE The purpose of this procedure is to detail the duties and responsibilities of personnel in the Emergency Recovery Organization filling the position of Computer Specialist. II. PREREOUISITE Both the primary and alternate individuals filling.the position of Computer Specialist have been fully trained and are aware of their duties and responsibilities. III. PRECAUTIONS None I

V. PROCEDURE

Upon activation.of the Emergency Recovery Organization, those individuals assigned to the position of Computer Specialist shall carry out the. assignment detailed in Appendix 1 of this-implementing procedure, i l I 5 S U E ;; 0M FC/EPIP/02 Ro 05-20-87. k

[ g e EPIP-RR-79-2 Fort Calhoun Station Unit No.1 EMERGENCY PLAN IMPLEMENTIN

G. PROCEDURE

-i ^ 'EPIP-RR-79 APPENDIX'l COMPUTER-SPECIALIST 'A. Personnel filling this cosition By Recovery Manager'0esignation. B. Reoorting Location Electric Building C. Reports To Administrative Logistics Manager D. Suoervises/ Coordinates Provide assistance for EAGLE VM System / Software problems. '.) - .a-E. Primary Resconsibility Provide Oose Assessment Team with troubleshooting and repair coordination for EAGLE VM System / Software problems., F. Basic Duties 1. Upon notification of Emergency Recovery Organization activation, .i the-primary and/or alternate Computer Specialist will' report to their assigned location listed in Section B.of this Appendix-and inform the~ Administrative Logistics Manager of his/her presence. I 2. Remains available for troubleshooting and repair. of EAGLE VM- ~ System / Software problems. 4 -3. Keeps Dose Assessment Team informed of status-of repair efforts,. and estimated down time. ^ !SSUED j FC/EPIP/02 MAY 2 01987 Ro 05-20-87' 1 s .,q + ..}}

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