Text

Public Version of Revised Emergency Plan Implementing Procedures,Including Procedures EPIP 1.1 Re Initial Classification,Epip 1.2 Re Plant Status & EPIP 1.3 Re Estimation of Source Term
	ML20077H413
Person / Time
Site:	Point Beach
Issue date:	07/01/1983
From:	Knorr J; WISCONSIN ELECTRIC POWER CO.
To:	;
Shared Package
ML20077H411	List: ML20077H412; ML20077H413;
References
	PROC-830701-01, NUDOCS 8308100147
	Download: ML20077H413 (124)
	v • d • e

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,[' July 1, 1983 s

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POINT BEACH NUCLEAR PLANT EMERGENCY PLAN LMPLEMENTING PROCEDURES The following revised procedures are attached and should be inserted into your manual. Please destroy all revisions removed from your manual.

1. EPIP 1.1, " Initial Classification," Revision 5, dated 07-01-83.

2. EPIP 1.2, " Plant Status," Revision 1, dated 07-01-83.

3. EPIP 1.3, " Estimation of Source Term," Revision 6, dated 07-01-83.

4. EPIP 1.7, " Evaluation of Core Damage," Revision 2, dated 07-01-83.

5. EPIP 1.8, " Emergency Off-Site Dose Estimations," Revision 3, dated 07-01-83.

6. EPIP 6.5, " Technical Support Center & Operations Support Center Acti-vation," Revision 2, dated 07-01-83.

7. EPIP 7.3.2, " Post-Accident Sampling & Analysis of Potentially High Level Reactor Coolant," Revision 6, dated 07-01-83.

8. EPIP 7.3.3, " Post-Accident Sampling of Containment Atmosphere,"

Revision 5, dated 07-01-83.

. 9. EPIP 16.1, " Nuclear Engineering Section, Notification & Response,"

Revision 4, dated 07-01-83.

l The following revised forms are attached for insertion into your manual.

l Please attach the form to the indicated procedure and destroy all revisions removed from your manual.

1. EPIP-05 (07-83) , "Worksheet for Status Report on Radiation Monitoring System for Unit" (attach to EPIP 1.2) .

2. EPIP-06 (07-83), "Worksheet for Status Report on Radiation Monitoring System for Plant" (attach to EPIP 1.2.) .

3. EPIP-24a (05-83) , Site Boundary Control Center Emergency Plan Inventory Checklist" (attach to EPIP 7.4.1) .

4. EPIP-25c (05-83), " Quarterly Emergency Plan Checklist" (attach to EPIP 7.4. 2) .

5. EPIP-25d (05-83) , " Semi-Annual & Annual Emergency Plan Checklist" (attach to EPIP 7.4. 2) .

6. EPIP-34 (07-8 3) , " Calculation of Xe-133 Equivalent Release Rates" (attach to EPIP 1.8) .

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8308100147 830729 PDR ADOCK 05000266 F PDR w

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Emergency Plan Implementing Procedures - 2 July 1, 1983 Also attached is a revised listing of the Table of Contents to be inserted into your manual. Please destroy the old revision.

Also attached is a revised Table of EEIP Forms. Please insert this '

table into your manual. Please destroy the old revision.

Please fill out the attached.qceipt form and return it to Ms. F. A. Flentje at the Point Beach Nuclear Plant.

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J. E. Knorr cmz Attachments s

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POINT BEACH NUCLEAR PLANT EMZRGENCY PLAN IMPLEMENTING PROCEDURES JULY 1, 1983 I hereby acknowledge receipt of EPIP 1.1, EPIP 1.2, EPIP 1.3, EPIP 1.7, EPIP 1.8, EPIP 6.5, EPIP 7.3.2, EPIP 7.3.3 and EPIP 16.1, and have insertad them into my manual.

Date

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o 07-01-83 TABLE OF CONTENTS Revision Date 1.0 CLASSIFICATION & ASSESSMENT 1.1 Initial Classification . .. . . . .. . . . 5 07-01-83 1.2 Plant Status . . . . . . . . . . ... . . . 1 07-01-83

. 1.3 Estimation of Source Term . . . . . . . . . . 6 07-01-83 1.4 Radiological Dose Evaluation . . . . . . . . 7 05-27-83 1.5 Protective Action Evaluation . . . . . . . . 4 09-10-82 1.6 Radioiodine Blocking & Thyroid Exposure Accounting . . . . . . .. . . . .. . .. 1 02-26-82 1.7 Evaluation of Core Damage . . . . . . . . . . 2 07-01-83 1.8 Emergency Off-Site Dose Estimations . . . . . 3 07-01-83 2.0 UNUSUAL EVENT IMPLEMENTING PROCEDURES 2.1 Unusual Event - Immediate Actions . . . . . . 1 09-10-82 2.2 Unusual Event - Plant and Company Personnel Notification . . . . . . . . . . . . . . . 1 07-01-81 2.3 Unusual Event - Off-Site Agency Notification 3 05-27-83 t

3.0 ALERT IMPLEMENTING PROCEDURES 3.1 Alert - Immediate Actions . . . . . . . . . . 2 09-10-82 3.2 Alert - Plant & Company Personnel Notification . . . . . . . . . . . . . .. 1 07-01-81 3.3 Alert - Off-Site Agency Notification . . . . 2 05-27-83 4.0 SITE EMERGENCY - IMPLEMENTING PROCEDURES

, 4.1 Site Emergency - Immediate Actions . . . . . 2 09-10-82 l 4.2 Site Emergency - Plant & Company Personnel Notification . . . . . . . . . . . . . . . 1 07-01-81 l

4.3 Site Emergency - Off-Site Agency Notification 1 09-10-82 5.0 GENERAL EMERGENCY - IMPLEMENTING PROCEDURES I

5.1 General Emergency - Immediate Actions . . . . 2 09-10-82 5.2 General Emergency - Plant & Company Personnel Notification . . . . . . . . . . 1 07-01-81 5.3 General Emergency - Off-Site Agency Notification . . . . . . . . . . . . . . . 1 09-10-82 l 6.0 EVACUATION 6.1 Limited Plant Evacuation . . . . . . . . . 2 01-28-83 6.2 Plant Evacuation . . . . . . . . . . . . . . 2 01-28-83 6.3 Exclusion Area Evacuation . . . . . . . . . . 1 01-28 6.4 Energy Information Center Evacuation . . . . 0 03-31-81 W~

6.5 TSC & OSC Ictivation . . . . . . . . . . . 2 07-01-83

0 Page 2 Revision Date 7.0 CHEMISTRY & HEALTH PHYSICS RESPONSE & PREPAREDNESS 7.1 Internal Chem & HP Group Personnel Notification /

Initial Response 7.1.1 Chem & HP Group Personnel Notification

& Initial Response when Chem & HP Personnel are On-Site ....... 2 04-30-82 7.1.2 Chem & HP Group Personnel Notification

& Initial Response when Chem & HP Personnel are Off-Site . ...... 1 03-17-82 7.1.3 HP Protective Actions by Operations Personnel Prior to Arrival of Chem

& HP Group Personnel . . . . . . . . 1 05-15-81 7.2 Health Physics Facility Activation 7.2.1 Activation of HP Facilities at Site Boundary Control Center ..... 2 03-17-82 7.2.2 Activation of HP Facilities at

Operations Support Center .... 2 05-27-83 7.2.3 DELETED 7.2.4 Health Physics Communications ... 2 05-27-83 ,

7.2.5 ' Control & Use of Vehicles ..... 1 03-17-82

7.3 Radiological Surveys 7.3.1 Airborne Sampling & Direct Dose i Rate Survey Guidelines . . . . . . 3 03-17-82

, 7.3.2 Post-Accident Sampling & Analysis of Potentially High Level Reactor Coolant ............. 6 07-01-83 7.3.3 Post-Accident Sampling of Contain-ment Atmosphere ......... 5 07-01-83 7.3.4 Movement of Required Chemistry Equip-ment & Material to the Technical Support Center Counting Room &

Mini-Laborato ry ...... ... 0 12-30-81 7.4 Emergency Equipment 7.4.1 Routine Check, Maintenance, Cali-bration & Inventory Schedule for Health Physics Emergency Plan Equipment . ........... 7 05-27-83 7.4.2 Emergency Plan Equipment Routine Check, Maintenance & Calibration 7.4.3 Use f Ba rd Modek b3b bingke Channel Iodine Spectrometer to ,

l Determine Airborne Iodine

~ Activities . . . ...... .. 1 05-15-81 n I

o Page 3 Revision Date 7.4.3.1 Use of Canberra Model 3100 Series 30 Multichannel Analyzer to Determine Airborne Iodine Activities .... 0 02-26-82 7.4.4 AMS-2 Air Particulate, Iodine &

Noble Gas Sampler / Detector . . . . 0 03-31-81 8.1 Personnel Assembly & Accountability . ... 3 01-28-83 9.1 Security . ......... ........ 0 03-31-81 10.0 Firefighting . ............... 0 03-31-81 11.0 FIRST AID & MEDICAL CARE 11.1 On-Site First Aid Assistance ........ 3 09-10-82 11.2 Injured Person's Immediate Care . ...... 1 05-15-81 11.3 Hospital Assistance . ............ 3 01-28-83 11.4 Personnel Decontamination . ......... 0 01-29-82 12.0 REENTRY & RECOVERY PLANNING-12.1 Reentry Procedures for Emergency Operations 1 03-17-82 12.2 Personnel Exposure & Search & Rescue Teams t 04-30-82

. ................. 2 12.3 Recovery Planning . . . . . . . . . . . . . . 0 03-31-81 12.4 Personnel Monitoring Exposure Guidelines .. 0 01-29-82 13.0 PRESS 13.1 Crisis Communications . ... ........ 2 09-10-82 14.0 COMMUNICATIONS 14.1 Testing of Communications Equipment . .... 0 03-31-81 15.0 TRAINING, DRILLS & EXERCISES 15.1 Employee Training . . . ... ..... ... 1 09-04-81 15.2 Off-Site Personnel Training . .... .... 0 03-31-81 15.3 Drills & Exercises .... ...... .. . 2 04-30-82 15.4 Emergency Preparedness Review . .. ... .. 0 09-10-82 16.0 WISCONSIN ELECTRIC GENERAL OFFICE PROCEDURES 16.1 Nuclear Engineering Section Notification &

Response . .... .... ...... .. 4 07-01-83

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' TABLE OF EPIP FORMS EPIP EPIP Form Title Procedure 01 Emergency Plan Airborne Radiation Survey Record Site Boundary Control Center (01-83) 7.3.1 .

02 Emergency Plan Survey Record Site Boundary Control Center (01-83) 7.3.1 03 Dose Factor Calculations for Specific Noble Gas Analysis Results (03-81) 7.3.1 04 Status Report on Plant Systems & Controls for Affected Unit (01-83) 1.2 05 Worksheet for Status Report on Radiation Monitoring System for Unit (07-83) 1.2 06 Worksheet for Status Report on Radiation Monitoring System for Plant (07-83) 1.2 07 For X/Q Determination (09-82) 1.4 08 Estimated Whole Body & Thyroid Projected Doses (09-82) 1.4 09 Estimated Whole Body Dose Calculation Worksheet for Specific Noble Gas Releases (09-82) 1.4 10 Estimated Ground Deposition Calculation Worksheet for Particulate,Radionuclide Releases (09-82) 1.4 10a Estimated Population Dose (09-82) 1.8 C',' 11 Summary of Radiological Dose Evaluation Calculations (09-82) 1.4 12 Incident Report Form (09-82) 2.3, 3.3, 4.3, 5.3 13 Status Update Form (09-82) 2.3, 3.3 4.3, 5.3 14 Unused 15 Unused 16 Event Data Checklist (03-81) 5.3 17 Accounting Short Form (04-82) 8.1 18 Assembly Area Roster (03-81) 8.1 19 Drill / Exercise Scenario (03-81) 15.3 20 Drill / Observation Sheet (03-81) 15.3 21 Drill / Exercise Evaluation Report (03-81) 15.3 22 Plant & Company Emergency Call List (01-83) Call List Tab 23 Offsite Agency Emergency Call List (01-83) Call List Tab 24a Site Boundary Control Center Emergency Plan Inventory Checklist (05-83) 7.4.1 24b TSC, ESC, South Gate & OSC Emergency Plan Inventory Checklist (09-82) 7.4.1 24c Emergency Plan Health Physics Supplies at Two Rivers

Community Hospital Inventory Checklist (09-81) 7.4.1 i 24d Control Room Emergency Plan Equipment Inventory Checklist (06-82) 7.4.1 24e Emergency Vehicle Inventory Checklist (04-82) 7.4.1 24f Emergency Plan First Aid Kit Inventory Checklist (09-82) 7.4.1

_ ___ _ _ . _ _ _ . . . _ . __. .___ ~____ . _ _ _ . . _ .

Page 2 m EPIP EPIP Forms Title Procedure 24g Emergency Plan Burn Kit Inventory (09-82) 7.4.1 24h Emergency Plan First Aid Room Inventory (09-82) 7.4.1 241 Emergency Plan Stretcher Inventory (09-82) 7.4.1 24j Emergency Trauma Kit Inventory Checklist (09-82) 7.4.1 25a Emergency Vehicle Checklist (06-82) 7.4.2 25b Monthly Health Physics Instrument & Air Sampler Functional Test Checklist (01-83) 7.4.2 25c Quarterly Emergency Plan Checklist (05-83) 7.4.2 25d Semi-Annual & Annual Emergency Plan Checklist (05-83) 7.4.2 26 Quarterly Communications Test (03-81) 14.1 27 Monthly Communications Test (01-83) 14.1 28 Emergency Plan Instrument Calibration Schedule (06-82) 7.4.2 29 Emergency Plan Counting Equipment & Frisker Calibration Schedule (06-82) 7.4.2 30 Reactor Coolant Post-Accident Sampling Analysis Report (09-81) 7.3.2 31 Containment Atmosphere Post-Accident Sampling Analysis Report (12-81) 7.3.3 32 Search & Rescue and Emergency Operations Checklist (04-82) 12.~2 33 Estimation of Core Damage (09-82) 1.7 34 Calculation of Xe-133 Equivalent Release Rates (07-83) 1.8 35 Dose Calculations (09-82) 1.8

.. 36 Unused ,

37 Medical Assistance Call List (09-82) 11.1 l

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t EPIP 1.1 MINOR Revision 5

. 07-01-83

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INITIAL CLASSIFICATION 1.0 GENERAL The purpose of this procedure is to provide a means of classifying an event or condition at the Point Beach Nuclear Plant into one of four emergency classifications as described in the Point Beach Nuclear Plant Emergency Plan. Each emergency classification requires emergency organization noti-fications, mobilizations, and actions to be performed in order to appro-priately react to the situation and provide for the health and safety of plant and public personnel. They are listed in order of increasing severity.

1.1 Unusual Event An unusual plant condition which either has occurred or might occur.

This condition could possibly lead to a degradation in overall safety.

This condition does not represent a significant radioactivity release, involves no offsite response, and may require no augmentation of plant personnel. In spite of the above, prompt notification of the r _

counties 'and state is required.

(~ 1.2 Alert Plant conditions in which events are in progress or have occurred which involve an actual or potential degradation of plant safety.

Radiation releases are not likely to cause an offsite hazard. Prompt offsite notification is necessary and the plant organization may have l to be augmented.

l l.3 Site Emergency i

Plant conditions in which events are in progress or have occurred which involve actual or probable major failures of plant functions.

Potential radioactive releases may have an impact on offsite people.

Prompt notification of offsite agencies is required. The plant organization must be augmented and the technical support center, onsite operations support center, and emergency support center will be operational. An evacuation may be necessary.

l 1.4 General Emergency Plant conditions in which events are in progress or have occurred which involve actual or imminent substantial core degradation and a potential for loss of containment integrity. Potential radioactive releases may have an impact on offsite people. Prompt notification

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. EPIP 1.1 Page 2 A

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of offsite agencies is required. The plant organization must be auginented and the technical support center, onsite operations .:upport center, and emergency support center will be operational. An evacu-ation may be necessary. The emergency news center will be opened.

i The Shift Supervisor has the responsibility and authority to take immediate action to mitigate the consequences of the emergency. He will consult with the Duty & Call Superintendent and assign the appropriate emergency classi-

, fication and initiate the necessary Emergency Plan implementing procedures.

2.0 REFERENCES

2.1 Nuclear Regulatory Commission NUREG-0654, Resision 1, published November, 1980.

4 2.2 Point Beach Nuclear Plant Emergency Plan Sections 4.1 and 5.1.

3.0 PRECAUTIONS AND LIMITATIONS 3.1 All actions and notifications should be appropriately logged.

3.2 Emergency Plan implementing procedures are not to be used to respond to security-threats. One hour notification of the NRC is required

(' using the red phone for security threats.

3.3 Certain events require notification to the NRC within one hour.

These items are included on Table 1-1. Those items which are noted as "NRC Only" means that there is no classification for the events and no notification other than the NRC is required. These notifi-

~ cations are not considered as starting the Emergency Plan.

3.4 The " Indications Used" in Table 1-1 are those which one may expect if that level of emergency occurs very quickly. For more slowly developing situations, other indications may be judged appropriate.

For example, a primary system leak rate of 40 gpm is an Unusual Event. Subsequently, charging may be lost and, in addition, the leak may worsen. One may not see charging flow 50 gpm greater than letdown flow when in fact an Alert should be declared.

3.5 For classification purposes where radiological dose is the primary parameter leading to a classification, use EPIP 1.8, " Emergency Off-Site Dose Estimation" for determination of dose.

4.0 INITIAL CONDITIONS - '

None.

NOTE: APPENDIX 1 0F NUREG-0654 (PAGE l-3) CONTAINS THIS SENTENCE: "THE TIME IS MEASURED FROM THE TIME AT WHICH OPERATORS RECOGNIZE (EMPHASIS ADDED) THAT EVENTS HAVE OCCURRED WHICH MAKE DECLARATION OF THE EMER- '

GENCY CLASS APPROPRIATE.

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EPIP 1.1

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5.0 PROCEDURE 5.1 Call the Duty & Call Superintendent for consultation to establish the initial classification. If he is unavailable, the shift Supervisor is responsible for classification.

5.2 Select affected categories related to plant events or conditions at this time. Check (J) all applicable categories.

Refer to Page Category in Table 1-1

1. _

_ Safety Systeu FLn:tiene 1

2. Abnormal Primary Leak Rate 1

3. Abnormal Coolant Temperature /

Pressure 2

4. Abnormal Primary / Secondary Leak 2

5. Core Fuel Damage 3

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6. Secondary Coolant Anomaly 4

7. Abnormal Effluent 5 l 8. Major Electrical Failures 5 l

9. Control Room Evacuation 6

10. Fire 6

11. Plant Shutdown Function 7

12. Abnormal Radiation Levels at Site Boundary 8

13. Fuel Handling Accident 8

14. Serious or Fatal Injury 9

15. Security Threat 9

16. Hazards to Plant Operations 9

17. Natural Events 10

18. Reactivity Transient 10 l

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EPIP 1.1 Page 4

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Refer to Page Category in Table 1-1

19. Load Transient 11

20. Other 11 5.3 Beginning at the indicated page in Table 1-1 (attached), review initiating conditions for all categories checked above. .

5.4 Record most severe emergency classification at this time.

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5.5 Record date/ time of initial classification (subsequent columns for reclassification at a later date and time are provided if reclassifi-cation is required).

Initial Subsequent Subsequent Date/ Time Date/ Time Date/ Time

, NOTE: IF THE SHIFT SUPERVISOR CANNOT COMMUNICATE WITH A DUTY &

( _. CALL SUPERINTENDENT, THE SHIFT SUPERVISOR MUST NOTIFT THE

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STATE AND TWO COUNTIES WITHIN 15 MINUTES OF THE DECLARATION OF ANY CLASS OF EMERGENCY.

5.6 If events or conditions are classified as an Unusual Event, perform EPIP 2.1, " Unusual Event - Immediate Actions."

5.7 If events or conditions are classified as an Alert, perform EPIP 3.1,

" Alert - Imediate Actions."

5.8 If events or conditions are classified as a Site Emergency, perform EPIP 4.1, " Site Emergency - Immediate Actions."

5.9 If events or conditions are classified as a General' Emergency, perform EPIP 5.1, " General Emergency - Immediate Actions."

NOTE:

"One hour" refers to the requirement to notify NRC within one hour (10 CFR 50.72).

"One hour - Open line" refers to the requirement to notify NRC within one hour and maintain an open line for continuous communication (10 CFR 50.72).

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Notes: DCS - Duty & Call Superintendent i DSS - Duty Shift Supervisor TABLE l-1 FSAR - Final Safety Analysis Report .

MASP - Modified Amended PBNP Security Plan i

EMERGENCY CLASSIFICATION Energency catequry Initiating condition Indication used classification

1. ' Safety System Functions unplanned initiation of emergency core cooling Any of the following first-out reactor trip Unusual Event with injection to the primary system panel annunciation with indicator confir- 1-Hour (7) nation noted:

I e. ;

1. " Containment press hi", j

[PI-945, PI-947, PI-949 (2/3 >5 psig)] i

2. " Steam line loop A lo-lo press" (PI-468, PI-469, PI-482 (2/3 <530 psig)]

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3. " Steam line loop 8 lo-lo press"

[FI-478, PI-479, PI-483 (2/3 <530 psig)] .,

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4. "Pressuriser lo press $1"

[FI-429, PI-430. PI-431 (2/3 <!?35 psig)l 5.. Wide range pressure <!500 psig l

Loss of containment integrity requiring When shutdown commences as determined by DSS Unusual Event (

l 1-Hour (5) shutdown by Technical Specifications and DCS !

Loss of engineered safety feature requiring When shutdown commences as determined by DSS Unusual Event j shutdown by Tecimical Specifications and DCS 1-Hour (5) .

I Loss of fire protection system function When shutdown commences as determined by DSS Unusual Ev'ent requiring shutdown by Technical Specifi- and DCS 1-Hour (5) cations (i.e., both fire pumps inoperable)

and no backup fire suppression system

2. Abnormal Primasy Exceeding Technical Specification primary system When shutdown commences as determined by DSS Unusual Event Leak Rate leak rate (10 ge) and DCS 1-Hour (5)

I (5) (7) See DCS 1.12.1 for 10 CFR 50.72 Natifications. 1 i

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Table 1-1 j Page 2 Emergency I Category Initiating Condition Indication Used Classification !

Leak rate >50 g a A J of the following: Alert

1. " Volume control tank level hi-lo"

[LI-lit and/or LI-ll2 (8%]

2. Decreasing pressurizer level

,, [LI-426,LI-427,LI-428]

I 3. " Charging pump speed hi"

4. Charging line flow (FI-128) >50 mm more than letdown flow (FI-134)

Leak rate in excess of available pump y of the following:

A Site Emergency capacity including charging, high head SI ,

and low head SI 1. " Volume control tank level hi-lo

[LI-141 and/or LI-il2 <8%)

2. No pressurizer level indicated (LI-426,L1-427,Ll-428] '

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3. All available pumps running as indicated by f the red light at the switch

4. Increasing core exit T/C temp as indicated by P-250 and confirmed on local readout.

3. Abnormal Coolant unexpected decrease in subcooling margin Both of the following: Unusual Event Temperature / Pressure

1. Alarm on P-250, if operable I

2. Confirmation by manual calculation M of the following:

Pressure >2735 psig l. Pressure >2735 psig on PR-420, and NRC only l l-hour open line (2) [

DNBR <!.30 2. Code safety, or PokV not closed indication >

4. Abnosmal Primary / Exceeding Tectunical Specification primary-secondary When shutdown commences as determined by DSS Unusual Event Secondary Leak leak rate and DCS 1-Hour (5)

(2) (5) See DCS 1.12.1 for 10 CtR 50.72 Notifications.

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Emergency Category Initiating condition Indication used Classification Gross failure of I SG tube (>400 @m) & loss of All of the following first-out reactor panel Alert offsite power (FSAR 14.2.4) annunciation with confirmation indications

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1. " Pressurizer to Press SI,"

(PI-429, PI-430, PI-431 (2/3 <!?35 psig)]

2. a. " Steam generator A level hi" r lL1-461, L1-462, LI-463 (2/3 >70%)] og

', b. " Steam generator B level hi" 1

' [LI-471, LI-472. LI-473 (2/3 >70%))

3. a. "4.16 kv bus undervoltage" ?

A. & 0 volts on A03 & A04 voltmeters. l

b. M04 to A03 ammeter on CO2 (0 amps) +

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c. XO4 to A04 ammeter on CO2 (0 amps) j

4. SI flow >400 w e indicated by FI-924 & '

I F1-925 and pump discharge pressure I corresponding to flow.

Rapid failure of >10 SG tubes (4000 gan) with A,ll of the following first-out reactor panel Site Emergency j i

i or without of fsite AC , annunciation with confirming indication: l

1. " Pressurizer lo press SI" .

I lP!-429 PI-430, PI-431 (2/3 <!735 psig)]

2. a. " Steam generator A level hi" ,

[LI-461, LI-462 LI-4o3 (2/3 >70%)] 4 or

)

b. " Steam generator B level hi" '

lLI-471, LI-472, LI-473 (2/3 >70%)l

3. SI flow >4,000 w e indicated by FI-626 FI-928, FI-924 & F1-925

5. Cos e Fuel Damage Gsoss fuel damage in. core indicated No. ! & 3 or No. 2 & 3 of the fo!!owing: Unusual Event .

1. Letdown line radiation monitor (IR9)

(sample line R109) 100 x alarm setpoint.

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2. Failed fuel monitor (2RE-109) 100 x

alarm setpoint.

i 3. Sustained offscale & chemical analysis shows.

d fission product concentratica increase by 100x.

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Table 1-1 l Page 4 Emergency Categoy Ini'iating Condition Indication used classification 4

Massive fuel damage 300 pCi/cc iodine-equivalent as determined by Alert by chemical analysis

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] 1. Massive loss of fuel clad integrity Initiating Conditions Nos. I, 2, 4 & 5 General Emergency f

! 2. With simultaneous loss of primary system integrity exist and No. 3 is possible: 1-Hour open line (3)

3. With potential loss of containment integrity ,,

4. 300 pC1/cc lodine-equivalent determined by ;

F chemical analysis *

2. Primary system leak >1000 mm indicated by SI flow >1000 gu (F1-924 & F1-925) and pump !

s. discharge pressure corresponding to flow 1

3. Manimum containment pressure suppression j equipment is not available (any of the ,

fo!!owing):

4

a. No fan cooler operating and <2 spray pumps. ;

b. No spray pump operating and <4 fan coolers

c. (2 fan coolers running with I spray pump -

l l 4. " Containment press hi" '

i I lPI-945, PI-947, PI-949 (2/3 >5 psig)l '

j 5. " Containment spray" with 2/3 + 2/3 >25 psig lPI-945,PI-947,PI-949)

(PI-946,P!-948,PI-950) l !

6. Secondary Coolant Reduction in feedwater enthalpy incident t

1. a. Decreasing feedwater temp indicated by Unusual Event ;

Anomaly (FSAR 14.1.7) 70-4184 & 70-438A on P-250 aj

b. confirmed by local temperature indicator on outlet of No. 5 feedwater heater.

2. Unexpected increasing power on encore nuclear <

instrumentation j Steam line break with primary to secondary leak A Q of the following first-out reactor trip Alert a rate in excess of 10 g m (FSAR 14.2.5) panel annunciation with confirmation: 1-Hour open 119 (3)I

1. Either:

a. " Steam line loop A Lo-Lo press"

[PI-468, PI-469, PI-482 (2/3 <530 psig))

or

b. " Steam line T~op o B Lo-Lo press" lPI-478, PI-479, PI-483 (2/3 <530 psig)l (3).Sgh DCS 1.12.1 for 10 CfR 60.72 Notification.

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Table 1-1 Fage 5 Emergency Categog Initiating Condition Indication used Classification

2. Confirmed primary-to-secondary leak rate of at least to was.

3. Either:

a. " Steam line loop A isol channel alert" (F1-464. F1-465 (1/2 >3.9x10' lb/hr)l or

b. " Steam line loop a isol channel alert" ,

r lF1-474, F1-475 (1/2 >3.9x10s Ib/hr)] ,

Secondary Coolant Transient initiated by loss of feedwater, fo!! owed ~All of the following: General Emergency s,Suomaly by loss of auxiliary !eedwater for >l hour 1-Hour open line (3)

(FSAR 14.t.It) 1. Decreasing SG levels -

"A" SG (L1-461, LI-462 LI-463]

"B" SG lL1-474,L1-472,L1-473) s

2. No auxiliary feedwater flow -

(F1-4002, F1-4007, F1-4014] j (F1-4036,F1-4037]

7 Abnormal Effluent Radiological effluent Technical Specification !!aits Airborne effluents only Unusual Event exceeded but (10 times the limit (FSAR 14.2.3) 1-Hour (8)

Radiological effluent Technical Specification limits Liquid effluents only Unusual Event exceeded (FSAR 14.2.2) 1-Hour (8) hadiological effluents >10 times Technical Airborne effluents only Alert Specification instantaneous limits. (An instan- 1-Hour (8) taneous rate which, if continued for >2 hours, would result in a dose of about I mR at the site boumlary umler tverage meteorological conditions.)

8. Major Electrical Sustained loss of offsite power >l5 minutes A of the following:

J Unusual Event Failures (FSAR 14.I.2)

1. "4.16 kv bus undervoltage" is 0 volts on A03 4 A04 voltmeters.

2. XO4 to A03 ammeter on CO2 (0 amps).

3. XO4 to A04 ammeter on 002 (0 amps) j 4

(3) (8) See DCS 1.12.1 for 10 CtX 50.72 Notification.

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Table 1-1 l Page 6 0

Emergency Category Initiating condition Indication used Classification Sustained loss of onsite AC power capability B of the following:

g Unusual Event

(>l3 minutes) 1-Hour (5)

1. a4.16 kv bus undervoltage" &

O volts on A05 and A06 voltmeters

2. " Emergency Diesel Starting System Disabled" for both Diesels .

Loss of all vital onsite DC power >l5 minutes B of the following:

M Site Emergency

1. " Annunciator power failure" on 001, C02, C03, and C04

s ,

2. <100 volts on the voltmeters for all batteries i Loss of offsite power and loss of all onsite AC AJ of the following: Site Emergency [

j power for >l5 minutes i

1. "4.16 kv bus undervoltage" J volts on ,

A03, A04, AOS, A06 & "Emerg Diesel starting system disabled" for both Diesels

2. XO4 to A03 ammeter on CO2 (0 amps)

3. XO4 to A04 ammeter on CO2 (0 amps) 1.oss of offsite and all onsite AC power with loss of A of the fo!!owing:

J General Emergency all auxiliary feedwater for >2 hours

1. Unit aux MW meter XO2 on CO2 (O MW)

2. Station aux MW meter XO4 on CO2 (0 NW)

3. XO4 to A03 ammeter on CO2 (0 amps)

4. XO4 to A04 ammeter on CO2 (0 amps)

6. XO2 to A01 ammeter on CO2 (0 amps)

7. a. No aualliary feedwater flow

[FI-4036, FI-4037]

b. Decreasing SG level -

"A" SG lLI-461, L1-462, LI-463l "B" SG lLI-471, LI-472, LI-473]

9. Control Hoom Evacuation Evacuation of control room >15 minutes & As required by DSS Site Emesgency no control at remote shutdown station I-tiour open line (3)

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10. Fase Fire in vital area or on the controlled side of plant As reported by Fire Brigade Chief unusual Event i lasting >10 minutes aften initial use of fire extinguishing equipment.

(3) , '.I) See DCS 1.12.1 for 10 Ctk 50.72 Notification.

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Table 1-1 Page 7 Emergency C

Agoy Initiating condition Indication Used Classification !

Fire affecting I train of safety systems. As reported by Fire Brigade Chief Alert Fire af fecting 2 trains of safety systems As reported by Firc Brigade Chief Site Emergency g II. Plant St.ut4 wn function Nonfunctional indications or alarms in the contro),, B of the following:

M Unusual Event f suam on primary system parameters requiring plant 1-pour (5) I

'. l. " Annunciator power failure" on C04.

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2. Failed indication as determined by DSS.

.J . Turbine mechanical failure with consequences @ of the following:

1. Annunciator " Turbine supe.visory." Unusual Event i

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2. Indication on TR-6019 of bearing vibration

>7 mils. l

3. Bearing vibration alarm on back of CO3.

4 4. Visual confirmat' ion of turbine housing penetration by a blade or disc.

i Significant loss of effluent monitoring capability & !. Loss of RE-223 during a release unusual Event meteorological instrtaments which impairs ability to or perform emergency assessment. Loss of effluent 2. Loss of RI8 (RE-218) during a release monitoring may/may not require plant shutdown. or

! 3. a. Loss of wind speed indication or wind direction indication and

b. Loss of Bl4 and RMS Il Channel I (RE-315. RE-317, RE-319) or j-

c. Loss of R15 (RE-215) and CR9 and lots Il Channel 5 (RE-225, RE-226) or

d. Loss of R21 (RE-221) and Rits !! Charunel 2 (RE-325, RE-327) '

or

c. Loss of RE-224 and Rits II channel 6 Failure of reactor protection system to A of the following:

J AlerO complete a trip which brings reactor syberitical 1-Hour open line (3)

1. Unplanned first out annunciator on C04 with confirmaticas from associated indicator

2. Intermediate range detector output not decaying

3. >l RCC RPI indicates fully withdrawn (3 ) See l#S L.12.1 Bor 10 Y 50.72 Notification

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Table 1-1 Page 6 i

J Emergency ?

Category Initiating condition Indication used Classification All alarms (annunciators) lost >15 minutes while unit 1. " Annunciator power failure" on Col, CO2 & Alert is not in cold shutdown 1(2)C03,1(2)C04 loss of functions needed for cold shutdown for >4 A_ng of the following: Alert hours while at cold shutdown

,, 1. Loss of service water Unit 1 = south & west header

Unit 2 = north & west header

2. i.oss of both trains of RHR

3. l.oss of component cooling

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! 12. Abnormal Radiation a. Effluent monitors detect levels corresponding to Altborne effluents only site Emergency Levels at site any of the followings f

, (1) >50 mR/hr for % hour .

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! (2) >250 mR/hr for % hour for the thyroid (3) >500 mR/hr whole body for 2 minutes {

(4) >2500 mR/hr to the thyroid for 2 minutes at the site boundary for adverse meteorology

b. Any of the above doses measured in the environs As reported to DSS by MP Supervisor a

c. Any of the dose rates projected, based on plant parameters

a. Effluent monitors detect levels corresponding to 41rborne effluents only General Emergency either:

li) 1 R/hr whole body

(2) 5 R/hr thyroid [

j at the site boundary under actual meteorological :

1 conditions - I

b. Either of the above doses measured in environs As reported to DSS by HP S e rvisor i

c. Either of above dose rates projected based on ;

, other plant parameters j l a j 13. Fuel Handling Accident Major damage to irradiated fuel in containment Both of the following: Alert i

1. As reported to DSS by Core Loading Supvr.

6 4 2. Alarm on Victoreen on manipulator & alarm on R211 t

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Emergency Category Initiating Condition Indication used Classification Fuel damage accident with release of radioactivity Both of the following: Alert to auxiliary building (FSAR 14.2.1)

4. As reported to DSS by Supvr in charge of fuel handling & drimusing area vent (R221) g

2. Alarm on Victoreen on spent fuel pit bridge. j i

'14. CSerious or Fatal Injury Transportation of seriously or fatally injured Reported as judged by DSS Urassual Event {

individual from site to hospital (espect hospitalisation for at least I (Reference EPIP !!.8) 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months ) f s o..

15. Security Thseat Security threat or attempted sabotage Per MASP Per MASP & Appendices or 1-Hour Red Phone Only Ongoing security compromise (open Line) (4)

It . Hazasds to Plant Unusual aircraft activity over facility Visual observation of Operations Supervisor of Unusual Event operation security force Near or onsite emplosion or flammable or tonic gas As reported to DSS by plant personnel making Unusual Event !

release visual observation Missile impacts from any source on facility visual observation by operations Supervisor Alert Missile impact causing damage to two trains of safety systems Visual observation by operatinns Supervisor Site Emergency Aircraft crash in protected area (within the fence) Visual observation b) Operations Supervisor Alert ,

Known explosion damage to facility atfecting plant Visual observation by operations Supervisor Alert operation. Toxic or flammable gases in facility i, environment excluding normal process gases Toxic or flammable gases entering into vital areas Visual observation by operations Supervisor Site Emergency (control room, auxiliary building, etc.) excluding normal process gases (4) See DCS !.12.1 for 10 CFk 50.72 Notification.

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Table 1-1 Page 10 t

Emergency Category Initiating Condition Indication used Classification

47. Natural Evesits Any eartlequake Activation of >2 accelerographs and verified by actual physical ground shaking or by con-tacting Dr. David Wi!!!s, University of Wisconsin,Itilwaukee Seismic Center at 1-484/ % 3- % 02. thiusual Event Any tornado visible from site Verification by Operations Supervisor Unusual Event r

, Low 1.ake Hicisigan water level With no CW pumps running, water level is 3.9' Unusual Event below O' on surge chamber level & confirmed by '

e' . - measuring forebay level at 10.9' below pumphouse floor (7' level) i Earthquake greater than operating basis earthsysake Earthquake with attendant structural damage of Alert I contaisunent or spent fuel pit Any tornado striking the facility Visual observation by Operations Supervisor Alert Seict.e near design level >6" of water in turbine hall Alert ',

1 I

Winds in excess of design levels Wind speed indicated as >100 mph Alert T 4

Wind with damage Structural damage to contairunent Site Emergency

'i' Failure of protection for vital equipment at low Any of the following: Site Emergency levels (i.e., caused by seiche > design levels) j 1. >3' water in both EDG rooms.

2. >2' water in vital switchgear roosa.

3. >2' water in auxiliary feed pump room.

18. Reactivity Transient Uncontrolled rod withdrawal (FSAR 14.1.1 de 14.1.2) Unusual Event CVCS Malfusiction (FSAR 44.4.5) Usiusual Event Accidesatal Criticality NRC Only (3)

(3) ee DCS 1.12.I for 10 CFR 50.72 Notification.

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Table I-!

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Emergency Category Initiating Condition Indication used Classification

19. Load Transient Loss of Electrical Load (FSAA 14.1.9) Unusual Event 20 '. Other Condition that warrants State ar.d/or local official DCS & DSS concurrence Unusual Event awareness Cosulitionthatwarrantsestablistmentoftechn[ cal DCS & DSS concurrence Alert f support center & emergency support center Condition that warrants use of monitoring tenans DCS & DSS concurrence Alert 1,

Persosunel contamination Health Physicist & DCS concurrence NRC-only 1-hour (10) i Any unplanned reactor trip DCS & DSS concurrence NRC-only 1-Hour (?)

Strike by employees or guard force DCS & DSS concurrence NRC-only 1-Hour (12)

Loss of red phone (ENS) DCS & DSS concurrence NRC-only 1-Hour (13)

Personnel or procedural error DCS & DSS concurrence NRC-only 1-Hour (6) 10 CFR 20.403 DCS & DSS concurrence NRC-only l-Hour (!!)

(6) (7) (10) (11) (12) (13) See DCS 1.12.1 for 10 CFR 50.72 Notification.

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EPIP 1.2 MINOR

[ Revision 1 07-01-83 PLANT STATUS 1.0 GENERAL The purpose of this procedure is to provide a checklist of contrS1 room instrumentation and parameters for various plant systems to assist in the determination of the plant emergency status and conditions.

2.0 REFERENCES

2.1 U.S. NRC Regulatory Guide 1.97, " Instrumentation for Light-Water-Cooled Nuclear Power Plants to Assess Plant and Environs Conditions During and Following An Accident," Revision 2, June 4,1980.

3.0 PRECAU'* IONS AND LIMITATIONS None 4.0 INITIAL CONDITIONS None 5.0 PROCEDURE 5.1 shift Supervisor / Plant Operations Manager 5.1.1 Designate an individual, such as the Operating Supervisor or Control Room Operator, to perform Section 5.2 of this procedure.

5.1.2 Designate an individual, such as the Auxiliary Operator or Duty & Call Superintendent, to perform Section 5.3 of this procedure.

5.1.3 Review all completed attachments performed in Section 5.2 and 5.3 of this procedure to evaluate plant conditions as to status of the emergency and plant safety.

5.1.4 Determine if reclassification of the emergency is necessary (refer to EPIP 1.1, " Initial Classification") and notify respective plant supervisors and managers of significant problems associated with the plant.

5.1.5 Continue to assess and re-evaluate the plant conditions until plant conditions are stabilized to allow the termi-nation or relaxation of the emergency.

l

3 l t EPIF 1.2 Page 2 5.2 Designee 1 5.2.1 Complete form EPIP-04 (attached) by evaluating and monitoring available control room instrumentation.

5.2.2 Forward the completed attachment to the Shift Supervisor.

He will see that the information is forwarded to the Plant Operations Manager, technical support center, and emergency support center as soon as possible.

5.2.3 Continue to update form EPIP-04 by re-evaluating and moni-toring appropriate control room instrumentation.

5.2.4 Relay updated and pertinent information and variables to the Shift Supervisor. He will see that the information is forwarded to the Plant Operations Manager, technical support center and emergency support center.

5.3 Designee 2 5.3.1 Complete forms EPIP-05 and EPIP-06 (attached) by evaluating ,

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and monitoring available readings from the radiation moni-toring system in the control room or technical support icenter.

NOTE 1: IT SHOULD BE FULLY UNDERSTOOD THAT INDIRECT DETERMINATION OF DOSE RATES BASED UPON READINGS FROM THE INSTALLED AREA AND PROCESS MONITORS ARE ESTIMATES ONLY. THESE ESTIMATES SHOULD BE VERIFIED AND SUPPLEMENTED WHENEVER POSSIBLE SY DIRECT DOSE RATE SURVETS AND AIR SAMPLING.

NOTE 2:. ATTACHMENTS 1.2-1, 1,2-2 & l.2-3 LISTINGS OF RMS INSTRUMENTS BY PLANT LOCATION, PROVIDES BASIC INFORMATION.

5.3.2 Forward the completed attachments to the Shift Supervisor.

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He will see that the information is forwarded to the Plant Operations Manager, technical support center and emergency support center as soon as possible.

5.3.3 Continue to update forms EPIP-05 and EPIP-06 by reevaluating and monitoring appropriate readings from the radiation monitoring system in the control room or technical support center.

5.3.4 Relay updated and pertinent information and variables to the Shift Supervisor. He will see that the information is forwarded to the Plant Operations Manager, technical support center and emergency support center.

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i ATTAODENTS 1.2-1, 2 & 3 The purpose of these attachments is to provide guidelines to aid in evalu-ating in-plant radiation dose rate estimates and the potential for airborne activity using the installed RMS instrumentation.

Area monitors calibrated in mrem / hour provide direct indications of area radiation dose rates. Process type monitors calibrated in counts-per-minute may be used to estimate area dose rates by converting the readout in counts-per-minute to mrem / hour. The conversion factor is approximately one mrem / hour for each 1,000 cpm indicated.

Vent stack monitors which measure noble gas discharges from plant areas should always be considered to indicate potential airborne activity.

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ATTACHMENT 1.2-1 LISTING OF RMS AREA MONITORS BY PLANT LOCATION j Monitor Location Indication IR9' Letdown Line Auxiliary building El. 26'. North Provides indication of reactor coolant end of demineralizer valve gallery. activity. Indicates potential dose rate ;

d.- levels in the area of the volume control 6 tanks, holdup tanks, letdown gas strippers, j related piping, and No. 2/No. 3 pipeway lower level. ;

i 1R4 Charging Pump Hallway Auxiliary building El. 8'. East Indicates dose rates in the hallway side of charging pump hallway. east of the charging pump cubicles.

IR6 Primary Sample Room Auxiliary building El. 26'. Inside Indicates dose rates inside sample rooms. l of primary sample room. Of minimal use unless sample system is in !

use. t i

1R22/2R22 Blowdown Tank Facade El.'26' next to blowdown tank. Provides dose rate indication around the blowdown tanks and facade El. 26'. The readout for these monitors is located !

on the blowdown evaporator control panel I at C59 panel area.

H3 Chemistry Lab Southeast corner of chemistry lab. Provides indication of dose rates in the j s

chemistry lab and associated hallways. ,

R5 Spent Fuel Pit Northeast side of the spent fuel Provides indication of dose rates in the pit, El. 66' auxiliary building. vicinity of the spent fuel pit. This monitor will be affected by high contain-ment radiation levels. 2 i

i R8 Drumming Area Inside of Atcor waste processing Provides dose rates within the Atcor cubicle. ,

cubicle.

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l Attachment 1.2-1, Cont'd. i 1 t j Monitor Location Indication R1 Control Room West side of control room. Indicates dose rates in the area of control room.

' I 1R2" Containment El. 66' Located inside of containment on Provides dose rates within containment '

Area Monitor east side of El. 66' near hatch. around east side of El. 66' near hatch. l IR7' Seal Table containment El. 46' by the seal Provides dose rates.in general area of table, seal table.

I RE-101 Control Room West wall of control room. Shifts control room ventilation to 100%

recirculation.

1(2)RE-102 Containment El. 66' near access hatch on east Provides dose rates within containment Low Range side. around access hatch. I RE-103 Chemistry Lab East wall of chemistry lab near Provides indication of dose rates in counting room door. chemistry lab and associated hallways.

1(2)RE-104 Charging Pump Hounted on west side of shield wall Indicates dose rates in hallways east of

, Room Low Range east of cubicles on El. 8' of aux. charging pump cubicles. !

building. I i

RE-105 Spent Fuel Pit Mounted on railing just northeast Provides indication of dose rates in the Low Range of spent fuel pit on El. 66' of aux, vicinity of the spent fuel pit. This i building.'

monitor is affected by high radiation !

levels in containment. !

1(2)RE-106 Primary Side Mounted on west wall, towards north When sampling system is in operation, it Saq)le Room Low Range corner of sample room on El. 26' of indicates dose rate inside sample room.

aux. building.

1(2)RE-107 Seal Table Mounted on wall just above seal table Provides an indication of general area on El. 46' of containment. dose rate near seal table.

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Attaciunent 1.2-1, Cont'd.

Monitor Location Indication I i

! RE-108 Drumming Station Mounted inside the Atcor waste Provides dose rate indication within the processing cubicle. ,,

drumming station. ,

1(2)'RE-109 Failed Fuel Mounted on south wall near east Provides an indication of failed fuel by 1

corner of primary side sample room monitoring the primary coolant sample J

, on El. 26' of aux. building. activity. ,f us..

RE-Il0 SI Pump Room Located on north wall just west of Provides an indication of the dose rate

passageway in SI pump room. in general area of SI pumps.

i RE-111 C59 Panel Mounted on top of C59 instrument Provides general area dose rate near C59 '

panel on El. 26' of aux. building. panel.

RE-112 Central PAB Hounted on north wall just east of Indicates general area dose rate on El. 8' pipeway No. 3 on El. 8' of aux. of aux. building.

building.

RE-113 El. -19' Auxiliary HoNated in general area of El. -19' Provides an indication of the dose rate in Building of aux. building. aux building sump and general area of El. -198 f RE-Il6 Valve Gallery Mounted by north entrance to valve Indicates general area dose rate in letdown *

, gallery on El. 26' of aux. building. valve gallery.

i ;

! 1(2)RE-134 Charging Ptump Hounted next to 1(2)RE-104 on west Provides an indication of general area dose ;

I Room liigh Range side of shield wall, east of cubicles rates in the event low-range monitor pegs I on El. 8' of aux. building. offscale high.

~

RE-135 Spent Fuel Pit Mounted next to 1(2)RE-105 on railing Provides an indication of general area ,

liigh Range 4 just northeast of spent fuel pit on dose rates in the event low-range El. 66' of aux, building. monitor pegs offscale high.

1(2)RE-136 Primary Side Mounted next to 1(2)RE-106 on west Provides an indication of general area ,

Sample Room liigh Range wall, towards north corner of sample doce rates in the event low-range

] g room on El. 26' of aux. building, monitor pegs offscale high.

,- 6 i

j At taciunent 1.2-1, Cont 'd.

,! Monitor Location Indication ,

i HE-140 SI Pump Room High Mounted next to RE-110 on north Provides an indication of general area Range wall just west of passageway in dose rates in the event low-range SI ptump room. monitor pegs offscale high. ;

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ATTACHMENT 1.2-2 4

LISTING OF RMS PROCESS MONITORS BY PLANT LOCATION i 1

1 Monitor Location Indication 1R16 Containment Fan Central section of aux. building, Provides indication of possible contamination Cooler Liquid Monitor El. 26'. of service water from the containment cooling ',

fans. May be used to estimate dose rates in !

the central section of the aux. building. l j 1R19 Steam Generator Blow- Outside of each primary sample room, Provides indication of steam generator blow-

down Monitors aux. building El. 26'. down activity. May be used to estimate dose j rates in vicinity of the monitor and aux.

building El. 26' hallways.

l 1R17 Component Cooling East side of aux. building central Provides indication of component cooling , ,

Liquid Monitor section, El. 8'. system liquid activity. May be used to

)

estimate dose rates in the vicinity of the monitors. May also be an indication of dose

.; rates around the component cooling tanks and heat exchangers on aux. building El. 468 I

R18 Waste Condensate Over- Aux. building El. 8', east side,' Monitors activity levels of waste water !

board Monitor near component cooling water pumps. being discharged overboard. May be used '

to estimate dose rates in the central section of aux. building El. 8'. i I

R20 Spent Fuel Pit Heat Aux. building El. 46', north side. Indicates activity of service water from the Exchanger Service Water spent fuel pit heat exchangers. May be used out Monitor to estimate dose rates in vicinity of the i Unit 2 facade. Due to location and shielding, this monitor should not be used to estimate i dose rates originating from Unit I contain-2 ment.

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Attachment 1.2-2, Cont'd.

Monitor Location Indication t

IRll Containment & Purge Facade, El. 52', east side. Indicates particulate activity in contain- '

Stack Part.iculate Monitor

ment and purge stack. May be used to indicate dose rates in facade, containment purge fans and purge stack.

IR12 Containment & Purge Facade, El. 52', east side. Indicates gaseous activity in containment st'ack Gas Monitor and purge stack. May be used to indicate dose rates in facade containment purge fans and purge stack.

l The following process monitors may be used in evaluating potential airborne contamination levels within the l plant. Whenever there are indications of high airborne activity being diecharged through a vent stack, there is always the possibility of in-plant airborne contamination.

R14 Aux. Building Vent In aux. building vent stack, Unit 1 Indicates gaseous releases from within Stack facade, just south of elevator. the aux. building. May indicate potential aux. building airborne activity.

l R21 Drumming Area Vent Northwest corner of Unit 1 facade Indicates gaseous releases from spent

Stack near roof. fuel pit area and drumming station. Indi-cative of potential aux. building airborne activity.

1R15/2R15 Steam Air Ejector West side of turbine building Indicative of steam generator primary-to-Monitor El. 46'. secondary leakage. May be indicative of i potential radiation exposure sources l' within turbine building.

i 1(2)RE-211 Contairunent Located in cubicle on east side of Indicates particulate activity inside con-Air Particulate El. 52' of containment facade. tainment, facade, or purge exhaust stack.

There are no control functions associated with .

this monitor. .

i 1(2)RE-211B Containment Located next to 1(2)RE-211 in Provides background activity for both 1(2)RE-211 Background cubicle. and 1(2)RE-212 for background subtraction, if used.

W f

Attactunent 1.2-2, Cont'd. l Monitor Location Indication f 1

I 1(2)RE-212 Containment Located in series with 1(2)RE-211 Provides indication of containment noble j Noble Gas on detector skid in the cubicle on gas activity. Isolates containment l El. 52' of containment facade." ventilation upon high activity.

I 1

RE-214 Aux. Building Vent Mounted on aux. building exhaust Indicates any gaseous release from spent Stack Noble Gas stack at about El. 80' in Unit 1 fuel pit area and the drumming station.

^

) facade just south of elevator. Indicative of potential aux. building air-

, borne activity. Shuts the vent gas release

[ valve and initiates aux. building exhaust

filtration.

j 1(2)RE-215 Condenser Air Mounted on west wall of El. 46' in Indicative of steam generator primary-to-Ejector Noble Gas turbine hall between MSR's. secondary leak. Hay be indicative of a potential airborne radiation exposure in turbine hall.

l 1(2)RE-216 Containment Fan In Unit 1, located west and slightly Provides indication of potential contamination Coolers Liquid Process south of C59 panel. In Unit 2, of cooling water.

located west and slightly north of C59

! panel. .

! l 1(2)RE-216B Containment Located next to 1(2)RE-216 monitor. Provides background data for the 1(2)RE-216 I Fan Coolers Background liquid process monitor, if used.

1(2)RE-217 Component In Unit 1, located in overhead Provides indication of component cooling Cooling Water Liquid just north of stairs going from water contamination. Shuts component Process El. 8' of aux. building to C59 cooling water surge tank vent.

panel area. In Unit 2, located in overhead just west of Unit 2 i component cooling water pumps.

f RE-218 Waste Disposal Located on east wall of waste con- Monitors waste condensate activity being dis-j System Liquid Process densate cubicle across from component charged. Upon exceeding high level setpoint, i

cooling water pump on El. 8' of aux. discharge of waste condensate is secured.

building.

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i Attachment 1.2-2, Cont'd, j Monitor Location Indication i RE-218B Waste Disposal Located next to RE-218 monitor. Provides background activity level for RE-218 system Background . , ,. liquid process monitor, if used.

I' 1(2)RE-219 Steam Generator Located outside each primary side Provides an indication of steam generator Blowdown Liquid Process sample room on El. 26' of aux. blowdown activity. Secures steam genrator i l building. blowdown and blowdown tank outlet valves, }

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and steam generator sample valves.

i i

1(2)RE-219B Steam Generator Located next to 1(2)RE-219 Provides background activity level for i Blowdown Background monitor. 1(2)RE-219 liquid process monitor, if used.-

j RE-220 Spent Fuel Pit Heat Located on El. 46' of aux. building Provides. an indication of service water con-

Exchanger Service Water on north wall just west of door to tamination from a spent fuel pit heat ,

Liquid Process Unit 2 containment facade, exchanger tube leak.

RE-220B Spent Fuel Pit Heat Located next tio RE-220 monitor. Provides background activity level for Exchanger Service Water RE-220 liquid process monitor, if used.

Background

RE-221 Drumming Area Vent Located in exhaust ducting above Indicates noble gas activity released from Stack Noble Gas drumming area SPING in northwest spent fuel pit and drumming area. May be corner of Unit 1 facade. indicative of a potential aux. building airborne release.

1(2)RE-222 Steam Generator Located on El. 26' of aux. building Provides an indication of activity level in }

Blowdown Tank outlet on east side of steam generator blow- blowdown tank. Secures blowdown of steam

Liquid Process down tank. generators and closes blowdown tank outlet I

valves.

!i RE-223 Waste Distillate Mounted on east side of "C" component Monitors activity of waste distillate during

Discharge Liquid Process cooling water heat exchanger on discharge. Secures discharge valves upon

] El. 46' of aux. building. exceeding setpoint.

RE-2238 Waste Distillate Located next to RE-223 monitor. Provides background activity level for RE-223

Dis 3 ti arge Background liquid process monitor, if used. !

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i Attachment 1.2-2, cont'd.

Monitor Location Indication l

l RE-224 Gas Stripper Vent Located in northeast corner of Indicates activity of gaseous release from

Stack Noble Gas Unit 2, El. 26' containment facade letdown gas stripper building.

i by exhaust duct.

RE-225 Combined Air Ejector Located above door on El. 46' of Indicative of primary-to-secondary leak in Low-Range Noble Gas turbine hall leading to I&c office. steam generatore.

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HE-226 Combined Air Ejector Located adjacent to RE-225 Ica-range Provides an indication of the noble gas

.; Low-Range Noble Gas noble gas monitor. activity in combined air ejector discharge l in the event RE-225 monitor is pegged off-scale high.

I 1(2)RE-229 Service Water For Unit 1, located on El. 8' of Monitors activity of service water discharge.

Discharge Process aux. building in vent area. For Unit 2, located in aux. feed pump room on east side of tunnel.

1(2)RE-229B Service Water Located adjacent to 1(2)RE-229 liquid Provides background activity level for process monitor. the 1(2)RE-229 monitor, if used.

Discharge Background RE-230 Retention Pond Located on El. 8' of turbine hall Monitors activity level in retention pond Discharge Liquid Process outside entrance to water treatmeat. effluent.

RE-230B Retention Pond Located adjacent to RE-230 liquid Provides background activity level for Discharge Background process monitor. RE-230 liquid monitor.

I 1(2)RE-231 - Line "A" Located on El. 88' of containment Monitors activity of steam released.

1(2)RE-232 - Line "B" facade in the area of atmospheric Steam Line Atmospheric relief valves - one per steam Release generator.

4 RE-234 Control Room Iodines Located on top of control room Honitors iodine activity in control room.

< building on El 46' of turbine hall. Initiates 100% recirculation (Mode 4) of control room ventilation upon exceeding 1 s setpoint.

1 .

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A /

N .

f s.

Attaclaient 1.2-2, Cont'd.

Monitor Location Indication ;

Provides background activity level for RE-234B Control Room Located adjacent to RE-234 iodine Background monitor. RE-234 iodine monitor.

REf235 Control Room Noble Located adjacent to RE-234 iodine Monitors noble gas activity in control room.

Gas monitor. Initiates 100% recirculation (Mode 4) of control room ventilation upon exceeding d- setpoint. ;

RE-237 Technical Support Located in ductwork on El. 18.5' Monitors iodine activity in the TSC air Center Iodines of TSC building, in northwest duct.

corner.

RE-237B Technical Support Located adjacent to RE-237 iodine Provides background activity for RE-237 Center Background monitor, monitor.

RE-238 Technical Support Located adjacent to RE-237 iodine Monitors the noble gas activity in TSC .

Center Noble Gas monitor, air duct.

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ATTACHMENT 1.2-3 .

i ,

LISTING OF HIGH-RANGE RMS-II PROCESS HONITORS BY PLANT LOCATION /

NOTE: THE RMS-II PROCESS MONITORS ARE HIGH-RANGE INSTRUMENTS CALIBRATED TO READOUT IN MREM / HOUR OR REM / HOUR.

l THEY ARE OFF-LINE TYPE MONITORS AND DO NOT DIRECTLY INDICATE VENT STACK DOSE RATE LEVELS. SHIELDING PREVENTS THEM FROM BEING USED AS AN INDICATION OF LOCAL DOSE RATES NOT ORIGINATING WITHIN THE PROCESS SYSTEM BEING MONITORED.

l Honitor Location Indication f s ,

RMS-II, Ch. 1/ Aux. Building Unit I facade, El. 66'. Provides a means of assessing high level Vent Stack releases from the aux. building. Any 1 indication on this monitor indicates a ;

j potential for hioh airborne activity and

high radiation dose rates within the aux. ,

building.

l RMS-II, Ch. 2/ Drumming Area Unit 1 facade, El. 6.5', west side. Provides a means of assessing high level Vent Stack releases from the central section (El. 46' l

& El. 66') of aux. building, spent fuel l

pit and drumming area. Any indication on this monitor indicates a potential for 1

high airborne activity and high radiation dose rates in the above areas. !

! RMS-II, Ch. 3 & 4/Contain- Unit 1: Unit I containment El. 66' Provides a means of assessing high level ment Purgc Stack entrance hatch platform. Unit 2: releases from containment. Any indication On top of Unit 2 R11/12 cubicle. on this monitor indicates a potential for high airborne activity and high radiation '

I dose rates in aux. building and in area of i purge fans and stacks.

RHS-II, Ch. 5/ Combined Air Above CR9 monitor on west; wall of Provides a ueans of assessing high level 1

Ejector Duct Unit I turbine hall, El. 46'. releases from the steam air ejector. Any indication on this monitor indicates a potential for high air ejecter airborne

activity or high radiation dose rates throughout the turbine building.

A

,9 3, 1 Attachment 1.2-3, Cont'd.

1 Monitor Location Indication l

RMS-II, Ch. 6/ Gas Stripper Unit 2, El. 668 fanroom, north side. Provides a means of assessing high level Building Vent Stack ,

releases from the gas stripper buildings.

Any indication on this monitor indicates a potential for high airborne activity within the gas stripper or high radiation dose rate i levels in the vicinity of the letdown gas strippers and related aux. building areas on El. 46' and El. 26'.

I i

V 3

. k

EPIP 1.3 MINOR Revision 6 07-01-83 ESTIMATION OF SOURCE TERM 1.0 GENERAL The purpose of this procedure is to estimate the source term (stack release rate in Ci/second) using the low range operational stack monitors, the Eberline RMS II Radiation Monitoring Systems or direct contact radiation measurements on the plant effluent vents. The plant effluent vent stacks are:

1.1 Auxiliary Building Vent, R14 (ABVNT) 1.2 Drumming Area Vent, R21 (DAVNT) 1.3 Gas Stripper Building Vent, RE-224 (GSBVNT) 1.4 Unit 1 Containment Purge, 1RE-305 l.5 Unit 2 Cdntainment Purge, 2RE-305 1.6 Combined Air Ejector Decay Duct, RE-225 (CAE) 1.7 Main Steam Safety Valves and Atmospheric Dump Valves, 1(2)RE-231 j and 1(2)RE-232 !

2.0 REFERENCE l

2.1 EDS Report to Wisconsin Electric Power Company concerning NUREG-0578, March 7, 1980.

3.0 PRECAUTIONS 3.1 If fuel damage or loss of reactor coolant system integrity has occurred, some or all of the following would be present: .

3.1.1 The letdown radiation monitor (1R9) or the failed fuel radiation monitor (2RE-109) may be unusually high or offscale.

3.1.2 The containment radiation monitors (1R11 and 1R12 or 2RE-211 and 2RE-212) may be unusually high or offscale.

3.1.3 The containment area monitors (1R2 and 1R7 or 2RE-102 and 2RE-107) may be unusually high or offscale.

3.1.4 The charging pump area monitor (IR4 or 2RE-104) may be .

t unusually high or offscale.

~7);

i l

l

b EPIP 1.3 Page 2 3.2 Health physics procedures and requirements must be followed when applicable (i.e., entering a high radiation area).

3.3 Evaluation of the radiation monitoring system readouts and radio-logical hazards must be completed prior to any attempt to enter the auxiliary building or facade to take a contact reading on any stack.

3.4 If this procedure is being used for determination of emergency classi-fication, use EPIP 1.8 " Emergency Off-Site Dose Estimations" for determination of projected dose off-site. EPIP 1.8 is a shorter, however more conservative procedure for determination of projected dose.

4.0 INITIAL CONDITIONS -

4.1 Applicable portions of EPIP 1.2, " Plant Status", is completed.

5.0 PROCEDURE FOR Xe-133 EQUIVALENT RELEASE RATE ESTIMATE - WORKSHEET NO. 1 5.1 Chemistry / Health Physics Supervisor or Designated Alternate 5.1.1 Obtain EPIP-05 and EPIP-06 of EPIP 1.2, " Plant Status," for sthe radiation monitoring systems.

(._ NOTE: IF EPIP-05 AND EPIP-06 IN EPIP 1.2, " PLANT STATUS," ARE NOT COMPLETED, OBTAIN THE METER READINGS FOR EACH PLANT EFFLUENT VENT STACK FROM THE REMOTE CONTROL ROOM READOUT AND RECORD THIS ON WORKSHEET NO. 1 AND THEN PROCEED WITH STEP 5.1.3.

NOTE: PLANT EFFLUENT VENT STACK MONITOR READINGS ARE ALSO AVAILABLE FROM THE TECHNICAL SUPPORT CENTER DATA LOGGER. SEE ATTACHMENTS 1.3-1, 2 & 3.

5.1.2 Enter the meter readings in the appropriate column on Work-sheet No. 1 for the indicated vents. If the readings are offscale, not monitored, or the monitors are inoperable, enter the appropriate word "offscale," "not monitored," or

" inoperable" in the meter reading column for the vent affected.

5.1.3 Designate individuals in accordance with ALARA concepts to f obtain meter readings of the vents whose Eberline RMS monitor .

readings are unavailable by performing Section 5.2 of this procedure, if required.

NOTE: IF STEP 5.1.3 NEEDS TO BE COMPLETED BECAUSE EBERLINE RMS MONITOR READINGS ARE UNAVAILABLE, THEN PERFORM i SECTION 5.3 0F THIS PROCEDURE AFTER APPROPRIATE

MEASUREMENTS HAVE BEEN TAKEN IN SECTION 5.2.

i ]

- _ _ _ _ . . __ _. 1 . . .- _

4 i

EPIP 1.3 Page 3 m .

5.1.4 Perform Section 5.3 of this procedure to determine the estimated Xe-133 equivalent release rate.

5.2 Direct Stack Survey Team Designees NOTE: THE FOLLOWING SECTION WILL NOT BE INITIATED UNTIL THE EVALUATION DISCUSSED IN PRECAUTION 3.3 M BEEN COMPLETED AND THE SITE MANAGER (DUTY & CALL SUPERINTENDENT), THE DUTY &

CALL HEALTH PHYSICS SUPERVISOR, AND THE DUTY SHIFT SUPERVISOR HAVE APPROVED INITIATION. THIS SECTION WILL BE ACCOMPLISHED UNDER THE DIRECTION OF HEALTH PHYSICS SUPERVISION.

5.2.1 Determine the most direct and desirable route to the plant effluent stack to be monitored.

5.2.2 Determine the Health Physics requirements to be met for the passage to the vent areas.

5.2.3 Determine the appropriate survey instrument to be used for the plant effluent vent to be monitored.

- 5.2.4 Proceed by the route determined in Step 5.2.1 to the stack and record the survey instrument reading in contact with the

( .

stack in the columns provided on Worksheet No.1, Part C, Plant Effluent Vent Stack Contact Readings.

NOTE: IN THE CASE OF THE MAIN STEAM SAFETY VALVES AND ATMOSPHERIC STEAM DUMP VALVES, THE READING WILL BE TAKEN IN CONTACT WITH THE CENTERLINE OF THE MAIN STEAM HEADER, THREE FEET FROM THE MAIN STEAM LINE.

SHIELD THE PROBE (WITH A MINIMUM OF .25 INCHES OF .

LEAD) ON THE SIDES FACING THE MAIN STEAM LINE AND THE CONTAINMENT.

5.3 Chemistrv/ Health Physics Supervisor or Designated Alternate 5.3.1 Choose the appropriate vent stack readouts in Part A, B, or

C of Worksheet No. I to convert readings to a Xe-133 equiva-l lent release rate. That is if the low. range monitors go l offscale, use the high range monitors. Conversely, if the normal monitors are onscale, use the normal monitors, or if both normal and high range monitors are offscale or inoper-able, use the vent stack contact readings.

5.3.2 Use the appropriate conversion factors for each of the plant effluent vent to convert the chosen vent stack readout, (cpm, pCi/cc or R/ hour) from Step 5.3.1 to an Xe-133 equivalent release rate in curies /second and record the value on Worksheet i No. 1, Part F, Estimate of Gross Xe-133 Equivalent Release .

l Rate. Enter the appropriate word "cffscale," "not monitored," - ::g or " inoperable" for the cases where the plant effluent vent *1 was not monitored, offscale, or inoperable.

NOTE: THE FOLLOWING QUALIFYING NOTES MUST BE RECOGNIZED.

EPIP 1.3 Page 4 7-s

1. If the actual flow rate is different than the assumed conversion factor flow rate, a ratio of:

Actual Flow Rate Assumed Flow Rate should be applied to determine the release rate.

(Ratio) x Release Rate Value = Adjusted Xe-133 Release Rate

2. Determine the steam line atmospheric vent, or the main steam header vent release rate in accordance with the following:

a. Obtain from the Shift Supervisor an estimated flow rate through the main steam header in lbm/ hour of steam being dumped to the environ-ment and the specific volume (v) of the steam.

NOTE: AT 1000 PSIA,' SPECIFIC VOLUME IS 0.446 FT3 /LBM. AT 500 PSIA, SPECIFIC t VOLUME IS 0.928 FT3/LBM.

C" lbm/hrxvfb; x 7.86 f

3 y,

b. Convert contact reading, if required, the main steam header to pCi/cc using the appropriate conversion factor (Worksheet No. 1 Sect. C).

pCi/cc

c. Multiply flow rate obtained in Step (a) by the concentration obtained in Step (b) to obtain the release rate (Xe-133 equivalent) from the main steam header.

Flow Rate Concentration _ Main Steam Header

~

(cc/sec.) * (pCi/cc) Release Rate 5.3.2 Sum the values (1) through (5) on Worksheet No.1, Part F, or use grab sample results #7 on Worksheet No. 1, Part F, to determine the gross Xe-133 equivalent release rate.

b_

EPIP 1.3 Page 5 NOTE: IF GRAB SAMPLE RESULTS ARE AVAILABLE, THE RESULT OF

, SUCH SAMPLES SHOULD BE MORE ACCURATE THAN GROSS MONITOR READINGS AND HENCE SHOULD BE USED IN LIEU OF THE RELEASE RATES CALCULATED ABOVE OR T.N ADDITION TO THE ABOVE IF THE RELEASE IS FROM AN UNMONITORED RELEASE PATH.

5.3.3 Report the calculated gross Xe-133 equivalent release rate to the Shift Supervisor and the Technical Support Manager.

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WORKSHEET NO. 1 Xe-133 EQUIVALENT RELEASE RATE 1

A. OPERATIONAL LOW-RANGE RELEASE MONITORS READOUTS Xe-133 Equiv.l Flow Rate Conversion Release Rate Meter Reading (cfm) Factor Ci/sec ,

9 Auxiliary Building Vent R/hr 61400 5.8 x 10 !

8 Drumming Area Vent R/hr 43100 1.3 x 10 :

U1 Containment Purge, 1 fan epm 12500 2.1 x 10 ~8 i 2 fans epm 25000 4.2 x 10 ~8 !

~ '

U2 Containment Purge, 1 fan epm 12500 2.1 x 10 6 2 fans epm 25000 4.2 x 10 ~8 Gas Stripper Building Vent pCi/cc 13000 6.2 ,

2 ,

Combined Air Ejector pCi/cc 25 1.6 x 10 l Steam Line Atmospheric Vent pCi/cc Refer to Section "E" B. EBERLINE RMS - II VENT STACK READOUTS !

i

( NOTE: THESE READINGS ARE ALSO AVAILABLE ON THE TECHNICAL SUPPORT CENTER !

l DATA LOGGER. ATTACHMENTS 1.3-1, 2, 3. j Xe-133 Equiv.

f Flow Rate Conversion Release Rate

Meter Reading (R/ hour) (cfm) Factor. Ci/sec 3

Aux. Building Vent, Ch. #1 61400 3.0 x 10 l Drumming Area Vent, Ch. #2 43100 2.1 x 103 l

U2 Containment Purge -

12500 1.6 x 10 4 l 25000 3.2 x 104 U1 Containment Purge 12500 1.6 x 10 4 25000 3.2 x 10 4 GS Bldg. Vent, Chg #6 4 13000 6.2 x 10 'Sh Combined Air Ejector Decay, Ch. #5 25 3.4

u s

[

4 C. PLANT EFFLUENT VENT STACK CONTACT READINGS Xe-133 Equiv.

Meter Reading Flow Rate Conversion Release Rate Vent (R/hr) (cfm) Factor Ci/sec Auxiliary Building 61400 300 2

Drumming Area 43100 2.3 x 10 U1 Containment Purge 12500 8.0 x 101 25000 1.6 x 102 U2 Containment Purge 12500 8.0 x 101 25000 1.6 x 102 4

Gas Stripper Building 13000 S x 10 2

Combined Air Ejector Decay 25 1.6 x 10 Main Steam Header Refer to Section "E" D. ACTUAL VERSUS CONVERSION CURVE FLOW RATE RATIO Actual Flow Rate x Release Rate Value = Adjusted Releast Rate Assumed Flow Rate I

( )x =

l l E. STEAM HEADER XE-133 EQUIVALENT RELEASE RATE CALCULATION

1. lbm/hr x specific volume, ft 3 /lbm x 7.86 cc-hr/ft 3-sec NOTE: At 1000 psia specific volume = .446 ft33/lba At 500 psia specific volume = .928 ft /lbm lbm/hr x ft3 /lbm x 7.86 cc-hr/ft 3-sec

= CC/seC

2. Contact reading from Section C (if contact reading used):

R/hr = pCi/cc .

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3. Steam header release rate:

Flow rate cc/sec x Concentration pCi/cc = Release Rate 6

cc/sec x x 10 Ci/cc = Ci/sec Step E.1 Step A.7 or E.2 F. ESTIMATE OF GROSS Xe-133 EQUIVALENT RELEASE RATE Xe-133 Equivalent Release Rate Vent (Curies /Sec.)

1. Auxiliary Building

2. Druauning Area

3. Gas Stripper Building

4. Combined Air Ejector Decay Duct

5. Main Steam Header

6. Unit 1 Containment Purge

7. Unit 2 Containment Purge s' 8. Sun (Gross Xe-133

) Equiv. Release Rate)

I SE

9. Grab Sample Results = Ci/sec.

l Completed By Time Date 4

ATTACHMENT 1.3-1 l'

DRCCCMG AREA VENT STACK Td4S-II CH =2 &

COMBINED AIR EJECTCR DISCHARGE F1 S-II CH =5 t

VOLTAGE TO R/HR CONVERSION TABLE RANGE 1 to 10" R/HR

'.*o l t s Units R/hr 0 0.001 0.1 0.00135 0.2 0.001847 0.3 0.002511 0.4 0.003414 0.5 . 0.004641 0.6 0.006309 0.7 0.008576 0.8 0.011659 0.9 0.015848 1.0 0.021544 1 1.1 0.029286

- 1.2 0.039810 1.3 0.054116 1.4 0.073564 1.5 0.100 1.6 0.135935 1.7 0.184784 1.8 0.251188 1.9 0.341454 2.0 0.464158 2.1 0.630957 2.2 0.857695 2.3 1.165914 2.4 1.584893 2.5 2.154434 2.6 2.925644 2.7 3.991071 2.8 5.411695 2.9 7.356422 3.0 10.0 i

. ATTACHMENT 1.3-2 6 '

s' UNIT 1 RMS-II CH #3 & UNIT 2 RMS-II CH #4 CONTAINMENT PURGE STACKS VOLTAGE M R/NR CONVERSION TABLES RANGE 10-' M 10

R/HR volts Units - R/RR

0. 0.1 0.1 .135 0.2 .184 0.3 .251 0.4 .341 0.5 .464 0.6 .630 0.7 .857

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0.8 1.165 O.9 1.584

1. 2.154 1.1 2.928 1.2 3.981 1.3 5.411 1.4 7.356 j

1.5 10.

l.6 13.593 1.7 18.478 1.8 25.118 1.9 34.145 2.

46.415 2.1 63.095 2.2 85.769 2.3 116.591 2.4 158.489 i 2.5 215.443 2.6 292.864 2.7 398.107 l

2.8 541.169 1 2.9 735.642

3. 1000.

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. ATTACHMENT 1.3-3

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AUXILIARY BUILDING VENT STACK RMS-II CH #1 &

GAS STRIPPER BUILDING VENT STACK RMS-II CH #2 VOLTAGE TO R/HR CONVERSION TABLE RANGE 10 TO 10' R/HR Volts Units - R/HR

0. 0.01 0.1 .013 0.2 .018 0.3 .025 0.4 .034 0.5 .046 0.6 .063 0.7 .085 0.8 .116 0.9 .158

1. .215 1.1 .292

. ._ 1 1.2 .398 7 [~ , 1.3 .541 1.4 .735 1.5 1.

1.6 1.35 1.7 1.847 1.8 2.511 1.9 3.'414

2. 4.641 2.1 6.309 2.2 8.576 2.3 11.659 2.4 15.848 2.5 21.544 2.6 29.286 2,~ 39.810

}1 54.116 l

>9 73.564 j 3. 100.

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EPIP 1.7 MINOR h,.- Revision 2 07-01-83 EVALUATION OF CORE DAMAGE 1.0 PURPOSE The purpose of this procedure is to estimate core damage using a mathema-tical model based on an actual primary coolant sample activity, estimated volume introduced into the primary system through safety injection and a correction factor based on the time since reactor shutdown. This evaluation should be performed by the Core Physics Coordinator or a Duty Technical Advisors and routed to the' Technical Support Manager and Site Manager.

2.0 REFERENCE Calculations performed by the Nuclear Engineering Section of Wisconsin Electric Power Company documented in a report to G. A. Reed dated October 5, 1981 "C & HP Items Related to NUREG-0737."

3.0 PRECAUTIONS 3.1 If fuel damage or loss of reactor coolant system integrity has occurred, some or all of the following would be present:

~ ~

3.1.1 The 1etdown radiation monitor (1R9) or the failed fuel radiation monitor (2RE-109) may be unusually high or offscale.

3.1.2 The containment radiation monitors (1Rll & 1R12 or 2RE-211 &

2RE-212) may be unusually high or offscale.

3.1.3 The containment area monitors (1R2 & 1R7 or 2RE-102 & 2RE-107) may be unusually high or offscale.

3.2 Health Physics procedures and requirements must be followed when applicable (e.g., when entering a high radiation area).

3.3 Evaluation of the radiation monitoring system readouts and radiologica'l hazards must be completed prior to any attempt to enter the auxiliary building to take a primary sample.

4.0 INITIAL CONDITIONS 4.1 Applicable portions of EPIP 1.2, " Plant Status," are completed.

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EPIP 1.7 Page 2 4.2 A reactor coolant sample has been taken and a contact reading of the sample bomb has been taken or a final total sample activity has been completed by implementing EPIP 7.3.2 " Post-Accident Sampling &

Analysis of Potentially High Level Reactor Coolant."

4.3 A contact reading of the sample bomb in R/hr was taken and listed on form EPIP-30 or an actual sample activity has been received from lab analysis.

5.0 PRIMARY COOLANT SAMPLE ACTIVITY ESTIMATE PROCEDURE 5.1 Note the time of the sample contact reading taken in Section 4.3 on form EPIP-33.

5.2 Determine the amount of time since reactor shutdown to sample contact reading using the equation:

Reactor Shutdown Time - Contact Reading Time = Time Since Shutdown 5.3 convert the R/hr reading obtained using the teletector to Ci/ml using Attachment 1.7-1.

5.4 Enter the-conversion factor from Section 5.3 on form EPIP-33.

5.5 Determin the estimated Sample Activity using the equation:

Estimated Sample Act2,ity (Ci/ml) =

Ci/ml Sample Bomb Contact Reading * (R/hr) x Conversion Factor R/hr

Contact reading is on shielded sample bomb which incorporates 3 inches of external solid lead shielding.

5.6 Enter the estimated Sample Activity on form EPIP-33.

1 6.0 EXAMPLE Coolant Sample Activity Estimate (Shielded Bomb)

Teletector reading = 2.75 R/hr Reading time = 1700 Reactor Shutdown Time = 0900 Time since shutdown: 1700 hours0.0197 days 0.472 hours 0.00281 weeks 6.4685e-4 months - 0900 hours0.0104 days 0.25 hours 0.00149 weeks 3.4245e-4 months = 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months 2.75 R/hr x 5.93 x 10 = 1.63 x 10- Ci/ml R/

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. i EPIP 1.7 l Page 3 7.0 CORE DAMAGE ESTIMATE PROCEDURE 7.1 Calculate the estimated percentage of core damage using the following formula and table of correction factors. Interpolate correction fac-tors for times between those listed.- Use best estimate for safety injection volume.

7.1.1 Estimated Sample Activity (ESA) Ci/ml 7.1.2 Estimated Safety Injection Volume (ESIV) gallons Available safety injection dilution sources are:

Accumulators: 2 at 1,000 gallons each Refueling water storage tank: 275,000 gallons Boric acid storage tank: 1 of 3 at 5,000 gallons each i

Spray additive tank: 2,574 gallons 7.1.3 Correction Factor for Time Since Shutdown [CF(t)] hours 7.1.4 Enter the values from Sections 7.1.1, 7.1.2, and 7.1.3 on form EPIP-33. Calculate the percent core damage using the following formula and Attachment 1.7-2 and enter the result on form EPIP-33.

( + }

Percent Core Damage (%) = '

CF(t) 7.1.5 Route form EPIP-33 to the Site Manager and Technical l Support Manager.

8.0 EXAMPLE Percentage Core Damage Estimate 1

Estimated Sample Activity = 1.63 x 10 Ci/ml Time since shutdown = 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months l

Estimated safety injection volume = 2,000 gallons

_1 Percentage core damage (%) = 1.63 x 10 Ci/ml x (32,500 + 2,000) 547 Percentage core damage (%) = 9%

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b EPIP 1.8 MINOR

/ . Revision 3 07-01-83 EMERGENCY OFF-SITE DOSE ESTIMATIONS 1.0 GENERAL The purpose of this procedure is to provide a method for the expeditious classification of an accident or event based on estimated off-site doses.

The procedure provides a methodology to quickly estimate (1) stack release rates (source terms) and (2) off-site whole body and thyroid doses.

2.0 REFERENCES

2.1 U. S. NRC Regulatory Guide 1.109, Calculation of Annual Dose to Man froa Routine Releases of Reactor Effluents for the Purpose of Evalu-ating Compliance with 10 CFR 50 Appendix I, Revision 1, October,1977 2.2 U. S. EPA " Manual of Protective Action Guides and Protective Actions for Nuclear Incidents," EPA-520/1-75-001, September 1975, Appendix D 2.3 TID 14844 " Calculation of Distance Factors for Power and Test Reactor Sites," March 23, 1982

( 2.4 EDS Report to Wisconsin Electric Power Company concerning NUREG-0578, e

March 7, 1980.

2.5 Point Beach Nuclear Plant, Final Safety Analysis' Report (FSAR) 3.0 PRECAUTIONS & LIMITATIONS 3.1 This procedure is intended for use in the control room by a person designated by the Shift Supervisor such as the Duty Technical Advisor.

l 3.2 This procedure is to be used only for humediate initial dose pro-jections. The initial dose projections are to be refined using EPIP 1.3 and 1.4 once more data becomes available, i.e., meteoro-logical data, air samples, and off-site survey dose measurements.

3.3 It is recognized that the RadCon/ Waste Manager, in conjunction with the Chemistry & Health Physics Supervisor, is ultimately responsible for off-site dose assessments. However, the initial projections will normally be done by the Shift Supervisor or designee for purposes of classifying the event.

}

EPIP 1.8 Page 2 4.0 INITIAL CONDITIONS 4.1 An emergency or potential condition which is anticipated to have off-site dose consequences.

4.2 A release of airborne radioactivity has occurred, or a release is anticipated, requiring a conservative estimate of the off-site dose consequences.

5.0 PROCEDURE 5.1 Calculation of Xe-133 Equivalent Release Rates (Source Terms) 5.1.1 Airborne effluents may be discharged from PBNP through the following vent stacks:

a. Auxiliary building vent (ABVNT)

b. Drumming area vent (DAVNT)

c. Unit 1 containment purge vent (Cont.1)

d. Unit 2 containment purge vent (Cont. 2)

e. Gas stripper building vent (GSBVNT)

f. Combined air ejector decay duct (CAE)

g. Main steam safety valves and atmospheric dump valves 5.1.2 The source terms (vent release rate in ci/second) may be estimated by using any of the following monitoring systems.

a. Low range operational stack monitors (designed to monitor low-level releases).

b. Eberline RMS II radiation monitoring system (designed to monitor high-level releases).

c. Contact readings using a hand-held survey meter (to be l used when other monitor systems are non-operable).

5.1.3 The decision as to which monitoring system is to be used to estimate the source terms is dependent on the level of j release and the operability of the monitor.

5.1.4 Meter readings are to be entered in the appropriate column on EPIP-34. If meter readings are "off-scale" or " inoperable,"

enter the appropriate comment in the meter reading column on EPIP-34. A source term estimate must be made for each vent which is exhibiting readings above normal operating readings. .

. EPIP 1.8 Page 3 i

5.1.5 Direct contact readings using a h'nd-held a survey meter are required when meter readings from the low range monitoring system or the RMS II system are not available.

5.1.6 Direct contact readings using a hand-held survey meter are not to be initiated until the following conditions are accomplished:

a. An evaluation of the radiological hazards must be com-pleted prior to any attempt to enter the auxiliary building or facade ~ to take survey' readings on any stack or vent.

b. Before the surveys are done, the proper survey meter and the most direct and desirable route to the stack to be monitored must be chosen.

c. The surveys will be accomplished under the direction of the Health Physics Supervisor. The surveys must ce approved by the Site Manager, Duty & Call Health Physics Supervisor, and the Duty Shift Supervisor.

5.1.7 For surveying the main steam safety valves and atmospheric dump valves, the reading will be taken in contact with the

(' s centerline of the main steam header, three feet from the a

main steam line. The survey probe is to be s'ielded with a minimum of 0.25 inches of lead on the side of the probe facing the main steam line and the containment.

5.1.8 The following data should be obtained from the Shift Super-visor in order to estimate release rates from the main steam safety valves and atmospheric dump valves:

a. Estimated flow rate of steam through the main steam l

header in lbs/hr.

b. Specific volume of the steam in ft 3/lb. At 1000 psia, specific volume of 0.446 ft3 /lb. At 500 psia, specific volume is O'.928 ft3 /lb.

Enter this data on the appropriate column in Section 3.0 of EPIP-34.

5.1.9 Sum the values on EPIP-34, Section 4.0, to determine the gross Xe-133 equivalent release rate.

i e

EPIP 1.8 Page 4 (m

5.2 Whole Body Dose Projections 5.2.1 Off-site whole body dose rates may be calculated at the site boundary using the following equation:

D(REM) = X/Q x Q x Kr HR.

Where:

D = whole body dose rate (Rem /hr)

X/Q = atmospheric dispersion coefficient (sec/m 3)

Q = gross Xe-133 equivalent release rate (Ci/sec.)

Kr = Dose Factor (rem-m3 )

Ci-hrs Projected off-site dose rates may be calculated by entering the total Xe-133 equivalent release rate calculated on EPIP-34 in the appropriate column on EPIP-35 and multiplying the variables in the equations.

5.2.3 Enter the estimated release duration in the appropriate

column on EPIP-35. (A dose per hour is calculated by entering an exposure period of one (1) hour.)

NOTE: THE X/Q VALUES LISTED ON EPIP-35 ARE ESTIMATED BASED ON CALCULATED ACCIDENT METER 0 LOGY FOR 0-2 HRS.

AS GIVEN IN THE FSAR. IF REAL TIME METEOROLOGICAL DATA IS AVAILABLE, X/Q VALUES CAN BE CALCULATED AS OUTLINED IN EPIP 1.4 SECTION 5.1. REFINEMENT OF THE PROJECTED OFF-SITE DOSES MAY BE ACCOMPLISHED BY SUBSTITUTING THE REAL TIME X/Q CALCULATED i VALUE FOR THE ESTIMATED X/Q ON EPIP-35.

5.3 Thyroid Dose Projection l

( 5.3.1 Calculate the projected thyroid dose at the site boundary on Section 2.0 of EPIP-35 by using the following equation:

Thyroid Dose = Whole Body Dose x Conversion Factor

! 5.3.2 The conversion factor is dependent on the type of accident which has occurred. Conversion factors are tabulated for the following accidents:

a. Loss of coolant accident (LOCA)

I

b. Gap activity accident

c. Fuel. handling accident

.- x

. 3 g

d. Steam generator tube rupture

EPIP 1.8 Page 5 7

5.3.3 Choose the apprcpriate type accident and calculate the thyroid dose in Section 2.0 of EPIP-35 by multiplying the whole body dose calculated in Sectin 1.0 by the conversion factor.

NOTE: IF THE TYPE OF ACCIDENT IS NOT KNOWN, USE THE LOSS OF COOLANT CONVERSION FACTOR.

5.4 Classification of the Event Based on Estimated off-Site Doses 5.4.1 The event is to be classified as a Site Emergency if the projected off-site doses meet any of the following criteria:

~

a. Effluent monitors detect levels corresponding to any of the following doses at or beyond the site boundary:

(1) >50 mR/hr whole body for \ hour (2) >250 mR/hr for \ hour for the thyroid (3) >500 mR/hr whole body for 2 minutes (4) >2500 mR/hr to the thyroid for 2 minutes s

b. Any of the above dose rates are projected, based on plant parameters.

5.4.2 The event is to be classified as a General Emergency if the projected off-site doses meet any of the following criteria:

a. Effluent monitors detect levels corresponding to any of the following doses at or beyond the site boundary:

(1) 21 R/hr whole body (2) 25 R/hr thyroid

b. Either of above dose rates are projected based on plant parameters.

5.5 Protective Action Recommendation 5.5.1 Due to the conservative nature of the calculations in this procedure, use EPIP 1.4 to calculate dose projections for protective action recommendations unless it is apparent there is no time to use EPIP 1.4.

5.5.2 Enter dose rates and release duration estimate on EPIP-35.

If duration of release is unknown, use a release duration of 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months .

M4 a

.=r. - ~ ,

F EPIP 1.8 Page 6 5.5.3 Recommend one of the following offsite protective actions based on results of EPIP-35 calculations.

Whole Body Dose

a. <1 R - No planned protective actions.

b. 1 to <5 R - Seek shelter as a minimum. Consider evacu-ation.

c. >5 R - Conduct evacuation in area affected and public within 2 miles of the plant.

Thyroid Dose

a. <5 R - No planned protective actions.

b. 5 to <25 R - Seek shelter as a minimum. Consider evacu-ation.

c. >25 R - Conduct evacuation in area affected and public within 2 miles of the plant.

s e

ne %

EPIP 6.5 '

MINOR

-' Revision 2 f-- 07-01-83 TECHNICAL SUPPORT CENTER & OPERATIONS SUPPORT CENTER ACTIVATION 1.0 PURPOSE 1.1 To provide instructions for the activation of the technical support center after the declaration of an alert, site emergency or general emergency.

1.2 To outline the technical support center ventilation system operation in the event of high airborne activity.

1.3 operation of the technical support center emergency power supply is also a part of this procedure.

2.0 ACTIVATION OF TECHNICAL SUPPORT CENTER 2.1 Se t up 10 tables as shown on Attachment 6.5-1. They may be obtained from the rooms in the adjacent operations support center and health physics areas of the technical support center.

2.2 Install phones from the technical support center storage cabinets in the appropriate areas as shown on Attachment 6.5-1.

2.3 Confirm that all phones operate by noting if dial tone is present when the receiver i~s lifted. Confirm that the numbers listed on the face of the phone matches the number on the wall jacks.

2.4 Distribute paper and pencils to each table.

2.5 Distribute a copy of the EPIP's to the table in front of the black-board and the table near the dose plotting map (see Attachment 6.5-1).

2.6 Obtain a copy of each of the following manuals from the front office l

area and bring them to the technical support center. They may be l

obtained from the office of the Manager, General Superintendent, or the training offices.

2.6.1 Operating procedures.

2.6.2 Emergency operating procedures.

l 2.7 Shift the ventilation system from the normal to the emergency opera-ting mode by implementing Section 4.0.

l l 2.8 Date and time the current charts on the safety parameter chart re-l corders in the technical support center.

i

l 1

}

EPIP 6.5

. _ Page 2 f .

. 2.9 Dispatch a person with systems training (Duty Technical Advisor) to control room to aid in communications between the TSC and control room.

3.0 DATA LOGGER OPERATION 3.1 Push the start button on the data logger. Enter the correct date and time on the printout from the data logger. The designations for each of the 37 channels are contained in Attachment 6.5-2.

3.2 The conversion of incore thermocouple NV to degrees Fahrenheit with the reference junction at.160* is accomplished by use of the incore thermocouple table (see Attachment 6.5-3) .

3.3 The conversion of radiological monitoring point volts to mR/hr or R/hr is accomplished by use of the conversion tables (see Attachments 6.5-4, 6.5-5 and 6.5-6).

4.0 EMERGENCY VENTILATION SYSTEM 4.1 The technical support center heating and ventilating system has a normal and emergency operating mode. Under normal operation the air intake is from the outside air vent on the east wall of the technical

. {' support center building. The intake air under normal operation is essentiaplyunfiltered.

In the emergency mode, there are two optional air intake locations.

One is adjacent to the normal intake on the east wall of the techni-cal support center building and the other is on the north wall of the Unit 2 turbine hall 1.

4.2 To shift the heating, ventilating and air conditioning system from the normal to emergency mode:

1 4.2.1 Turn the auto /off/ occupied switch on Panel M-1 to the occu-pied position. See Attachment 6.5-7 for location of Panel M-1.

4.2.2 Turn the normal / emergency control switch on Panel M-1 to the emergency position.

4.2.3 Select the north or south (east) emergency intake depending on meteorological conditions. Select the upwind intake duct.

5.0 TECHNICAL SUPPORT CENTER AUXILIARY AIR CONDITIONING 5.1 Turn on compressor CH2 on the west wall of El.18.5' of the technical support center building (see Attachment 6.5-7).

5.2 Turn on air handling unit AHU2 on the technical support center building room north wall (see Attachment 6.5-1). ,

t.

7$

i

EPIP 6.5 Page 3 t ,-

6.0 EMERGENCY POWER 6.1 Tne power source for the technical support center in 1801 480 V. See Attachment 6.5-8 for the main disconnect location.

6.2 The emergency power source is not operational at this time.

t i

l am 9 i

l

j ATTACHMENT 6.5 .

'./

k ,g

. s.

TSC/HP ,

Blackroarc Air J//d-2

Conditioner E

a .

3 $ Table m g Site Mgr.

g y .

Plant Cperations ~

E Table Manager E,

$ o o eo

> -

oM m cr.

j g mg u "

S "

Table $"

aaral m/cc8he" Sa fety

?arameter Display Tech. E ,

S to. co rt Mgr.

Red Phene

(,_I, S torage Cabinet

' l N#AS Table Table Offsite -

y Co::=tunicator t

CSC Core Coord.

Data Proc. Coord. Table Chem /HP W 3 "P # ~

Table RQ Environmental Dose Plot 3 Status Map lC

!D TECHNICAL SUPPCRT CENTER

ArrACHMENT 6.5-2 Data Loccer Point Designations Channel Unit Status Parameter

~ 0 mV Active 3) )

1 mV Active 12) Unit 1 Reference )

2 mv Active 13) RTD ) n t 1 Incore 3 mV Active 18) Junction Box "A" ) Thermoccuples 4 mV Active 22) )

5 mV Active 26) Unit 1 Reference )

6 mV Active 29) RTD )

7 mV Active 36) Junction Box "B" )

8 mV Active 3) )

9 mV Active 12) Unit 2 Reference )

10 mV Active 13) RTD ) nit 2 Incore 11 mV Active 18) Junction Hox "A" )

.hermoccuples g-- 12 mv Active 22) )

13 mV Active 26) Unit 2 Reference )

((- 14 15 mV mV s Active Active 29) 36)

RTD Junction Box "B"

)

16 *F Active Unit 1 Reference RTD Junction Box "A" 17 *F Active Unit 1 Reference RTD Junction Sox "B" 18 *F Active Unit 2 Reference RTD Junction Box "A" 19 *F Active Unit 2 Reference RTD Junction Box "B" 20 gpm Active Unit 1 Auxiliary Feed Flow "A" SG 21 gpm Active Unit 1 Auxiliary Feed Flow "B" SG 22 gpm Active Unit i SI Flow Train "A"

, 23 gpm Active -

Unit 1 SI Flow Train "B" 24 ft Unit 1 Containment Sump Level 26- V Active Unit 1 Containment Purge Stack fes-II ch 43 27 gpm Unit 2 Auxiliary Feed Flow "A" SG 28 gpm Unit 2 Auxiliary Feed Flow "B" SG l 29 gpm Active Unit 2 SI Flow Train "A" l 30 gpm Active Unit 2 SI Flcw Train "B" 31 ft Unit 2 Containment Sump Level ,

33 V Active Unit 2 Containment Purge Stack RMs-II ch 44 34 V Active Drumming Area Vent Stack RMS-II Ch #2 35 V Active combined Area Vent Stack RMS-II Ch 45 l 36 V Active Gas Stripper Building Vent Stack RMS-II ch 46 l 37 V Active Auxiliary Building vent Stack res-II ch 41 e

e

<r= N 'N I01HI 16. . . i . . . HilCl. EAR PLANI - INCORE T/C TABLE - RL -.. dCE JUNCTION AT 160. DEG F MV DEG F MV DED F HV DEG F MV DEG F MV DEG F MV liEti F 0.00 lb2. 7.60 497. 15.20 024. 22.00 1146. 30.40 1472. 30.00 1811.

0.20 161. 7 . t10 505. 15.40 033. 23.00 1155, 30.60 1481. 30.20 1821.

0 40 170. 0.00 514. 15.JO 041. 23.20 1163. 30.00 1490. 30.40 1029.

0.60 100. O.20 523. 15. Os 850. 23.40 1171. 31.00 1498. 30.60 1039.

0.00 10Y. 0.40 532. 16.00 058. 23.60 1180. 31.20 1507, 30.00 1040.

1.00 1Y0. 0.60 540. 16.20 067. 23.80 1100. 31.40 1516. 39.00 1057.

1.20 200. 0.00 549. 16.40 075. 24.00 1197. 31.60 1524. 39.20 1067.

1.40 217. 9.00 550. 16.60 003. 24.20 1205. 31.80 1533. 39.40 1876. .

227. 9.20 567. 16.80 092. 24.40 1214. 32.00 1542. 39.60 1005.

1.'l60 39.00 1894.

1.00 236. 9.40 576. 17.00 901. 24.60 1223. 32.20 1551.

2.00 245. 9.60 504. 17.20 909. 24.80 1231. 32.40 1560. 40.00 1904.

2.'20 254. 9.00 593. 17.40 918, 25.00 1240. 32.60 1560. 40.20 1913.

2.40 263. 10.00 602. 17.60 926. 25.20 1240. 32.00 1577. 40.40 1922, 2.60 272. 10.20 610. 17.00 935. 25.40 1256. 33.00 1506. 40.60 1932.

2.00 202. 10.40 610. 10.00 943. 25.60 1265. 33.20 1595. 40.00 1941.

3.00 2Y0. 10.60 627. 18.20 951. 25.80 1274. 33.40 1604. 41.00 1950.

3.20 300. 10.00 636. 10.40 960. 26.00 1202. 33.60 1613. 41.20 1960.

3.40 309. 11.00 445. 18.60 968. 26.20 1291. 33.00 1621. 41.40 1969.

3.60 310. 11.20 653. 10.00 977. 26.40 1299. 34.00 1631. 41.60 1979.

3.00 327. 11.40 662. 19.00 985. 26.60 1300. 34.20 1639. 41.80 1908.

4.00 336. 11.60 671. 19.20' 993. 26.80 1316, 34.40 1649, 42.00 1998.

4.20 345. 11.00 679. 19.40 1002. 27.00 1325. 34.60 1657. 42.20 2008.

4.40 354. 12.00 600. 19.60 1010. 27.20 1334, 34.80 1666. 42.40 2017.

4.60 363. 12.20 696. 19.00 1019. 27.40 1342. 35.00 1675. 42.60 2026.

4.110 372. 12.40 705. 20.00 1028. 27.60 1351. 35.20 1684. 42.00 2036.

5.00 301. 12.60 714. 20.20 1036. 27.00 1359, 35.40 1493. 43.00 2045. p 5.20 390. 12.00 722.- 20.40 1045. 20.00 1360. 35.60 1702. 43.20 2055. Q 5.40 399. 13.00 730. 20.60 1053. 20.20 1377. 35.00 1711. 43.40 2065. M 5.60 4011. 13.20 739, 20.80 1061. 28.40 1305. 36.00 1720. 43.60 2074..

5.00 417. 13.40 747, 21.00 1070. 28.60 1394. 36.20 1729, 43.80 2004.

6.00 426. 13.60 756. 21.20 1078. 20.00 1403. 36.40 1738. 44.00 2094.

6.20 434. 13.00 765. 21.40 1007. 29.00 1411. 36.60 1740, 44.20 2103. P 4.40 444. 14.00 773. 21.60 1095. 29 20 1420. 36.00 1756. 44.40 2113. T 4.60 452. 14.20 702. 21.00 1103. 29.40 1429. 37.00 1766. 44.60 2123.

6 . 11 0 462. 14.40 790. 22.00 1112. 29.60 1437. 37.20 1775. 44.80 2133.

7.00 470. 14.60 799. 22.20 1120. 29.00 1446, 37.40 1784. 45.00 2142.

7.20 479. 14.00 007. 22.40 1129. 30.00 1455. 37.60 1793. 45.20 2152.

7.40, 400. 15.00 016. 22.60 1138. 30.20 1463. 37.80 1802, 45.40 2162.

1 Jt n An #"t Till'It A c. 4 0 Mll PFFFRF~NPC lilH P T T O M AT 140, DFG F

' ~I ATTACHMENT 6.5-4 )

l e

DRUMMING AREA VENT STACK RMS-II CH #2 &

COMBINED AIR EJECTOR DISCHARGE RMS-II CH #5 VOLTAGE TO R/HR CONVERSION TABLE RANGE 1 to 10" R/HR Volts Units R/hr 0 0.001 0.1 0.00135 0.2 0.001847 0.3 0.002511 0.4 0.003414 0.5 0.004641 0.6 0.006309 0.7 0.008576 0.8 0.011659 0.9 0.015848 1.0 0.021544 1.1 0.029286 C_,- 1.2 0.039810

' t 1.3 0.054116 1.4 0.073564 1.5 0.100 1.6 0.135935 1.7 0.184784 1.8 0.251188 1.9 0.341454 2.0 0.464158 2.1 0.630957 2.2 0.857695 2.3 1.165914 2.4 1.584893 2.5 2.154434 2.6 2.928644 2.7 3.981071 2.8 5.411695 2.9 7.356422 3.0 10.0

~ . . . - .

ATTACHMENT 6.5-5

\

UNIT 1 RMS-II CH #3 & UNIT 2 RMS-II CH #4 CONTAINMENT PURGE STACKS VOLTAGE TO R/HR CONVERSION TABLES RANGE 10 ' TO 10' R/HR Volts Units - R/HR

0. 0.1 0.1 .~ 135 0.2 .184 0.3 .251 0.4 .341 0.5 .464 0.6 .630 0.7 .857 0.8 1.165 1.584

(- s 0.9 1.

1.1 2.154 2.928 1.2 3.981 1.3 5.411 1.4 7.356 1.5 10.

1.6 13.593 1.7 18.478 1.8 25.118 1.9 34.145

2. 46.415 2.1 63.095 2.2 85.769 2.3 116.591 2.4 158.489 2.5 215.443 2.6 292.864 2.7 398.107 2.8 541.169 2.9 735.642

3. 1000.

.* 9p 2

l

ATTACHMENT 6.5-6

f. AUXILIARY BUILDING VENT STACK RMS-II CH #1 &

GAS STRIPPER BUILDING VENT STACK RMS-II CH #2 VOLTAGE TO R/HR CONVERSION TABLE RANGE 10-' TO 10' R/HR Volts Units - R/HR

0. 0.01 0.1 .013 0.2 .018 0.3 .025 0.4 .034 0.5 .046 0.6 .063 0.7 .085 0.8 .116 0.9 .158

1. .215 1.1 .292 1.2 .398 1.3 .541 1.4 .735

(

l.5 1.

' ~ l.6 1.35 1.7 1.847 1.8 2.511 1.9 3.414

2. 4.641 2.1 6.309 2.2 8.576 2.3 11.659 .

2.4 15.848 2.5 21.544 2.6 29.286 2.7 39.810 2.8 54.116 2.9 73.564

3. 100.

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i EPIP 7.3.2 MINOR Revision 6 07-01-83 POST-ACCIDENT SAMPLING AND ANALYSIS OF POTENTIALLY HIGH LEVEL REACTOR COOLANT

1.0 INTRODUCTION

1.1 This procedure outlines the steps necessary to collect, handle and analyze a high level reactor coolant sample which could result from gross fuel failure.

1.2 Equipment List Set up the following in the primary sample hood prior to collecting your sample:

1.2.1 The equipment detailed in Figures 2A and 2B, with the exception of the sample bomb.

,- 1.2.2 Two magnetic stirrers and two 50 ml poly beakers and a 50 ml beaker.

1.2.3 'A pH/mv meter, pH probe and chloride / reference elec-trodes.

1.2.4 A piston burette.

1.2.5 A lead brick wall of sufficient size to store residue from analysis.

1.2.6 Chemical transfer pump.

1.2.7 Remote handling tools located in the cabinet below the hood.

1.2.8 Prepare a 1.0 liter s_dearm flask with a correctly sized solid stopper and rubber septum over the sidearm.

1.2.9 Gas syringe The following equipment is also necessary for this procedure and need to be made ready.

1.2.10 The gas partitioner for H2 Enalysis.

- -. 3 -,

EPIP 7.3.2

.- Page 2 1.2.11 The special cart ' used for transport of the sample bomb.

1.2.12 Tools for connecting and disconnecting sample bomb; i.e.,

11/16" open end wrench or equivalent.

1.2.13 Remote valve turning tool. This tool as well as those mentioned in Step 1.2.12 are necessary for sampling and should be taken along and placed on the sample bomb transport cart.

The following reagents are also necessary and need to be prepared.

1.2.14 0.1N NaOH for boron. Obtain a supply from normal boron analysis.

1.2.15 2.0M HNO3 for chloride analysis.

1.2.16 Manitol for boron analysis.

1.3 Preliminary Steps Initials 1.3.1 Standardize the pH meter.

1.3.2 organize as much of the equipment as possible behind lead brick walls in an arrangement

'that allows for unobstructed view of all operations with the aid of the convex mirror.

(See Figure 2.)

1.3.3 Put new rubber septum on gas bomb.

1.3.4 Condition and check out chloride / reference electrodes.

1.3.5 Check out and prime the piston burette with fresh 0.1N NaOH solution.

1.3.6 Check out the operation and calibration of the chemical transfer pump by pumping chloride free deionized water through the pump.

2.0 PRELIMINARY EVALUATION NOTE: THIS EVALUATION SHALL BE COMPLETED PRIOR TO ANY ATTEMPT TO ENTER THE AUXILIARY BUILDING OR SAMPLE ROOM TO OBTAIN A REACTOR COOLANT SAMPLE UNDER EMERGENCY CONDITIONS. .

.: y

EPIP 7.3.2 t Page 3 ,

s 2.1 Possible Indication of Fuel Damage , Initials Some or all of the following would be present if fuel damage had occurred:

2.1.1 The letdown radiation monitor (R9) would be unusually high or offscale.

2.1.2 The Unit 1 containment radiation monitors (R11 and R12) or the Unit 2 containment radiation monitors (R211 and R212) would be unusually high or offscale.

2.1.3 The Unit 1 containment area monitors (R2 and R7) or the Unit 2 containment area monitors (R102 and R107) would be reading unusually high or offscale.

2.1.4 The auxiliary building stack monitor (R14) would show a significant increase due to auxiliary building airborne activity from the letdown and charging pump areas.

\ 2.1.5 Evaluation of Sample Room Conditions (A- a. The sample room area monitor (R6 or R106) and charging pump area monitor (R4 or R104) would give an indication of conditions in the auxiliary building and sample room.

b. After evaluation of the radiation monitoring system readouts, Health Physics will determine what airborne and radiation surveys would be appropriate before auxiliary building entry.

I

c. Verify the requirements for auxiliary building sample room entry, i.e., (1) RWP requirements, (2) clothing requirements, (3) respiratory

. requirements, and (4) dosimetry requirements including extremity dose monitoring requirements, and (5) health physics coverage requirements including timekeeping.

2.2 Possible Loss of Component Cooling 2.2.1 Verify that component cooling is still in service

to the affected unit. Sample cannot be obtained l without component cooling to sample room heat exchangers.

i

EPIP 7.3.2

,. Page 4 s 3.0 REACTOR COOLANT SAMPLING PROCEDURE NOTE: THIS PROCEDURE SHALL NOT BE INITIATED UNTIL THE EVALUATIONS DIS-CUSSED IN SECTION 2.0 HAVE BEEN COMPLETED AND REVIEWED BY DUTY &

CALL SUPERINTENDENT (COORDINATOR), DUTY HEALTH PHYSICS SUPERVISION AND THE DUTY SHIFT SUPERINTENDENT, AND THEIR APPROVAL HAS BEEN GRANTED.

3.1 Collecting a Pressurized Sample (Refer to Figure 1) Initials NOTE: THE FOLLOWING STEPS WILL BE ACCOMPLISHED UNDER THE DIRECTION OF HEALTH PHYSICS SUPERVISION AND ONLY AFTER COMPLETING SECTION 1.0 0F THIS PROCEDURE.

3.1.1 The following steps (a through f) must be accomplished before opening the incontainment sample isolation valve 955 (Step 3.1.2) and the hot leg sample isolation valve 966C (Step 3.1.3).

a. Verify that the demineralized water header pressure is approximately 100 to 120 psi.

NOTE: THIS STEP MAY BE DELETED IF REACTOR MAKEUP WATER IS USED FOR THE FLUSH.

n

b. Proceed to the sample room and install the shielded sample bomb on the outside wall of the sample room using the fittings provided.

CAUTION: BEFORE REMOVING THE SWAGE LOCK CAPS TO INSTALL THE BOMB, OPEN VALVES 939, 940 AND 941 TO RELIEVE SYSTEN PRESSURE. CLOSE VALVES 939, 940 AND 941. PLACE A WASTE BUCKET DIRECTLY UNDER THE BOMB. USE A PAPER TOWEL SHIELD AND RUBBER GLOVES WHEN f REMOVING CAPS.

c. Verify that the dimineralized water line is connected from valve 945 to the de-mineralized water manifold. Open demineralized H 2 O valves 945, 947, and 948 and check for leaks on the bomb' fittings.

Shut valves 945, 947, and 948.

d. Enter the sample room and close the following valves on the sample panel.

1. 961C - Normal hot leg sample bomb inlet.

2. 964C - normal hot leg sample bomb outlet.

m

EPIP 7.3.2 Page 5

\, -

Initials

3. 965C - normal hot leg sample bypass.

4. 968 - normal hot leg return valve.

5. 971 - normal hot leg sink sample valve.

e. Notify the control room that the failed fuel monitor (RE-109) will be taken out-of-service for sampling; then close:

938B - detector RE-109 inlet valve 938A - detector RE-109 outlet valve

f. Further verify that there is component cooling flow to the sample room heat exchangers by viewing the local flow indicator (FI-603) on the sample panel.

g. Open the following valves on the sample panel. l

1. 969A - sample system purge to volume

'T control tank.

t

2. 956C - normal hot leg sample supply valve.

3. 990 - residual heat removal sample supply valve.

h. Leave the sample room and position the following valves wide open. These valves are located on the sample room wall.

1. 939 - sample bomb inlet.

2. 940 - sample bomb bypass.

3. 941 - sample bomb outlet.

NOTE: FOR DRILLS AND TRAINING EXERCISES OPEN VALVE 941 ONLY ONE-QUARTER TURN TO ELIMINATE N-16 GAMMAS.

EPIP 7.3.2

. Paqa 6 3.1.2 Open the incentainment hot leg sample isolation valve (valve 955) and the residual heat removal sample isolation valve (valve 959) by means of the switches located outside the No. 1 pipeway for Unit 1 or No. 4 pipeway for Unit 2.

CAUTION: IF THE AFFECTED UNIT IS ALREADY ON RHR, LEAVE THE AREA IMMEDIATELY AFTER OPENING VALVE 959 AS BACK-GROUND RADIATION LEVELS WILL RISE SHARPLY.

If the valve No. 955 will not open because of containment isolation, perform the following steps (1 through 3).

1. Request that the Control Room reset the con-tainment isolation signal.

2. Turn the local control switch positions for valves 951, 953, and 955 to the "close" position.

s 3. Turn the local control switch for valve e s 955 to the "open" position.

r NCTE: SECTION 3.1.2 MUST BE ACCOMPLISHED BEFORE SAFETY INJECTION RECIRCULATION HAS BEGUN.

3.1.3 Leave the area and request control room to open the hot leg sample isolation valve (966C).

3.1.4 Verify sample flow by an increase in the radiation level.

3.1.5 After a recirc time of 30 minutes, return to the sample station and using the remote valve operating tool, fully open valves 9B and 9A and 8A and 8B.

a. Slowly and completely close valve 940.

b. Leave the primary auxiliary building.

NOTE: THE VALVE OPERATING TOOL SHOULD BE USED TO OPERATE ALL VALVES EXCEPT 945, 946, l 947 AND 948 (FLUSH VALVES).

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EPIP 7.3.2 Page 7 3.1.6 After 15 additional minutes, return to collect the sample. Close valves 9B and 9A and then valves 8B and 8A uniaq the remote valve operating tool. M de icte of the sample collection time.

NOTE: DO NOT DISCONNECT THE Sc 'LE BOMB UNTIL SAMPLE FLOW IS SECURED AND THE DI FLUSH IS COMPLETE AS EVIDENCED BY REDUCTION IN RADIATION LEVELS.

3.1.7 Request control room to immediately close the hot leg sample isolation valve 966C.

NOTE: M IS VERY IMPORTANT THAT THE HOT LEG SAMPLE ISOLATION VALVE 966C IS CLOSED PRIOR TO STARTING THE DI FLUSH. WAIT FOR CONFIRMATION FROM THE' CONTROL ROOM.

4.0 SAMPLE LINE FLUSHING 4.1 Leave valve 939 open, and fully open valves- 940 and 941.

_ Open valves 945 and 946. Allow the lines to flush for at least 15 minutes.

NOTE: 'D0 NOT REMAIN IN THE AREA 0F TUK' SAMPLE STATION DURING THIS FLUSH.

4.2 After about 15 minutes return and measure radiation levels.

If a Chemistry & Health Physics Supervisor determines that the levels are satisfactory, close whitey valve 946 and using the remote valve tool, close valve 939. Then open valve 947 and valve 948 and allow about a 15-minute DI flush.

4.3 After about 15 minutes, close valves 940 and 941 with the.

remote valve tool and then close valves 945, 947, and 948.

Disconnect valve 945 from the domineralized water manifold and cap both ends. Disconnect the sample bomb from the fittings using a paper towel to prevent spraying. Remove' the shielded sample bomb from its support. Remove excess liquid from the top and bottom bomb fittings with a syringe and dispose behind lead shielding. Replace the Swagelok caps on the wall fittings and on the bomb.

Transport the bomb, remote valve tool and wrenches to the chemistry lab on a cart.

NOTE: AFTER DRILLS AND TRAINING EXERCISES, RETURN ALL EQUIPMENT AND VALVE LINEUPS TO NORMAL.

~.

EPIP 7.3.2 Page 8 ,

s 4.4 Enter the sample room and notify control room that the failed fuel monitor (RE-109) is to be returned to service and that the sample room valving is to be returned to normal.

4.4.1 Close Valves:

1. 966C - Containment hot leg sample isolation valve

2. 961C - Normal hot leg sample bomb inlet

3. 964C - Normal hot leg sample bypass

4. 965C - Normal hot leg sample bypass

5. 971 - Normal hot leg sink sample valve

6. 939 - High level sample bomb inlet

7. 940 - High level sample bomb bypass

8. 941 - High level sample bomb outlet

.9 . 945, 946, 947, 948 - Demineralized water flush valves NOTE: DISCONNECT VALVE 945 FROM THE DEMINERALIZED WATER MANIFOLD. ADVISE CONTROL TO REDUCE DEMINERALIZED WATER HEADER PRESSURE TO NORMAL.

4.4.2 Open Valves:

1. 938B - Failed fuel monitor RE-109 inlet

2. 938A - Failed fuel monitor RE-109 outlet

3. 956C - Normal hot leg sample supply valve

4. 968 - Normal hot leg return valve

5. 969A - Sample system purge to volume control tank 5.0 SEPARATICN OF THE PRESSURIZED SAMPLE AND ANALYSIS OF THE GASEOUS AND LIQUID COMPONENT (Refer to Figures 2A and 28) 5.1 Collecting the Gaseous Sample From the Pressurized Sample 5.1.1 Place the shielded sample bomb in the sample holder in the primary sample hood.

z

EPIP 7.3.2

- Page 9

\

5.1.2 Connect the sample bomb to the shielded gas collection bomb by means of the fittings provided. Place lead bricks in the area of this connection for shielding.

5.1.3 Connect the valve manifold to the opposite end of l the sample bomb and verify that valve 11 on the manifold is open. l l

5.1.4 Hake sure that the vacuum line is attached to !

the gas collection bomb at the valve 1 location.

Open valves 1 and 2. Secure valve 3. Evacuate the gas bomb and connecting lines. With vacuum still on, secure valve 1. Secure vacuum.

5.1.5 Before proceeding, make sure no inleakage has occurred into the gas bomb by observing the vacuum gauge reading. Using the remote valve tool, fully open valves 9A and 9B. Open valve 8A one-quarter turn. Crack open valve 8B and control degassing by throttling valve 8B.

Allow the system to de cas for 5 minutes. .

_ Check that valves 9A t- ' 9B, 8A and 8B are fully open. Close valve 2.

NOTE: OBSERVE THE VACUUM GAUGE. THE VACUUM SHOULD DROP VERY SLOWLY. IF THE DROP IS TOO RAPID, CLOSE VALVE 8B SLIGHTLY.

DROP SHOULD BE 5-10" HG/ MIN.

5.2 Analysis of Gaseous Sample 5.2.1 Hydrogen Use a syringe to draw a 1 cc sample. Use the injection port on the gas partitioner for this analysis.

5.2.2 Radioactive Noble Gas Use a syringe to draw a 1/2 cc sample and inject this into the flask prapared in Section 1.2.8.*

Allow 30 minutes for thermal mixing. Draw a 1/2 cc sample of this dilution and proceed as no rmal.

Additional dilution should be performed if the contact reading is >l mr/ hour.

NOTE: SEE SECTIONS 7.0 AND 8.0.

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Pag Ib s

5.3 Collecting the Liquid Sample From the Pressurized Sample 5.3.1 Add one drop of 2 M nitric acid (per 10 ml of sample) to the chloride beaker for pH

, adjustment.

NOTE: IT IS EXTREMELY IMPORTANT TO VERIFY THAT VALVE 2 HAS BEEN CLOSED BEFORE PROCEEDING.

5.3.2 Open valve 3 slowly. Allow the liquid sample to drain into the 50 m1 beaker. Direct a slow stream of air through the vent line on valve 3 if necessary to recover the total liquid sample.

5.3.3 Close valves 8A and 88, 9A and 98, and valve 3.

5.4 Analysis of Liquid Samples 5.4.1 Boron /pH Analysis

a. Transfer a 5 mi sample using the chemical transfer pump into a 50 ml poly beaker containing a stir bar.

NOTE: OBSERVE THE TRANSFER PUMP OPERATION.

t WHEN SAMPLE BEGINS TO ENTER THE BEAKER, THE TRANSFER RATE IS 0.5 ML/SECOND OR 10 SEC = 5 ML.

h. After transfer is complete, record the pH.

c. Plug in the magnetic stirrer, add mannitol, and proceed with the boron analysis.

NOTE: IF THE PRIMARY SYSTEM HAS BEEN BORATED, 5 ML OR MORE OF TITRANT MAY BE NEEDED TO REACH AN ENDPOINT.

NOTE: AFTER DRILLS AND PRACTICE RUNS, THE BORON TITRATOR MUST BE FLUSHED WITH DEIONIZED WATER AND PUT IN DRY LAY-i UP. ALL ELECTRODES SHOULD BE PLACED IN LAYUP SO THEY ARE CONDITIONED FOR

' IMMEDIATE USE.

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3

EPIP 7.3.2 )

.. Page 11 l

\ -

5.4.2 Chloride Analysis Transfer the remainder of the sample to the second poly beaker containing the chloride electrode. Start the stirring action and record the potential. Use the calibration curve for the chletide electrode to determine chloride concentration.

NOTE: CHLORIDE SAMPLE MUST BE pH ADJUSTED.

SEE STEP 5.3.1.

I NOTE: HIGH LEVELS OF RADIOACTIVE IODINE IN THE COOLANT WILL INTERFERE WITH THE CHLORIDE ANALYSIS. REFER TO THE ATTACHED IODINE /

CHLORIDE CORRCCTION CURVE (ATTACHMENT 7.3.2-1) TO MAIG THE PROPER ADJUSTMENT TO THE CHLORIDE ELECTRODE ANALYSIS.

i 5.4.3 Iodine Analysis and Gamma Scan Using the specially prepared 2 cc syringe, with-draw 0.3 cc of the sample from the poly beaker used for the chloride analysis and inject this sample into a 1000 ml poly bottle containing

( ' demineralized water. Make additional dilutions in the same manner.* Count as normal.

Additional dilution should be performed until the contact reading is <1 mr/ hour.

NOTE: SEE SECTIONS 7.0 AND 8.0.

5.5 Reporting.of Results Complete and forward Reactor Coolant Post-Accident Sampling Analysis Report (EPIP-30).

6.0 SAMPLE RESIDUE Placs all sample residue in the specially prepared lead pig for disposal.

7.0 SAMPLES TAKEN TO KEWAUNEE NUCLEAR PLANT FOR COUNTING

Reference:

Post-accident counting agreement with Wisconsin Public Service, Kewaunee Nuclear Plant.

Kewaunee Nuclear Plant does not utilize the 5 ce glass vial and 1 cc test tube geometries. Therefore, " normal" samples will have to be diluted and placed in one liter poly bottles if they are sent to the Kewaunee Nuclear Plant for analysis. ,

7_ .,

- - _ _ _ - - - - . _ _-_ ~ . . . -. .

EPIP 7.3.2 l

' Page 12 8.0 LABELING OF SAMPLES Label all chloride, noble gas, iodine and gama scan samples with all pertinent information such as: sample number, name of sample, date and time of sampling, sample volume and dilution (s).

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ATTACHMENT 7.3.2-1 CORRECTION FOR REACTOR COOLANT IODINE INTERFERENCE WITH CHLORIDE ELECTRODE RESPONSE

1.0 INTRODUCTION

Chloride as determined by the chloride specific ion electrode is subject to interference caused by the presence of high levels of other halogens, specifically iodine. The following procedure outlines the method used to estimate the correction for this interference.

2.0 PROCEDURE 2.1 Measure the chloride concentration using the chloride specific ion electrode.

2.2 Perform the iodine analyses as outlined in Section 5.4.3 of this procedure.

2.3 Convert the iodine concentration to ppm using the following conversion factors.

Isotope pCi/cc + ppm Conversion Factor I-130 2.54 X 105 I-131 3.85 X 10~4 -

I-132 4.71 X 10 s -

I-133 4.23 X 10 s I-134 1.78 X 10~8 -

I-135 1.38 X 10 s 2.4 Sum the concentration, in ppm, of each iodine isotope and use the attached curve to determine the estimated chloride concentration correction factor.

3.0 EXAMPLE Reactor Coolant Analysis Estimated Isotope Concentration (pCi/cc) Conversion-Factor Concentration (pcm)

I-131 2.56 X 103 pCi/cc X 3.85 X 10 ~4 9.86 X 10~1 I-133 1.45 X 102 pCi/cc X 4.23 X 105-6.13 X 103 2

I-135 5.65 X 103 pCi/cc X 1.38 X 10 s 7.80 X 10 TOTAL 1.07 ppm Chloride concentration from correction curve (ppm) = 0.42 ppm Subtract this value from the chloride concentration determined by the chloride electrode.

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/- Revision 5 07-01-83 POST-ACCIDENT SAMPLING OF CONTAINMENT ATMOSPHERE

1.0 INTRODUCTION

This procedure outlines the steps necessary to collect, handle and an'alyze a potentially high radioactive containment atmosphere sample resulting from gross fuel failure and loss of reactor coolant system integrity to determine hydrogen and radioactive gas concentrations.

2.0 PRELIMINARY EVALUATION NOTE: THE FOLLOWING EVALUATION SHALL BE COMPLETED PRIOR TO ANY ATTEMPT TO ENTER THE FACADE OR R11/R12 CUBICLE TO PERFORM A VALVE LINEUP OR COLLECT A CONTAINMENT ATMOSPHERE SAMPLE.

2.1 Indications of Possible Fuel Damage Some or all of the following would be present if fuel damage or loss of reactor coolant system integrity had occurred:

2.1.1 Yhe letdown radiation monitor (R9) would be unusually high or offscale.

2.1.2 The containment radiation monitors 1R11 and 1R12 or 2R11 and 2R12 would be unusually high or offscale.

l 2.1.3 The containment area monitors 1R2 and 1R7 or 2RE-102 and 2RE-107 would be unusually high or offscale.

2.1.4 The automatic actions of EOP-1A have caused containment isolation.

2.2 Evaluation of Radiological Hazards in Access Areas i Required for Sampling Initials 2.2.1 After evaluation of the radiation monitoring system readouts, verify with Health Physics that the appropriate airborne and radiation surveys have been made before entering the facade.

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. % g

e EPIP 7.3.3 (3 ,

Initials 2.2.2 Verify the requirements for facade and 1R11/1R12 or 2R211/2R212 cubicle entry.

a. Radiation work permi.t requirements

b. Clothing requirements

c. Respiratory requirements

d. Dosimetry requirements including extremity dose monitoring

e. Health physics coverage requirements including timekeeping 3.0 CONTAINMENT ATMOSPHERE SAMPLING PROCEDURE NOTE: THE FOLLOWING PROCEDURE SHALL NOT BE INITIATED UNTIL THE EVALUATION DISCUSSED IN SECTION 2.0 HAS BEEN COMPLETED.

THE DUTY & CALL SUPERINTENDENT (COORDINATOR), THE DUTY AND CALL, HEALTH PHYSICS SUPERVISOR AND THE DUTY SHIFT C SUPERINTENDENT,SHALL APPROVE THE IMPLEMENTATION OF THIS PROCEDURE. THE FOLLOWING STEPS WILL BE ACCOMPLISHED UNDER THE DIRECTION OF HEALTH PHYSICS SUPERVISION.

3.1 Unit 1 Containment Atmosphere Sampling Using the IR11/1R12 Sampling System 3.1.1 Valve Lineup

a. Entry into the facade shall be from the potable water room.

l b. Verify with the control room that the 1R11/1R12 sampling system is lined up in accordance with the following:

(1) The containment isolation valves 3200C and 3200G are closed.

(2) The 3200A&B A0V's in the R11/R12 cubicle are closed.

(3) The IP707 forced purge pump is secured.

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EPIP 7.3.3 Page 3

, Initials

c. Proceed to the R11/R12 cubicle, and place the IR11/1R12 sampling system in the following lineup:

(1) Open valves 1-3200J and 1-3200L.

(2) Close valve 1-3200H.

(3) Verify that the continuous vent valves 1-3200S and 1-3200W are closed.

(4) Close valve 1-3200K.

d. Connect the 100' service air hose to the Chicago fitting located on the elevator side of the cubicle established for sample line purging.

NOTE: THE AIR HOSE IS STORED INSIDE THE 1R11/1R12 CUBICLE, ALONG WITH THE

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PRESSURE REGULATOR.

e. Verify that valve 1-3200Y is closed.

f. Proceed to the potable water room with the air hose and pressure regulator.

NOTE: FOR RCT TRAINING EXERCISES, DO NOT USE THE POTABLE WATER ROOH SERVICE AIR H00KUPS, SINCE THEIR USE UNNECESSARILY BURDENS SECURITY. THE

_ SERVICE AIR H00KUPS LOCATED ON THE EL. 66' CONTROLLED SIDE FACADE WILL ARE TO BE UED.

(1) Attach the service air hose to an available turbine hall service air outlet using the pressure regulator !

designated for this procedure.

(2) Lock the Chicago fitting in place l using the appropriate pins. i

g. Verify with the control room that the !

1Ril/1R12 monitor is lined up to monitor, !

and discharge to, containment atmosphere.

h. Request the control room to open the ,

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~ 1-3200A&B A0V sample system valves and the containment isolation valves 1-3200C&G.

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EPIP 7.3.3 Page 4 Initials

1. Verify sample flow by an increase in radiation levels in the cubicle, and return to the potable water room.

j. Allow five minutes for sample recircing.

Request control room to stop the pump and close 1-3200A, B, C & G A0V isolation valves.

3.1.2 Sample Collection

a. Enter the facade and proceed to the sample point. Take along two 2 cc gas syringes in a hollowed-out lead break.

CAUTION: CONTAINHENT PRESSURE COULD THE0RETI-CALLY BE AS HIGH AS 60 PSIG. HOLD THE SYRINGE PLUNGER SECURELY. ALSO, LOCK THE GAS SYRINGES BEFORE WITH-DRAWING FROM THE SAMPLING SEPTUM, USING THE BUILT-IN LOCKING DEVICE ON

(,

THE SYRINGE..

b. Remove the set screw and take one 1/2 cc and one 1 cc gas sample. Place the syringes in ,

the lead brick. t

c. Replace the set screw and open valve 32007.

d. Leave the facade and return to the potable water room.

l 3.1.3 Sample Line Purging (

l

a. Open service air valve and adjust the ;

pressure to a setting of 10 psig greater i than measured containment pressure. !

i

b. Request control room to open the pump discharge isolation valve 1-3200A.

c. Request control room to start the 1Rll/lR12 pump. ;

d. Verify purge effectiveness by measuring ,

the reduction in radiation levels after !

approximately 15 minutes of purging.

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O EPIP 7.3.3 Page 5 Initials

e. Request control room to close the 1-3200A pump discharge isolation valve, secure the pump, and open the 1-3200B&C pump suction isolation valves.

f. Continue the purge for 10 more minutes.

g. Request control room to close all pump discharge and suction isolation valves and secure the pump.

h. Secure the service air valve.

NOTE: AF'hERDRILLSANDPRACTICERUNS, RETURN ALL EQUIPMENT AND VALVE LINEUPS TO THE AS-FOUND CONDITION.

3.2 Unit 2 Containment Atmosphere Sampling Using the R211/R212 Sampling System, After Containment Isolation 3.2.1 1 Valve Line-Up

a. Verify with the control room that the following conditions exist:

(1) Valves 3200A, B & C have closed.

(2) Pumps P707A&B have been secured.

(3) Valve 3200G is lined up to return g i flow to the containment (Position 1). g I

( (4) Valves 3200D, E, F & H are closed.

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b. Obtain the designated 100' service air l line and pressure regulator from the Unit 1 ;

Rll/R12 cubicle. j j

c. Proceed to the nonnuclear room with the ;

l air hose and pressure regulator.

NOTE: FOR RCT TRAINING EXERCISES, DO NOT f USE THE NONNUCLEAR ROOM SERVICE AIR i H00KUPS, SINCE THEIR USE UNNECESSARIL*l BURDENS SECURITY. THE SERVICE AIR H00KUPS LOCATED ON EL. 66' CONTROLLED SIDE FACADE WALL ARE TO BE USED.

-M.

5

EPIP 7.3.3 Page 6 Initials (1) Attach the service air hose to an available turbine hall service air outlet using the pressure regulator designated for this procedure.

(2) Lock the Chicago fitting in place using the appropriate pins,

d. Proceed to the R211/R212 cubicle and perform the following:

(1) Connect the service air hose to the Chicago fitting attached to valve 3200Y, and verify that the valve is closed.

NOTE: THE VALVE IS LOCATED ON THE i ELEVATOR SIDE OF THE R211/R212 l CUBICLE.

(2) Enter the cubicle and close valve C' '

3200B.

e. Request the control room perform the following:

NOTE: THE SAMPLE PUMP P707A SEALS ARE l RATED FOR 5 PSI. TO PREVENT AN j INADVERTENT AIRBORNE RELEASE, !

THE CONTAINMENT PRESSURE MUST .

BE LESS THAN 5 PSI BEFORE PRO- ;

CEEDING. ;

(1) Verify containment pressure is less than 5 psi.

(2) Place the sample mode switch located in the cable spreading room in the septum position.

(3) Open A0V's 3200A&B and the containment valve 3200C.

(4) Start sample pump P707A.

f. Verify sample flow by observing a flow increase on FIT-3288 and return to the nonnuclear room. .

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EPIP 7.3.3 Page 7 Initials

g. Allow 15 minutes for sample recirculation and return to the cubicle with two (2) 2 ce gas-tight syringes in a hollowed-out lead brick.

3.2.2 Sample Collection

a. Withdraw a 1 cc gas sample and a 1/2 ce gar., sample from the sample septum by means of the gas-tight syringes. Lock the syringe plunders, and place them in the lead bricks.

b. Establish a service air pressure of 5 psi on the regulator and open valve 3200Y.

3.2.3 Sample Line Purge

a. Return to the nonnuclear room and request the control room close valves 3200B&C.

s C b. Allow the system to purge forward for 25 minutes, then request the control room perform the following:

(1) Secure the sample pump P707A.

(2) Close AOV-3200A and place valve j

3200S in its normal position (Position 1). !

1 (3) Open A0V-3200B&C. f I

c. Allow the system ta purge backwards 25 minutes and request the control room close !

valves 3200B&C. } !

d. Return to the cubicle and perform the following:

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(1) Verify the effectiveness of the purge .

i by measuring the reduction of the !

radiation levels.

(2) Enter the cubicle and open valve 3200K. -

(3) Close the service air hose regulator, ,

close valve 3200Y and remove the .

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. service air hose. '59k$

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EPIP 7.3.3

. Page 8 i

Initials

e. Notify the control room that sampling has been completed.

NOTE: AFTER ALL DRILLS AND PRACTICE EXERCISES, ALL EQUIPMENT AND VALVE LINEUPS ARE TO BE RETURNED TO THE AS-FOUND CONDITION.

4.0 CONTAINMENT ATMOSPHERE ANALYSIS 4.1 Volume Adjustments 4.1.1 Before proceeding with the hydrogen, radioactive noble gas and radioactive iodine analyses, the sample contents of both syringes must be brcught to atmospheric pressure. Use the shielded sacri-ficial glass bomb for this purpose.

4.1.2 Insert the syringe through the rubber septum of the glass bomb. Unlock the syringe locking device

,and let the syringe and bomb equilibrate for

(, _ approximately 30 seconds. Relock the syringe and withdraw from the septum. Store the syringe in the lead brick.

4.1.3 Repeat Step 4.1.2 for the other syringe.

4.2 Hydrogen Assuming normal equipment setup and preparations are complete, use the sample injection port on the gas partitioner, inject a.1 cc gas sample and proceed using normal hydrogen analysis procedures.

4.3 Radioactive Noble Gases and Iodines 4.3.1 Couple the silver zeolite column and the evacuated glass l bomb as shown in Figure 3.

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4.3.2 Inject the 1/2 cc containment atmosphere sample into the column through the septum. Remove the syringe and store i

in the lead brick.

4.3.3 Using a second syringe tip (without the plunger) puncture the column septum and fully insert the syringe tip.

1

EPIP 7.3.3 Page 9 4.3.4 Crack open the upper glass bomb valve between the bomb and the column. Allow air to slowly bleed thrcugh the column to the bomb. Flow can be serified by observing the gentle reflux of the siver zeolite surface. When flow has ceased, momentarily open the upper valve wide open and then close completely.

4.3.5 Remove the column from the bomb and store the bomb behind shielding in the hood.

Iodines 4.3.6 Remove the septum from the column and pour the silver zeolite

- into a counting test tube and cap.

1 4.3.7 If the silver zeolite / iodine sample is less than one mR/hr contact, count the sample directly on the MCA using the test tube geometry and a multiplying factor of two.

4.3.8 If the iodine activity is too high to count directly, determine the radiation level in mR/hr at one foot. Also, determine the percent isotopic composition using the MCA and attenu-ating the sample if necessary.

s Determine the total iodine activity using the following equation:

= R/hr @ l' C(Ci) 6E T

The following table can be used as an aid in determining the

" Total Average Energy /y" (E T

) f r the above equation.

(1) (2) (1) x (2)

Average y % Composition Weighted y Isotope Enerov (Fraction) Energy 131 I 0.380 -

I 0.731 I 0.530 I 0.857 I 1.238 Total Energy /y (ET) l Each individual iodine isotopic concentration (pCi/cc) is calculated as follows:

pCi/cc I g) =C x 2 x 10 x (Fractional % Composition) l- - ,. 4, l

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Noble Gases 4.3.9 Withdraw a 1/2 cc gas sample from the glass bomb in Step 4.3.5.

If the contact reading en the syringe is less then one mR/hr, inject the sample into a , cc vial and count as normal. If the contact reading is greater than one mR/hr, further dilutions are necessary.

Dilution procedure: Inject the 1/2 cc sample into a one liter poly bottle and allow to equilibriate. Withdraw 1/2 cc from the poly bottle. If the contact reacing is less then one mR/hr, inject into a 5 cc vial and count as normal. If the contact reading is greater than one mR/hr, dilute further.

NOTE: FOR EACH DILUTION, USE A NEW SYRINGE OR MAKE CERTAIN l

THE OLD SYRINGE IS COMPLETELY PURGED. LABLE NOBLE GAS SAMPLES AND IODINE SAMPLES WITH ALL PERTINENT INFOR-MATION SUCH AS: SAMPLE NUMBER, NAME OF SAMPLE, DATE AND TIME OF SAMPLE, SAMPLE VOLUME AND DILUTION (S).

5.0 CALCULATIONS

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5.1 Pressure.and Temperature Correction Ask control room supervision for the temperature and pressure in containment (psig and *F). Determine atmospheric pressure (mm Hg) and temperature ('F) from the barometer in the laboratory. Convert this to psia. Apply the following correction factor to all results:

Initials concentration (pci/cc) X (P + 14.7) T + 459 P

yg Tyg + 459 i

P

(*" "U}

lab (psia)= 14.7 psi x 760 mm Hg I

i 5.2 Complete and forward Containment Atmospheric Post-Accident Sampling Analysis Report (EPIP-31).

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,~a g s t mconsn ascmc Pows coww Q, et Q "2 Point Beach Nuclear Plant l

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e' EPIP 16.1 -

MINOR

/ Revision 4 '

07-01-83 NUCLEAR ENGINEERING SECTION NOTIFICATION AND RESPONSE 1.0 GENERAL 1.1 Establish notification responsibilities and sequence for designated Nuclear Engineering Section personnel.

l 1.2 Provide notification to personnel from the plant and corporate offices.

1.3 Augment the emergency organization human resources as necessary during a Point Beach Nuclear Plant condition classified as an alert, site emergency, or general emergency in accordance with the Emergency Plan Implementing Procedures.

.Necessary phone numbers are included in Attachment 16.1-1, " Nuclear Engineering Section Emergency Call List."

2.0 PRECAUTIONS AN6 LIMITATIONS 2.1 If immediately unable to contact an individual, continue with the notification of the other individuals and then attempt to contact the persons who have not been notified.

2.2 All notifications should be appropriately logged on Attachment 16.1-1.

3.0 INITIAL CONDITIONS Alert, site emergency or general emergency conditions exist.

4.0 PROCEDURE 4.1 Emergency Support Manager 4.1.1 Contact the Emergency Director listed on Attachment 16.1-1 and discuss the emergency classification.

4.1.2 Contact as appropriate other designated Engineering Section personnel listed on Attachment 16.1-1. Have one of the Nuclear Engineering Section personnel contact the Duty Public Information representative at 271-7117 if appropriate.

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EPIP 16.1 Page 2 7,

4.1.3 Report to the emergency support center and assume responsi-bility for operations of the emergency support center.

4.1.4 Establish a communications link with the technical support center as soon as practicable.

4.1.5 Contact INPO and inform them that we have had an emergency.

Give them the classification and a short description of the incident. Also inform them that we will give them an update on conditions at the plant when more information is available.

See form EPIP-23, "Offsite Agency Emergency Call List."

4.2 Emergency Director 4.2.1 Report to WE corporate headquarters and coordinate as appro-priate WE general office emergency response and recovery operations in support of the plant's emergency and recovery effort.

4.2.2 Establish a ' communications link with the emergency support center as soon as practicable.

4.3 Designated Nuclear Engineering Section Personnel (Technical (1 -

Support Center) i (See Emergency Plan Figures 5-4, 5-5, and 5-6 for individual reporting l to the technical support center. This would include the Chemistry /

Health Physics Supervisor and the System / Analysis & Procedural Support Coordinator.)

Report to the technical support center.

4.4 Designated Nuclear Engineering Section Personnel (Emergency ,

Support Center)

(See Emergency Plan Figures 5-5 and 5-6 for individual reporting to the emergency support center. This would include the RadCon/ Waste Manager and Radwaste/ Technical Support Coordinator.

Report to the emergency support center.

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EPIP 16.1 Page 3 ATTACHMENT 16.1-1 WE General Office Company Ext. Home Phone Time Notified Sol Burstein Executive Vice President (Emergency Director)

C. W. Fay, Vice President-Nuclear Power '

(Emergency Support Manager) i D. K. Porter (Design, Construction & l Planning Manager)

R. A. Newton (Systems, Analysis & Procedural Support Coordinator) ,

G. D. Frieling :

(Radwaste Technical Support Coordinator)

E. J. Lipke (Rad Con / Waste Manager) ,

i S. Cartwright !

(Utility Engineering Director)

{

R. K. Hanneman i I

f J. Jacovitch .

i D. F. Johnson ,

C. W. Krause (Licensing Support Coordinator)

S. A. Schellin R. D. Seizert ,

T. J. Rodgers T. A. Hanson ',

s .

W. F. Geisheker _ c _M i

W. M. Adams ,

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EPIP 16.1 Page 4

/

WE Ceneral Office Company Ext. Hcme Phone Time Notified D. M. Stevens G. M. Krieser R. E. Heiden i

D. R. Dunham D. B. Robinson W. F. Wilson i

N. H. Palmer i D. C. Kois -

C. N. Peters ,

J. E. Peters l t

J. R. Donahue I i

J. E. Knorr j a

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POINT BEACH NUCLEAR PLANT WORKSHEET FOR STATUS REPORT ON RADIATION MONITORING SYSTEM FOR UNIT

1. Unit Process Monitors Monitor No. Instrument Reading Comment 1R11/2RE-211 Containment Particulate Monitor 1R12/2RE-212 Containment Gas Monitor 1(2)R15 Steam Air Ejector 1R16/2RE-216 Containment Fan Cooler Liquid Monitor 1R19/2RE-219 Steam Generator Blowdown Liquid Monitor RMS-II Channel 3 Containment Purge Stack (Unit 1)

RMS-II Channel 4 Containment Purge Stack (Unit 2) :

2. Unit Area Monitors Monitor No. Instrument Reading Comment 1R2/2RE-102 Containment El. 66' 1R4/2RE-104 Charging Pump Hallway 1R6/2RE-106 Primary Side Sample Room l 1R7/2RE-107 Seal Table El. 46'

( Containment I

1R9/2RE-109 Primary Coolant Failed-Fuel Completed By Time Date "4.,

l EPIP-05 (07-83)

~ -

O POINT BEACH NUCLEAR PLANT WORKSHEET FOR STATUS REPORT ON RADIATION MONITORING SYSTEM FOR PLANT

1. Plant Process Monitors channel No. Instrument Reading Comment R14 Auxiliary Building Vent Stack R21 Drumming Area Vent Stack RE-225 Combined Steam Air Ejector RE-224 Gas Stripper Building Vent Stack RMS-II Auxiliary Building Channel 1 Vent Stack RMS-II Drumming Area Vent Channel 2 Stack t

RMS-II Combined Steam Air Channel 5 Ejector RMS-II Gas Stripper Building Channel 6 Vent Stack

2. Plant Area Monitors Channel No. Instrument Reading Comment R1 Control Room

. R3 Chemistry Lab R5 Spent Fuel Pit R8 Drumming Area Completed By Time Date

,. '4

--3 EPIP-06 (07-83)

~

POINT BEACH NUCLEAR PLANT SITE BOUNDARY CONTROL CENTER EMERGENCY PLAN INVENTORY CHECKLIST Item No. Item Required On Hand Sampling Equipment and Supplies

1. AC generator (5,000 watt) 1

3. Electric high volume air sampler 1

4. Poly gas sample bottles 12

5. Charcoal cartridges for air sampler, high volume 48

6. Charcoal cartridges for air sampler, low volume 50

7. Silver zeolite cartridges for air sampler, low volume 5

8. Filters for air samplers (pkg. of 100) 2

9. Gasoline for AC generator (gallons) 2

10. Sample tags 50

11. Plastic bags 50

12. 100' extension cord 2

13. Planchets 20 13a. High volume air sample filter 1 13b. 1 qt. oil 1 13c. 1 pt. OB oil 1 13d. Flask 1 13e. Dizalene : 1 Respiratory Protection Equipment

14. Full-face respirators 4

16. Full-face filter cartridge 12 Fire Protection Equipment

18. Fire extinguisher, dry chemical 1 Radiation Survey and Monitoring Instrument

19. Jordan Radgun (.01 mR/hr - 10 kR/hr) 1

20. Radector III (.1 mR/hr - 1,000 R/hr) 1

21. Victoreen Model 490 Thyac III 1

22. PIC-6A survey instrument (1 mR/hr - 1,000 R/hr) 2

23. RM3C personnel survey frisker 1

24. Johnson Associates, GSM-5, 0-50k cpm, 0-200 mR/hr 1

25. MSC-1 sampler holder for GSM-5 1

26. Check sources; 2 - Cs-137 and 1 - Sr-90 3 7

27. Filters for smears (pkg. of 100) 2 l 28. Nuclear Chicago counter scaler 1 l 29. Coin envelopes (box) 1/2 HP-210 probe

30. 2

31. SH4 probe holder 1

32. Earphones for Thyac III survey instrument 3 i 33. Side window probe 2 _

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EPIP-24a (05-33) l

Page 2 Required On Hand

34. Cord, BNC-BNC connector 2

35. Cord, amphenol - BNC connector 2 3ba. HPI 1010 1 Personnel Monitoring Equipment

36. Personnel Thermoluminescent Dosimeters (TLD) 100

37. Radiological dosimeters, 0-5 R 12

38. Radiological dosimeter charger 2 38a. "AA" batteries, boxes 3 First Aid and Decontamination Supplies

39. First aid kit 1

40. Burn kit 1

41. Emergency drinking water tablets (bottles; 50 tables per bottle) 5

42. Water (gallons) 20

43. Decon soap, powder (5 lb.) 1

44. Decon soap, liquid (qt.) 1

45. Hand brush 4

46. Cotton applicators (box) 1

47. Potassium permanganate (4 oz.) 1

48. Sodium bisulfate (1 lb.) 1

49. Kim towels (box) 1

50. Masselin (pkg.) 1

51. "409" cleaner (btl.) 2

52. " Spic'n Span" (box) 2

53. Masselin mop , 1

54. Regular sponge map 2

55. Rag mop 1

56. Wringer 1

57. Large mop bucket 1

58. Kimwipes (box) 1

59. Bucket, plastic 2

60. Cotton swabs (packets) 5

61. Gauze sponges, 2" x 2" (100 per pkg.) 1

62. Nail brushes 4 62a. Potassium Iodide tablets, bottles 4 Radiation Hazard Signs and Supplies

63. Radiation warning tape (roll) 1

64. Radiation placards 10

65. Radioactive material and radiation hazard tags 10

66. Radiation contamination hazard tags 10

67. Contamination, high radiation, radioactive -

material, and radiation area inserts (ea.) 10

68. Yellow / magenta ribbon (rolls) 8

69. Yellow / magenta rope (roll) 1 f

EPIP-24a (05-83)

4 Page 3 Required On Hand Communication Equipment and Supplies

70. Portable 2-way radio KRQ-717 1

71. Telephone, plan PBX-extension with outside line capability 1

72. WE telephone book 1

73. Two Rivers /Manitowoc telephone book 1 Traffic Signs and Equipment

74. Traffic cones for barricades 20 l 75. Traffic lights for barricades 8

76. Chains and padlocks for barricades 4

77. Traffic flashlight 4

78. " Closed Area" placards 6

79. Traffic warning light batteries (spare)

Clothing and Toiletry Supplies 79a. Cloth hoods 25

80. Coveralls 25

81. Rainwear 6

82. Rubber boots 10

83. Shoe covers; plastic 25

84. Overshoes, winter 6

85. Gloves, rubber disposable 6

86. Gloves, cotton disposable 6

87. Mittens, winter .6

88. Towels 12

69. Washcloths 12 89a. Lineman's gloves 6 89b. Rubber gloves, boxes 1 Stationery and Miscellaneous Supplies

90. Desk table and chair 1

91. Writing paper (pad) 1

92. Pens and pencils -

Assortment

93. Tape, masking (rolls) 2

94. Tuck tape (rolls) 10 l 95. Logbook ~ 1

96. Absorbent paper (roll) 1

97. Paper cups (box) 1

98. Plastic bags 50

99. Scissors 1 100. Pocketknife 1 101. Screwdrivers (set) 1 102. Plastic funnel 4 103. Flashlight 1 104. Batteries (for flashlight and survey instruments) 50 105. Flashlight bulbs 6 '.___,

- ' .36.,

, EPIP-24a (05-83)

O Page 4 Required On Hand Statior.ery and Miscellaneous Supplies, continued ...

106. Bulbs, incandescent 8 107. Electric clock 1 108. Electric heater 5 109. Wet / dry vacuum cleaner 1 110. Metal drum (55-gallon) 1 111. Dz1-lene (quart), gasonline stabilizer 1 112. Lead bricks 12 113. Safety solvent (low) 1 114. Metal funnels 2 115. Pencil sharpener 1 116. Chalk 1 117. Bulletin Board 1 118. Chalkboard 1 119. Table (reg.) 1 120. Picnic tables 2 121. Calculator 1 Emergency Plan Documents 122. PBNP Emergency Plan 1 122a. Emergency Plan Implementing Procedures 1

, 123. Health Physics Administrative Control Policies

& Procedures Manual 1 124. Dose Isopleth/ Map Package 1 125. Personnel Roster 10 126. Potassium Iodide Approval, Use List 1 127. DOE, Region V, Radiological Assistance Handbook 1 128. State of Wis. Peacetime Radiological Response Plan 1 EPIP Forms 129. EPIP-01, Emergency Plan Airborne Radiation Survey 5 130. EPIP-02, Emergency Plan Survey Record 5 131. EPIP-03, Dose Factor Calculation Sheet 5 132. EPIP-04, Status Report on Plant Systems and Controls 5 133. EPIP-05, Worksheet for Status Report on RMS for Unit 5 134. EPIP-06, Worksheet for Status Report on RMS for Plant 5 135. EPIP-07, X/Q Determination 5 136. EPIP-08, Estimated Whole Body and Thyroid Projected 5 137. EPIP-09, Estimated Whole Zody CalcuIation Worksheet 5 138. EPIP-10, Estimated Ground Deposition Calculation 5 139. EPIP-17, List of Missing Personnel 5 140. EPIP-18, Assembly Area Roster 5 141. Xe-133 Equivalent Release Rate, Worksheet No. 1 5 141a. EPIP-36, Master Dose Logsheet 5 EPIP-24a (05-83)

i Page 5 Required On Hand EPIP Procedures 142. EPIP 1.4, Radiological Dose Evaluation 5 143. EPIP 1.5, Protective Action Evaluation 5 144. EPIP 7.1.1, Chemistry & Health Physics Personnel Notification and Initial Response when Chemistry G

-Health Physics Personnel are On-Site 5 145. EPIP 7.2.1, Activation of Health Physics Facilities at Site Boundary Control Center 5 146. EPIP 7.2.2, Activation of Health Physics Facilities at Operations Support Center 5 147. EPIP 7.2.3, Activation of Health Physics Facilities at Technical Support Center 5 CHP Forms 148. CHP-02, Iodine Airborne Survey (pad of 50) 1 149. CHP-21, Miscellaneous Survey (pad of 50) 1 150. CHP-31, Radiation Work Permit (pad of 50) 1 151. CHP-34, Dosimeter Rezero (pad of 50) 1 152. CHP-37, Irregular or Offscale Dosimeter Report (pad of 50) 1 153. CHP-22, Air Particulate Sample (pad of 50) 1 154. CHP-25, Counting Log Sheet (pad of 50) 1 155. CHP-33b, Visitors Monitored per 10 CFR 20 (pad of 50) 1 156. CHP-33c, Visitor Personnel Monitoring Record (pad of 50) 1 157. CHP-35, Dosimeter Summary Sheet (pad of 50) 1 158. CHP-38, Lost or Damaged TLD Report (pad of 50)

~

1 159. CHP-39, Personnel Contamination Report (pad of 50) ~1 160. CHP-40, Visitor TLD Badge Issue Report (pad of 50) 1 160a. CHP-41, Occupational Exposure History 1 161. CHP-44, Timekeeping Log - High Radiation Work Location (pad of 50) 1 162. CHP-56, Personal Bioassay. Evaluation (pad of 50) 1

, 163. CHP-106, Occupational External Radiation Exposure History (pad of 50) 1 Emergency Plan Sampling Kits i

! 164. Emergency Plan Sampling Kits - Each kit contains I

the following: 2

1. Battery powered air sampler l'

2. Scott cartridge holder 1

3. Silver Zeolite cartridge holder 1

4. Stop watch with batteries 1

5. Air Particulate filters (env.) 1

6. Silver Zeolite cartridge 5

7. Scott charcoal cartridge 5

8. PIC-6A survey meter 1 lb,t

9. Water filled gae sample bottle (1 liter) 2 i EPIP-24a (05-83)

/

Page 6 Required On Hand

10. Liquid sample cubitainers (1 liter) 2

11. Scissors 1 pair

12. Plastic suit 2 sets

13. Gloves (surgeons) 6 pair

14. Dosimeters (0 - 5,000 mR) 2

15. Dosimeter charger 1

16. Plastic Bags 12 x 18 inch size 6 5x8 inch size 6 3x5 inch size 12

17. Flashlight with spare bulb and batteries 1

18. Smears (100 each/ box) 2

19. Tuck Tape (roll) 1

20. Sharpie, Flair pen, grease pencil and pencil 4

21. Sample ID tags (pad) 1

22. Sampling Procedures EPIP 7.3.1 Airborne Sampling and Direct Dose Rate 5 ea Survey Guidelines EPIP 7.3.1 Atmospheric Radioactive Iodine Sample 5 ea Attachment Collection and Counting

23. Sampling Forms 5 EPIP-01 Airborne Radiation Survey Record 5 ea EPIP-02 Emergency Plan Survey Record 5 ea Sample Identification Survey Map 5 ea 2 and 5 Mile Sample Identification Survey Map 5 ea -

24. CHP-34 Rezero Sheet 5 ea

25. 9 V batteries 2 ea

26. CS source 1 ea '

27. Clipboard 1 ea

28. Fuses 5 ea

29. Coin envelopes 1 pkg By Date Reviewed By Date (Health Physics Supervisor)

~.

4 5

EPIP-24a (05-83)

9 POINT BEACH NUCLEAR PLANT QUARTERLY EMERGENCY PLAN CHECKLIST DATE

Reference:

EPIP 7.4.1 - Routine Check, Maintenance, Calibration and Inventory of Schedule of Health Physics Emergency Plan Equipment EPIP 7.4.2 - Emergency Plan Equipment Routine Checks, Maintenance and Calibration Instructions SITE BOUNDARY CONTROL CENTER RESPIRATORY EQUIPMENT Item No. Tyge Serial No. Inspection

1. Full-face

2. Full-face

3. Full-face

4. Full-face ,

COMMUNICATIONS Portable Radio Functional check with control room WARNING LIGHTS, TRAFFIC Traffic Warning Lights All traffic warning lights functioning AC GENERATOR (Gasoline Powered)

Functional Test EPIP-25c (05-03)

DRY CELL BATTERY REPLACEMENT (For SBCC, Control Room, TSC, OSC, ESC and Two Rivers Hospital),

NOTE: If alkaline batteries are used, battery changeout is required annually rather than quarterly. If carbon or mercury batteries are used, a quarterly 5-minute test shall be completed to verify operability.

Item Battery Changed /

No. Type of Equipment Tgge ' Quantity Tested

Date Due

1. Traffic Warning Light i 2. Survey /Frisker

Instruments 1

3. Flashlights

4. Portable Radio __

5. Stop Watch

6. Dosimeter Charger CONTROL ROOM RESPIRATORY EQUIPMENT Item Type of Serial Functional No. Equipment 1 Number Inspection Test

1. Bio-Pak

2. Bio-Pak

3. MSA-SCBA

4. MSA-SCBA

5. Supplied Air Mask

6. Supplied Air Mask

7. Supplied Air Mask

8. Supplied Air Mask

9. Supplied Air Mask

10. Supplied Air Mask

11. Supplied Air Mask Hose .

12. Supplied Air Mask Hose

13. Supplied Air Mask Hose

14. Supplied Air Mask Hose

15. Supplied Air Mask Hose

16. Supplied Air Mask Hose

17. Supplied Air Valve

18. Supplied Air Valve

19. Supplied Air Valve

20. Supplied Air Valve

21. Supplied Air Valve

22. Supplied Air Valve

~ '.

EPIP-25c (05-83)

Control Room Respiratory Equipment, continued ...

Item Type of Serial No. Equipment Number Inspection

23. Supplied Air Hose

24. Supplied Air Hose

25. Supplied Air Hose 26: Supplied Air Hose

27. Supplied Air Hose

28. Supplied Air Hose

29. Spare Mask

30. Spare Mask

31. Spare Mask

32. Spare Mask

. 33. Bio-Pak

34. Bio-Pak

35. Bio-Pak

l. 36. Bio-Pak TECHNICAL SUPPORT CENTER /0PERATIONS SUPPORT CENTER RESPIRATORY EQUIPMENT Item Type of Serial Serial No. Equipment Number Inspection Number Inspection

1. Clear-Vue 1. 4.

2. -

5.

3. 6.

2. Ultra-Vue 1. 4.

2. 5.

3. 6.

3. Bio-Pak 1. 6.

2. 7.

3. 8.

4. 9.

5. 10.

COMMUNICATIONS Portable Radio Functional Test with (2 units) Control Room i

- )

EPIP-25c (05-33)

, , - - - , - , , - . , - --n,y .,-.. . - , , -

. ,re - --- r-w- w--- -+------? -,-w * - - - - + - - - - - - - ---7--e- - -- -----r

P EMERGENCY SUPPORT CENTER RESPIRATORY EQUIPMENT Item Type of Serial Serial No. Equipment Number Inspection Number Inspection 1 1. Clear-Vue 1. 4.

2. 5.

3. . 6.

2. Ultra-Vue 1. 4.

2. 5.

3. 6.

TWO RIVERS COMMUNITY HOSPITAL RESPIRATORY EQUIPMENT Item Tvpe of Serial No. Equipment Number Inspection

1. Clear-Vue 1.

2.

3.

4.

2. Ultra-Vue '

1. _

2.

3.

4.

REMARKS:

I l

NOTE: Include maintenance request numbers for all items requiring repairs.

Inventory By Date Reviewed By Date l

l l

l EPIP-25c

-: 3 i

(05-83) l

POINT BEACH NUCLEAR PLANT

}

SEMI-ANNUAL & ANNUAL EMERGENCY PLAN CHECKLIST DATE

Reference:

EPIP 7.4.1 - Routine check, Maintenance, Calibration and Inventory of Schedule of Health Physics Emergency Plan Equipment

~

EPIP 7.4.2 - Emergency Plan Equipment Routine Checks, Maintenance and Calibration Instructions SITE BOUNDARY CONTROL CENTER AIR SAMPLERS Item Preventive Flow Rate No. Tm Maintenance Calibration

1. High Volume (115 V AC)

2. DC Battery Powered i

DOSIMETERS Pocket Dosimeters Drift / Response Checked --

Date.Last Completed Date Due TLD's ,

TLD's Changed

Date Changed Date Due

Includes TLD's from emergency vehicle.

M4 EPIP-25d (05-83)

O RESPIRATORY EQUIPMENT i Item Serial Periodic No. Number Functional Test Maintenance

1. Bio-Pak EMERGENCY SUPPORT CENTER AIR SAMPLERS Item Preventive Flow Rate No. m T Maintenance Calibration

< 1. Low Volume (115 V AC) _

DOSIMETERS Pocket Dosimeters Drift / Response Checked Date Last Completed Date Due SOUTH GATE AIR SAMPLERS Item Preventive Flow Rate No. Equipment Maintenance Calibration

1. Low Volume CONTROL ROOM DOSIMETERS Pocket Dosimeters Drift / Response Checked Date Last Completed Date Due -

}

EPIP-25d (05-83)

RESPIRATORY EQUIPMENT Item Type of Serial Periodic No. Equipment Number Inspection Functional Test Maintenance

1. Bio-Pak

2. Bio-Pak

3. Bio-Pak

4. Bio-Pak

5. Bio-Pak

6. Bio-Pak

7. MSA SCBA N/A

8. MSA SCBA N/A

9. Supplied Air / N/A (Comp. Unit) / N/A

10. Supplied Air / N/A (Comp. Unit): / N/A

'/

11. Supplied Air / N/A (Comp. Unit) / N/A

12. Supplied Air / N/A (Comp. Unit) / N/A

13. Supplied Air / N/A (Comp. Unit) / N/A

14. Supplied Air / N/A (Comp. Unit) / N/A TWO RIVERS COMMUNIT'I HOSPITAL DOSIMETERS Pocket Dosimeters Drift / Response Checked Date Last Completed Date Due 74

'~ t I EPIP-25d (05-83)

f r

A s REMARKS:

Checked By Date Reviewed By Health Physics Supervisor r

.s t s

~

\

i EPIP-25d

- '3~*

(05-83)

-s s

i POINT BEACH NUCLEAR PLANT CALCULATION OF Xe-133 EQUIVALENT RELEASE RATES Responsibility - Shift Supervisor or designee.

Frequency - During classification only.

1.0 LOW RANGE OPERATIONAL VENT STACK READINGS Flow Rate Meter Readi'ng Conversion Factor Release Rate C

(c fm) (pCi/cc or cpm) (Curies /sec)

Auxiliary Building 61400 x 5.8 x 10','d =

-' Drumming Area 43100 x 1.3 x 10 8 =

Gas Stripper Building 13000 x 6.2 =

Combined Air Ejector 25 x 1.6 x 10 _2 =

/

2.0 EBERLINE RMS-II VENT STACK READOUTS Flow Rate Meter Reading Conversion Factor Release Rate (CFM) (R/hr) Curies - Hrs (Curies /sec) sec-R Auxiliary Building 61400 x 3.0 x 10 3 =

Drumming Area 43100 x 2.1 x 10 3 =

Gas Stripper Building 13000 x 6.2 x 102 =

Combined Air Ejector 25 x 3.4 =

O EPIP-34 (07-83 '

.x

.llt

L

.)

.i 3.0 PLANT EFFl.UENT VENT STACK CONTACT READINGS Flow Rate Meter Reading Conversion Factor Release Rate (CFM) (R/hr) Curies-hr (Curies /sec) sec-R Auxiliary Building 61400 x 3.0 x 10 2 =

prumming Area 43100 x 2.3 x 102 =

Gas Stripper Building 13000 x 8.0 x 104 =

Combined Air Ejector 25- x 1.6 x 10 2 =

o .

Estimated Steam x Specific x Conversion x Meter X Conversion Release Rate '

Release Volume F' actor Reading Factor = Ci/sec 3

(lb/hr) (ft /lbm) hr-cm3 (R/hr) Curies-hr sec-ft3 cm 3 -R Main Steam lleader x x 7.86 x x 8.0 x 10 1 = '

Assume 1000 psia steam which will give i a conservative specific volume. At 1000 psia specific volume = .446 fta/lbm.

Steam generator safety valve rating is 8.33 x 105 lb/hr per valve. Atmospheric 4

relief valve capacity is 3.3 X 10 lb/hr with both valves open.

EPIP-34 (07-83)J u

3

% 3 )

/

4.0 ESTlHATE OF GROSS Xe-133 EQUIVALENT RELEASE RATES Xe-133 Vent Equivalent Release Rate (Curles/sec)

Auxiliary Building Drunuming Area "

Gas Stripper Building Combined Air Ejector Decay Duct ,,

Main Steam Header TOTAL completed By , Date _ Time

, t L

O EPIP-34

( 07-8ll }

>

ML20077H413

Text

2.0 REFERENCES

2.0 REFERENCES

Background

2.0 REFERENCES

1.0 INTRODUCTION

Reference:

1.0 INTRODUCTION

1.0 INTRODUCTION

Reference:

Reference:

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