ML20053C451
ML20053C451 | |
Person / Time | |
---|---|
Site: | Point Beach ![]() |
Issue date: | 04/30/1982 |
From: | WISCONSIN ELECTRIC POWER CO. |
To: | |
Shared Package | |
ML20053C448 | List: |
References | |
PROC-820430-01, NUDOCS 8206020174 | |
Download: ML20053C451 (100) | |
Text
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TABLE OF CONTENTS Revision Date 1.0 CLASSIFICATION & ASSESSMENT 1.1 Initial Classification . . . . . . . . . . 2 04-30-82 1.2 Plant Status . . . . . . . . . . . . . . . . 0 03-31-81 1.3 Ettimation of Source Term . . . . . . . . . . 1 04-30-82 1.4 Radiological Dose Evaluation . . . . . . . . 4 04-30-82 1.5 Protective Action Evaluation . . . . . . . . 3 04-30-82 1.6 Radioiodine Blocking & Thyroid Exposure Accounting . . . . . . . . . . . . . . . . 1 02-26-82 1.7 Evaluation of Core Damage . . . . . . . . . . 0 04-30-82 1.8 Emergency Off-Site Dose Estimations . . . . . 0 04-30-82 2.0 UNUSUAL EVENT IMPLEMENTING PROCEDURES 2.1 Unusual Event - Immediate Actions . . . . . . 0 03-31-81 2.2 Unusual Event - Plant and Company Personnel Notification . . . . . . . . . . . . . . . 1 07-01-81 2.3 Unusual Event - Off-Site Agency Notification 1 02-26-81 V 3.0 ALERT IMPLEMENTING PROCEDURES 3.1 Alert - Immediate Actions . . . . . . . . . . 0 03-31-81 3.2 Alert - Plant & Company Personnel Notification . . . . . . . . . . . . . . . 1 07-01-81 3.3 Alert - Off-Site Agency Notification . . . . 0 03-31-81 4.0 SITE EMERGENCY - IMPLEMENTING PROCEDURES 4.1 Site Emergency - Immediate Actions . . . . . 0 03-31-81 4.2 Site Emergency - Plant & Company Personnel Notification . . . . . . . . . . . . . . . 1 07-01-81 4.3 Site Emergency - Off-Site Agency Notification 0 03-31-81 5.0 GENERAL EMERGENCY - IMPLEMENTING PROCEDURES 5.1 General Emergency - Immediate Actions . . . . 0 03-31-81 5.2 General Pnergency - Plant & Company Personnel Notification . . . . . . . . . . 1 07-01-81 5.3 General Emergency - Off-Site Agency Notification . . . . . . . . . . . . . . . 0 03-31-81 6.0 EVACUATION 6.1 Limited Plant Evacuation . . . . . . . . . 1 04-30-82 g 6.2 Plant Evacuation . . . . . . . . . . . . 1 02-26-82 6.3 Exclusion Area Evacuation . . . . . . . . . 0 03-31-81 6.4 Energy Information Center Evacuation . . . 0 03-31-81 8206020174 820525 PDR ADOCK 05000266 F. PDR
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\m,/ 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 i
Boundary Control Center . . . . . 2 03-17-82 7.2.2 Activation of HP Facilities at Operations Support Center . . . . 1 03-17-82 7.2.3 DELETED
, 7.2.4 Health Physics Communications . . . 1 03-17-82 s 7.2.5 Control & Use of Vehicles . . . . . 1 03-17-82 7.3 Radiological Surveys 1
7.3.1 Airborne Sampling & Direct Dose Rate Survey Guidelines . . . . . . 3 03-17-82 7.3.2 Post-Accident Sampling & Analysis
[ of Potentially High Level Reactor Coolant . . . . . . . . . . . . . 3 12-30-81 7.3.3 Post-Accident Sampling of Contain-ment Atmosphere . . . . . . . . . 3 12-30-81 7.3.4 Movement of Required Chemistry Equip-ment & haterial to the Technical Support Center Counting Room &
Mini-Laboratory . . . . . . . . . 0 12-30-81 7.4 Emergency Equipment 7.4.1 Routine Check, Maintenance, Cali-bration & Inventory Schedule for Health Physics Emergency Plan Equipment . . . . . . . . . . . 5 04-30-82 7.4.2 Emergency Plan Equipment Routine Check, Maintenance & Calibration h Instructions . . . . . . . . . . . 3 04-30-82
'~) 7.4.3 Use of Baird Model 530 Single Channel Iodine Spectrometer to Determine Airborne Iodine Activities . . . . . . . . . . 1 05-15-81 l
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d' Page 3-N 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 . .. . 2 04-30-82 9.1 Security . .. ... . ... .. . ... . . 0 03-31-81 l 10.0 Firefichting . ...... . .. . ... .. 0 03-31-81 11.0 FIRST AID & MEDICAL CARE 11.1 On-Site First Aid Assistance .. . . .. . . 2 02-26-82 11.2 Injured Person's Immediate Care . . ... .. 1 05-15-81 11.3 Hospital Assistance . ... .. .. ... . . 1 04-30-82 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
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Teams . . .. ... ... . ... ... . . 2 04-30-82 12.3 Recovery Planning . . . .. . .. . ... .. 0 03-31-81
., 12.4 Personnel Monitoring Expo 3ure Guidelines . . 0 01-29-82 13.0 PRESS 13.1 Crisis Communications . . . . .. . ... .. 1 09-04-81 14.0 COMMUNICATIONS 14.1 Testing of Communications Equipment . ... . 0 03-31-81 15.0 IRAINING, 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 16.0 WISCONSIN ELECTRIC GENERAL OFFICE PROCEDURES 16.1 Nuclear Engineering Section Notification &
Response ..... . . .. . . . .. .. . 3 09-04-81 b
04-30-82 TABLE OF EPIP FORMS EPIP EPIP Form Title Procedure 01 Emergency Plan Airborne Radiation Survey Record Site Boundary Control Center (03-81) 7.3.1 02 Emergency Plan Survey Record Site Boundary Control Center (09-81) 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 (03-81) 1.2 05 Worksheet for Status Report on Radiation Monitoring System for Unit (03-81) 1.2 06 Worksheet for Status Report on Radiation Monitoring System for Plant (03-81) 1.2 07 For X/Q Determination (09-81) 1.4 08 Estimated Whole Body & Thyroid Projected Doses (09-81) 1.4 09 Estimated Whole Body Dose Calculation Parksheet for Specific Noble Gas Releases (09-81) 1.4 i 10 Estimated Ground Deposition Calculation Worksheet for Particulate Radionuclide Releases (09-81) 1.4 v 11 Summary of Radiological Dose Evaluation Calculations (09-81) 1.4
, 12 Unusual Event Incident Report Form (03-81) 2.1 4
13 Alert Incident Report Form (03-81) 3.3 14 Site Emrgency Incident Report Form (03-81) 4.3 15 General Emergency Incident Report Form (03-81) 5.3 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 (02-82) Call List Tab 23 Offsite Agency Emergency Call List (02-82) Call List Teb 24a Site Boundary Control Center Emergency Plan Inventory Checklist (04-82) 7.4.1 24b TSC, ESC, South Gate & OSC Emergency Plan Inventory Checklist (04-82) 7.4.1 24c Emergency Plan Health Physics Supplies at Two Rivers Community Hospital Inventory Checklist (09-G1) 7.4.1 24d Control Room Emergency Plan Equipment Inventory Checklist (09-81) 7.4.1 24e Emergency Vehicle Inventory Checklist (04-82) 7.4.1 24f Emergency Plan First Aid Kit Inventory Checklist (02-82) 7.4.1 24g Emergency Plan Burn Kit Inventory (02-82) 7.4.1 24h Emergency Plan First Aid Room Inventory (05-81) 7.4.1 p/ 24i Emergency Plan Stretcher Inventory (09-81) 7.4.1
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EPIP Forms Title EPIP Procedure 25a Emergency Vehicle . Checklist (10-81) 7.4.2 25b Monthly Health Physics Instrument .i Air Sampler Functional Test Checklist (04-82) 7.4.2 25c Quarterly Emergency Plan Checklist (04-82) 7.4.2 25d Semi-Annual & Annual Emergency Plan Checklist (04-82) 7.4.2 26 Quarterly Communications Test (03-81) 14.1 27 Monthly Communications Test (03-81) 14.1 28 Emergency Plan Instrument Calibration Schedule (05-81) 7.4.2 29 Emergency Plan Counting Equipment & Frisker Calibration Schedule (07-81) 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 (04-82) 1.7 -
34 Calculation of Xe-133 Equivalent Release Rates (04-82) 1.8 35 Dose Calculations (04-82) 1.8
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EPIP 1.1 MINOR Revision 2
% ,/ 04-30-82 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 appropriately 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.
l This condition could possibly lead to a degradation in overall safety.
This condition does not represent a significant radioactivity release, (h
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involves no offsite response, and may require no augmentation of plant personnel. In spite of the above, prompt notification of the.
counties and state is required.
1.2 Alert l
l Plant conditions in which events are in progress or have occurred which involve an actual or potential degradation of plant safety.
l Radiation releases are not likely to cause an offsite hazard. Prompt
! offsite notification is necessary and the plant organization may have to be augmented.
1.3 Site Emergency l
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.
1.4 General Emergency l
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 l
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v ) releases may have an impact on offsite people. Pronipt notification
EPIP 1.1 Page 2 O
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 evacu-ation may be necessary. The emergency news center will be opened.
The Shift Supervisor has the responsibility and authority to take humediate 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, Revision 1, published November, 1980.
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 notifications 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 fire 50 gpm greater than letdown flow when in fact an Alert should be declared.
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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4 EPIP 1.1 Page 3 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 System Functions 1
- 2. Abnormal Primary Leak Rate 1
- 3. Abnormal Coolant Temperature /
Pressure 2
- 4. Abnormal Primary / Secondary Leak 2
- 5. Core Fuel Damage 3
- 6. Secondary Coolant Anomaly 4
- 7. Abnormal Effluent 5
- 8. Major Electrical Failures 5
- 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 l 16. Hi.zards to Plant Operations 9
- 17. Natural Events 10
- 18. Reactivity Transient 10
EPIP 1.1 Page 4 v
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.
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 s
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j NOTE: IF THE SHIFT SUPERVISOR CANNOT COMMUNICATE WITH A DUTY &
CALL SUPERINTENDENT, THE SHIFT SUPERVISOR MUST NOTIFY THE STATE AND TWO COUNTIES WITHIN 15 MINUTES OF THE DECLARATION OF ANY CLASS OF EMERGENCY.
i 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, parform EPIP.3.1,
" Alert - Immediate 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 e General Emergency, perform EPIP 5.1, " General Emergency - Immediate Actions."
NOTE:
"Gae 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 DSS - Duty Shift Supervisor TABLE l-1 FFDSAR - Final Facility Description &
i Safety Analysis Report MASP - Modified Amended P8NP Security Plan EMERGENCY CLASSIFICATION 1
Emergency Category 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-mation noted:
- 1. " Containment press hi",
i lPI-945, PI-947, PI-949 (2/3 >5 psig)]
- 2. " Steam line loop A lo-lo press"
[PI-468, PI-469, PI-482 (2/3 <530 psig)]
- 3. " Steam line loop B lo-lo press"
[PI-478, PI-479, PI-483 (2/3 <530 psig)]
- 4. " Pressurizer lo press SI"
[PI-429, PI-430, PI-431 (2/3 (1735 psig)]
- 5. Wide range pressure-<l500 psig i Loss of containment integrity requiring When shutdown commences as determined by DSS Unusual Event .
shutdown by Technical Specifications and DCS i
Loss of engineered safety feature requiring When shutdown commences as determined by DSS Unusual Event l shutdown by Technical Specifications and DCS l
1 Loss of fire protection system function When shutdown comumences as determined by DSS Unusual Event ,
i requiring shutdown by Technical Specifi- and DCS cations (i.e., both fire pumps inoperable)
- 2. Abnormal Primary Exceeding Technical Specification primary system When shutdown commences as determined by DSS Unusual Event Leak Rate leak rate (10 gun) and DCS
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EPIP 1.1 Table 1-1 Page 2 Emer gency Category Initiating Condition Indication Used Classification Leak rate >50 gpm All of the following: Alert
- 1. " Volume control tank level bi-lo"
[LI-141 and/or LI-ll2 <8%)
- 2. Decreasing pressurizer level lL1-426, LI-427, LI-428]
- 3. Charging ptanp speed hi"
- 4. Charging line flow (F1-120) >50 gpm more than letdown flow (FI-134)
Leak rate in excess of available pump All of the following: Site Emergency capacity including charging, high head SI
- and low head SI 1. " volume control tank level hi-lo lLI-141 and/or LI-ll2 <8%]
- 2. No pressurizer level indicated
[LI-426, LI-427, LI-428]
- 3. All available pumps running as indicated by 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 Coolar.t Unexpected decrease in subcooling margin Both of the following: Unusual Ev?nt Temperature / Pressure
- 1. Alarm on P-250, if operable
- 2. Confirmation by manual calculation Pressure >2735 psig Pressure >2735 psig on PR-420 and NRC only
" Code, safety or PORV not closed" l-hour open lins (2)
DNBR <l.30
- 4. Abnormal Primary /- Exceeding Technical Specification priaary-secondary When shutdown comme'ces as determined by DSS Unusual Event Secondary Leak leak rate and DCS
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- i EPIP 1.1 Table 1-1 Page 3 Emergency l Ca tegory Initiating Condition Indication Used Classification Gross failure of 1 SG tube (>400 gpm) & loss of All of the following first-out reactor panel Elert ,
offsite power (FFDSAR 14.2.4) annunciation with confirmation indication:
1
- 1. " Pressurizer Lo Press SI," !
l lPI-429, PI-430, PI-431 (2/3 <!735 psig)l
- 2. a. " Steam generator A level hi"
[LI-461, LI-462, LI-463 (2/3 >70%)) or
- b. " Steam generator a level hi" (
[LI-471, LI-472. LI-473 (2/3 >70%)l i
- 3. a. d4.16 kv bus undervoltage" l
& 0 volts on A03 & A04 voltmeters,
- b. XO4 to A03 asumeter on CO2 (0 amps)
' c. XO4 to A04 ammeter on CO2 (0 aura)
- 4. SI flow >400 we indicated by FI-924 & j F1-925 and pump discharge pressure l corresponding to flow. j Rapid failure of >10 SG tubes (4000 gpm) with T of the folivaing first-out reactor panel 3 Site Emergency or without offsite AC annunciation with confirming indication 1
- 1. " Pressurizer lo press SI"
[PI-429, PI-430, PI-431 (2/3 <1735 psig)l
- 2. a. " Steam generator A level hi" lLI-461, LI-462, LI-463 (2/3 >70%))
or
- b. " Steam generator B level hi" (LI-471, LI-472, LI-473 (2/3 >70%))
- 3. SI flow >4,000 gpm indicated by FI-626 &
FI-928.
- 5. Core Fuel Damage Gross fuel damage in core indicated Both of the following: Unusual Event
- 1. Letdown line radiation monitor (R9) 100 x alarm setpoint.
- 2. Sustained offscale & chemical analysis shows fission product concentration increase by 100X.
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EPIP 1.1 Table 1-1 Page 4 Emergency Category Initiating condition Indication Used Classification Massive fuel damage 300 pCi/cc iodine-equivalent as determined by Alert by chemical analysis
- 1. Massive loss of fuel clad integrity Initiating Condition Nos. 1 & 2 exist and No. 3 General Emergency
- 2. With simultaneous loss of primary system integrity is possit,le
- 3. With potential loss of contairment integrity chemical analysis
- 2. Primary system leak >1000 p m indicated by SI flow >1000 g o (FI-924 & FI-925) and pump discharge pressure corresponding to flow
- 3. Miniasm contairment pressure suppression equipment is not available (any of the following):
- a. No fan cortlers operating and <2 spray Pumps.
- b. No spray pumps operating and <2 fan coolers
- c. <2 fan coolers running with I spray pump
- 4. " Containment press hi"
[PI-945. PI-947. PI-949 (2/3 >5 psig)]
- 5. " containment spraya with 2/3 + 2/3 >25 psig
[PI-945, FI-947, PI-949] '
[PI-946, PI-948, PI-950)
- 6. Secondary Coolant Reduction in feedwater enthalpy incident 1. a. Decreasing feedwater temp indicated by Unusual Event
, Anomaly (FFDSAR 14.1.7) TO-418A & 70-438A on P-250 g
- b. confirmed by local temperature indicator on outlet of No. 5 feedwater heater.
- 2. Unexpected increasing power on excore nuclear +
instrumentation Steam line break with primary-to-secondary leak l A l of the following first-out reactor trip Alert rate in excess of 10 go panel annunciation with confirmation:
(FFDSAR 14.2.5)
- 1. Either:
- a. " Steam line loop A Lo-Lo press" (PI-468, PI-469, PI-482 (2/3 <530 psig)]
or
- b. " Steam line loop B Lo-Lo press"
[PI-478, PI-479, PI-483 (2/3 <530 psi-)
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EPtr 1.1 Table 1-1 Fage 5 Emergency Category Initiatir.g Condition Indication Used Classification
- 2. Confirmed primary-to-secondary leak rate of at least 10 gpe.
- 3. Either:
- a. " Steam line loop A isol channel alert"
[FI-464, FI-465 (1/2 >3.9x10' lb/hr)] EE
- b. " Steam line loop B isol chana:1 alert"
[FI-474, FI-475 (1/2 >3.7x10s ib/hr)) Secondary Coolant Transient initiated by loss o' feedwater, followed All of the following: General Emergency Anomaly by loss of auxiliary feedwater for >l hour (FFDShR 14.1.11) 1. Decreasing SG levels -
"A" SG [LI-461, LI-462, LI-463] *B" SG lLI-471, LI-472 LI-473)
- 2. No auxiliary feedwater flow -
lFI-4002, F1-4007, FI-4014] lFI-4036, F1-4037]
- 7. Annormal Effluent Radiological efflues.t Technical Specification limits Airborne effluents only Unusual Event exceeded but <10 times the limit (FFDSAR 14.2.3)
Radiological effluent Technical Specification limits Liquid effluents only Unusual Event exceeded (FFDSAR 14.2.2) Radiological effluents >10 times Technical Airborne effluents only Alert Specification instantaneous limits. (An instan-taneous rate which, if continued for >2 hours, would result in a dose of about I mR at the site boundary under average meteorological conditions.)
- 8. Major Electrical Sustained loss of offsite power >l5 minutes All of the following : Unusual Event Failures (FFDSAR 14.1.2)
- 1. "4.16 kv bus undervoltage"
& 0 volts on A03 & A04 voltmeters.
- 2. XO4 to A03 ammeter on'CO2 (0 amps).
- 3. XO4 to A04 ammeter on CO2 (0 amps) ,
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EPIP 1.1 Table 1-1 Page 6 Emergency Category Initiating Condition Indication Used Classification Sustained loss of onsite AC power capability Both of the following: Unusual Event (>15 minutes)
- 1. "4.16 kv bus undervoltage" &
O volts on A05 and A06 voltmeters and
" Emergency Diesel Starting System Disabled
- for both Diesels
, Loss of all vital onsite DC power >l5 minutes Both of the following: Site Emergency ] 1. " Annunciator 5swer failure" on 4 C01, CO2, C03, and C04
- 2. <100 volts on the voltmeters for all batteries Loss of offsite power and loss of all onsite AC All of the following: Site Emergency power for >l5 minutes
- 1. "4.16 kv bus undervoltage" 0 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 ammettr on CO2 (0 amps)
Loss of offsite and all onsite AC power with loss er All of the following: General Emergency all auxiliary feedwater for >2 hours
- 1. Unit aux W meter -XO2 on CO2 (0 W)
- 2. Station aux W meter XO4 on CO2 (0 W )
{ 3. XO4 to A03 ammeter on CO2 (0 amps) , 4. XO4 to A04 amater on CO2 (0 amps) , 6. XO2 to A01 ammeter on CO2 (0 amps)
- 7. a. No auxiliary feedwater flow lFI-4036, FI-4037]
- b. Decreasing SG level -
"A" SG [LI-461, LI-462, LI-463] "D" SG [L1-471, LI-472, LI-473]
- 9. Control Room Evacuation Evacuation of control room >l5 minutes & As required by DSS Site Emergency no control at remote shutdown station
- 10. Fire Fire in vital area or on the controlled side of plant As reported by Fire Brigade Chief Unusual Event lasting >10 minutes after initial use of fire extinguishing equipmenc.
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EPIP 1.1 Table 1-1 Page 7 Emergency Category Initiating Condition Indication Used classification ] -- ' Fire affecting 1 train of safety systems. As reported by Fire Br,igade Chief Alert Fire affecting 2 trains of safety systems As reported by Fire Brigade Chief Site Emergency Unusual Event
- 11. Plant Shutdown Function Nonfunctional indications or alarms in the control Both of the following:
room on primary system parameters requiring plant ,
- 2. Failed indication as determined by DSS.
Turbine mechanical failure with consequences 1. Annunciator " Turbine supervisory." Unusual Event
- 2. Indication on TR-6019 of bearing vibration
>7 mils.
- 3. Bearing vibration alarm on back of C03.
- 4. Visual confirmation of turbine housing penetration by a blade or disc.
i Significant loss of effluent monitoring capability & l. Loss of LW16 during a release Unusual Event meteorological instruments which impairs ability to or perform emergency assessment. Loss of effluent 2. Loss of R18 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 R14 and RMS II channel 1 or
- c. Loss of R15 and CR9 and RMS 11 Channel 5 or
- d. Loss of R21 and RMS II channel 2 or
- e. Loss of GWil2 and RMS 11 Channel 6 Alert Failure of reactor protection system to All of the following:
complete a trip which brings reactor suberitical Unplanned first out annunciator on C04 with confirmation from associated indicator and intermediate range detector output not decaying and >l RCC RPI indicates fully withdrawn
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L) . EPIP 4.1 Table 1-1 Page 8 Emergency Category Initiating Condition Indication Used Classification 4 All alarms (annunciatcrs) lost >l5 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 Any of the following: Alert hears while at cold shutd m
- 1. Loss of service water Unit 1 = south & west header .
Unit 2 = north & west header
- 2. Loss of both trains of RHR
- 3. Loss of component cooling
- 12. Abnormal Radiation a. Effluent monitors detect levels corresponding to Airborne effluents only Site Emergency Levels at Site any of the follewing:
(1) >50 mR/hr for \ hour (2) >250 mR/hr for 4 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 HP Supervisor
- c. Any of the dose rates projected,'ased o on plant parameters
- a. Effluent monitors detect levels corresponding to Airborne effluents only General Emergency either:
(1) 1 R/hr whole body (2) 5 R/hr thyroid at the site boundary under actual meteorole gical conditions
- b. Either of the above doses measured in environs As reported to DSS by HP Supervisor
- c. Either of above dose rates proiected based on other plant parameters
- 13. Fuel Handling Accident Major damage to irradiated fuel in containment Both of the following: Alert
- 1. As reported to DSS by Core Loading Supvr.
- 2. Alarm on Victoreen on manipulator & alarm on Rt!
N . (p) j v N (m N.,
- EPIP 1.1 Table 1-1 Page 9 Emergency Indication Used Classification Category initiating Condition Both of the following: Alart Fuel damage accident with release of radioactivity to auxiliary building (FFDSAR 14.2.1)
- 1. As reported to DSS by Supvr in charge of fuel handling & drumming area vent (R21)
- 2. Alarm on Victoreen on spent fuel pit bridge.
Reported as judged by DSS Unusual Event
- 14. Serious or Fatal Injury Transportation of seriously or fatally injured individual from site to hospital (expect hospitalization for at least (Reference EPIP 11.1) 48 hours)
Per NA5P Per MASP & Appendices
- 15. Security Threat Security threat or attempted sabotage 1-Hour Red Phone only
- or (open Line) (4)
Ongoing security compromise Visual observation of Operations Supervisor or Unusual Event
- 16. Hazards to Plant Unusual aircraft activity over facility security force Operation As reported to DSS by plant personnel making Unusual Event Near or onsite explosion or flammable or toxic gas release
- 1 observation Missile impacts from any source on facility Visual observation by Operations Supervisor Alert Missile impact causing damage to two trains of Site Emergency safety systems visual observation by Operations Supervisor A'ircraft crash in protected area (within the fence) Visual observation by operations Supervisor Alert Known explosion damage to facility affecting plant Visual observation by operations Supervisor Alert operation. Toxic or flammable gases in facility environment excluding normal process gases Visual observation by Operations Supervisor Site Emergency Toxic or flammable gases entering into vital areas (control room, auxiliary building, etc.) excluding normal process gases
r i R
\, \s x .
EPIP 1.1
- Table 1-1 Page 10 Emergency Category Initiating Condition Indication Used Classification
- 17. Natural Events Any earthquake Activation of >2 accelerographs and verified by actual physical ground shaking or by con-tacting Dr. David Willis. University of Wisconsin, Milwaukee Seismic Center at 1-414/963-4602. Unusual Event Any tornado visible from site Verification by Operations Supervisor Unusual Event Low Lake Michigan water level With no CW pumps running, water level is 3.9' Unusual Eveat below O' on surge chamber level & confirmed by measuring forebay level at 10.9' below pumphouse floor (7' level)
Earthquake greater than operating basis earthquake Earthquake with attendant structural damage of Alert containment or spent fuel pit Any tornado striking the facility Visual observation by operations Supervisor Alert Seiche near design level >6" of water in turbine hall Alert Winds in excess of design levels Wind speed indicated as >100 mph Alert Wind with damage Structural damage to containment Site Emergency Failure of protection for vital equipment at low Any of the following: Site Emergency le;vels (i.e., caused by seiche > design levela)
- 1. >3' water in both EDG rooms.
- 2. >2' water in vital switchgear room.
- 3. >2' water in auxiliary feed pump room.
- 18. Reactivity Transient Uncontrolled rod withdrawal (FFDSAR 14.1.1 & 14.1.2) Unusual Event CVCS Malfunction (FFDSAR 14.1.5) Unuaual Event Accidental Criticality NRC Only (3)
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iPIP 1.1 Table 1-1 Page 11 Emergency Category Initiating Condition Indication Used Classification
- 19. Load Transient Loss of Electrical Load (FFDS&R 14.1.10) Unusual Eve.at
- 20. Other condition that warrants State and/or local official DCS & DSS concurrence unusual Event awareness Condition that warrants establishment of technical DCS & bJS concurrence Alert support center & emergency support center l
Condition that warrants use of monitoring teams DCS & DSS concurrence Alert Personnel contamination Health Physicist & DCS concurrence NRC-only 1-hour (10) Any unplanned reactor trip DCS & DSS concurrence NRC-only 1-Hour (7) Strike by employees or guard force DCS & DSS concurrence NRC-only 1-Hour (~2) 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 1-Hour (!!)
4 EPIP 1.3 MINOR Revision 1 [) v 04-30-82 ESTIMATION OF SOURCE TERM 1.0 GENFRAL 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 (ABVNT) 1.2 Drumming Area Vent (DAVNT) 1.3 Unit 1 Containment Purge Vent (CONT 1) 1.4 Unit 2 Containment Purge Vent (CONT 2) 1.5 Gas Stripper Building Vent (GSBVNT) 1.6 Combined Air Ejector Decay Duct (CAE) 1.7 Main Steam Safety Valves and Atmospheric Dump Valves 2.0 REFERENCE 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 (R9) may be unusually high or offscale. 3.1.2 The containment radiation monitors (R11 and R12) may be unusually high or offscale. 3.1.3 The containment area monitors (R2 :.nd R7) may be unusually high or offscale.
- 3.1.4 The charging pump area monitor (R4) may be unusually high or offscale.
pd
EPIP 1.3 Page 2 V. 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 ESTIMAIE - 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 (f-~ the 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. 5.1.2 Enter the meter readings and flow rates in the appropriate columns on Worksheet No. I 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 obtain meter readings of the vents whose Eberline RMS II data is not available and the main steam header by performing Section 5.2 of this procedure if required. NOTE: IF STEP 5.1.3 NEEDS TO BE COMPLETED BECAUSE EBERLINE RMS II DATA IS NOT AVAILABLE, OR IF A STEAM GENERATOR TUBE RUPTURE IS BELIEVED TO HAVE OCCURRED WHICH PRODUCES THE POTENTIAL FOR RELEASES, OR RELEASES ARE IN PROGRESS FROM THE MAIN STEAM HEADER OR THE ATMOSPHERIC STEAM DUMP, THEN PERFORM SECTION 5*.3 0F s,) THIS PROCEDURE AFTER APPROPRIATE MEASUREMENTS HAVE BEEN TAKEN IN SECTION 5.2. 5.1.4 Perform Section 5.3 of this procedure to determine the gross Xe-133 equivalent release rate estimate.
EPIP 1.3 Page 3 a 5.2 Direct Stack Survey Team Designees NOTE: THE FOLLOWING SECT.!0N WILL NOT BE INITIATED UNTIL THE EVALUATION DISCUSSED IN PRECAUTION 3.3 HAS BEEN COH?LETED AND THE SITE MANAGER (DUTY & CALL SUPERINTENDENT), THE LUTY & 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. ' (/3 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 Chemistry / Health Physics Supervisor or Designated Alternate 5.3.1 Choose the appropriate vent stack readouts in Part A, B, or C of Worksheet No. 1 to convert readings to a Xe-133 equivalent release rate. That is if the low range monitors go 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 inoperable, use the vent stack contact readings. 5.3.2 Use the appropriate attached conversion curves for each of the plant effluent vent to convert the chosen vent stack readout, (cpm or R/ hour) and flow rate, from Step 5.3.1 to an Xe-133 equivalent release rate in Curies /second and record the value on Worksheet No. 1, Part D, Estimate of Gross Xe-133 Equivalent Release Rate. Enter the appropriate word "offscale," "not monitored," or " inoperable" for the i cases where the plant effluent vent was not monitored, s) offscale, or inoperable.
=
EPIP 1.3 Page 4 O NOTE: THE FOLLOWING QUALIFYING NOTES MUST BE RECOGNIZED.
- 1. If the actual flow rate is different than the conversion curves flow rate, a ratio of:
Actual Flow Rate Conversion Curve Flow Rate should be applied to determine the release rate. (Ratio) X Conversica Curve = Adjusted Xe-133 Release Rate
= Equiv. Release Rate
- 2. If the main steam header vent release rate needs to be determined, the following steps must be applied.
- a. Obtain from the shift Supervisor an estimated flow rate through the main steam header in Ibm / hour of steam being dumped to the environ-ment and the specific volume (v) of the steam. ,
At 1000 psia, specific volume is 0.446 ft.3/lbm. A, At 500 psia, specific volume is 0.928 ft.3/lt.n.
}
c lbm/hr x v x 7.86 ft eb
- b. Convert contact reading obtained at the main steam header to pCi/cc using the appropriate attached conversion curve for the main steam header.
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 (7) on Worksheet No. 1, Part D, to determine the gross Xe-133 equivalent release rate. O
i . EPIP 1.3
; Page 5 O NOTE: IF GRAB SAli?LE 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 IN 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. I e f 4 1 4 . f 1 i )
,.___c~-.-.,_-.n --.,,,--_,.,__,,.,_n,, ,, - , , , - - - ,,,, .n. ,. - - - . . _ - . _ , , , - - . , . _ ,- ,-,, ,, ,,,
WORKSHEET NO. 1 Xe-133 EQUIVALENT RELEASE RATE A. LOW RANGE OPERATIONAL VENT STACK READOUTS 1 Meter Reading Flow Rate Conversion Curve Vent (cpm) (cfm) Attachment No. Auxiliary Building 61400 1.3-1 Drumming Area 43100 1.3-2 Unit 1 Containment Purge 12500/25000 1.3-3 and 1.3-4 Unit 2 Containment Purge 12500/25000 1.3-5 and 1.3-6 Gas Stripper Building 13000 1.3-7 Combined Air Ejector Decay 1.3-8
.e g B. EBERLINE RMS - II VENT STACK READOUTS Meter Reading Flow Rate Conversion Curve Vent (R/ hour) (cfm) Attachment No.
Auxiliary Building 61400 1.3-9 , Drumming Area 43100 1.3 i Unit 1 Containment Purge 125C0/25000 1.3-11 and 1.3-12 Unit 2 Containment Purge 12500/25000 1.3-11 and 1.3-12 l Gas Stripper Building 1.3-13 l Combined Air Ejector Decay 1.3-14 i i J. . ~ n J l
. - . - . . - . , , _ + - . . . ~ , . _ _ . . , , , _ . . - . , _ , . . . , , . . - _ _ _ . . . . _ _ - . . . . , . - . _ - . _ . . - - -
4 4 O C. . PLANT EFFLUENT VENT STACK CONTACT READINGS Meter Reading Flow Rate Conversion Curve Vent (mr/hr or R/hr) (cfm) Attachment No. Auxiliary Building 61400 1.3-15 Drumming Area 43100 1.3-16 Unit 1 Containment Purge 12500/25000 1.3-17 and 1.3-18 Unit 2 Containment Purge 12500/25000 1.3-17 and 1.3-18 Gas Stripper Building 13000 1.3-19 Combined Air Ejector Decay 1.3-20 Main Steam Header 1.3-21 D. ESTIMATE OF GROSS Xe-133 EQUIVALENT RELEASE RATE O Xe-133 Equivalent Release Rate
'- Vent (Curies /Sec.)
- 1. Auxiliary Building i 2. Drumming Area
- 3. Unit 1 Containment Purge
- 4. Unit 2 Containment Purge
- 5. Gas Stripper Building
- 6. Combined Air Ejector Decay Duct
- 7. Main Steam Header
- 8. Sum (Gross Xe-133 Equiv. Release Rate)
QR
- 9. Grab Sample Results = Ci/sec.
Completed By Time Date
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4 EPIP 1.4 MINOR (g Revision 4 Q 04-30-82 RADIOLOGICAL DOSE EVALUATION 1.0 GENERAL The purpose of this procedure is to provide a method to quickly estimate (1) X/Q using meteorological overlays, (2) thyroid and whole body dose using X/Q and (3) ground deposition using an approximation of D/Q.
2.0 REFERENCES
2.1 U. S. NRC Regulatory Guide 1.109, Calculation of Annual Doses to Man from Routine Release of Reactor Effluents for the Purpose of Evaluating Compliance with 10 CFR Part 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. See Appendix D " Technical Bases for Methods that Estimate the Projected (bT Thyroid Dose and Projected Whole Body Gamma Dose from Exposure to k _,) Airborne Radioiodines and Radioactive Noble Gases." 2.3 U. S. NRC Regulatory Guide 1.4, " Assumptions used for Evaluating the Potential Radiological Consequences of a Loss-of Coolant accident for Pressurized Water Reactors,d Revision 2, June 1976. 2.4 TID 14844, " Calculation of Distance Factors for Power and Test Reactor Sites," March 23, 1962. 3.0 PRECAUTIONS & LIMITATIONS 3.1 Ensure that all PBNP maps to be used by this procedure and their corresponding meteorological overlays are based on the same scale. 3.2 This procedure will be accomplished in the technical support center by a person designated by the Shift Supervisor or the Technical Support Manager. It will usually be done in conjunction with the Chemistry / Health Physics Supervisor when available. 3.3 This procedure will also be accomplished in the ESC by a person designated by the RadCon/ Waste Manager. 3.4 Wind speed and wind direction must be average values obtained from the analog recorders in the control room. , O NOTE: DO NOT USE INSTANTANEOUS VALUES. U
EPIP 1.4 Page 2 f'i G 3.5 If the radiological release duration is unknown, assume a duration of 8 hours. 3.6 If the meteorological parameters cannot be obtained from the control room, obtain the data from the following priority backup list. 3.6.1 Kewaunee Nuclear Power Plant. 3.6.2 National Weather Service in Green Bay. Ask for Two Rivers Coast Guard information, if available. 4.0 INITIAL CONDITIONS 4.1 A release of airborne radioactivity has occurred or a release is anticipated. 4.2 An emergency or potential emergency condition is anticipated to have offsite dose consequences. 5.0 PROCEDURE 5.1 Determination of X/Q. Atmospheric Dispersion Factor , r-.) I 5.1.1 Obtain the following information from the indicated source
'~' and enter this data in the appropriate space on form EPIP-07 (attached).
Source Data EPIP-04 (a) wind speed in-mph EPIP-04 (b) wind direction EPIP-04 (c) time of reactor shutdown EPIP-04 (d) time of release to containment EPIP-04 (e) time of release from the plant Health Physics or Operating (f) duration or expected duration logs or projected estimate of the release in-hours (see EPIP 1.3 results note) (g) gross Xe-133 equivalent release rate in Ci/sec NOTE: IF RELEASE DURATION IS UNKNOWN, ASSUME 8 HOURS. 5.1.2 Visually check cloud cover and incoming solar radiation. With this information, use Attachment 1.4-1 to ascertain the appropriate stability class. Enter the stability class (h), on form EPIP-07. NOTE: IF INCOMING SOLAR RADIATION IS STRONG AND WINDS ARE FROM THE EAST OR SOUTHEAST, IT IS A POSSIBILITY THE O, WIND IS PRODUCED BY A LAKE EFFECT. CALL THE GREEN BAY NATIONAL WEATHER SERVICE FOR AID IN THIS DETER-MINATION. IF A LAKE BREEZE IS SUSPECTED, OFFSITE SURVEY TEAMS MUST BE REMINDED TO PAY CLOSE ATTENTION TO WIND DIRECTION.
EPIP 1.4 Page 3
-~s 5.1.3 Place the overlay corresponding to the stability class on the map. Using the plant location as a pivot point, align the centerline of the overlay to the downwind direction from the plant.
NOTE: THE " TICK" MARKS ON THE CENTERLINE OF THE OVERLAYS ARE ONE MILE APART. s.l.4 Determine the distance (i) to the dose projection location if different from the standard centerline distances listed on form EPIP-07. Note the location description, sector, and distance on form EPIP-07. Enter the Xu/Q value (j) for the distances of site boundary, two miles, five miles, and ten miles on EPIP-07. The Xu/Q values (j) can be obtained from the overlay in the table in the lower righthand corner of the overlay. If a possible location other than the standard specified location is on a line, enter the Xu/Q (j) value for that line from the overlay on form EPIP-07. If the location is not on a line, move to the next inner-most line (toward the centerline) and
. enter the Xu/Q (j) value for that line on form EPIP-07.
r'N s_,/ a Example: 6 m 2 Class "C" Xu/Q for 5 miles equals 1.21 x 10 5.1.5 Calculate the X/Q value from the Xu/Q value by using the equation: X sec.
= 2.24(sec./m) X WindXu/Q (m 2)
Speed (mi/hr.) Q m" (hrs./mi) Enter the X/Q values on form EPIP-07. 5.2 Whole Body Dose Estimate NOTE: IF THE NOBLE GAS SOURCE TERM IS DETERMINED BY GRAB SAMPLE RESULTS WHICH GIVES AN INVENTORY OF SPECIFIC NUCLIDES, THEN A CONSERVATIVE WHOLE BODY DOSE ESTIMATE CAN BE MADE BY COMPLETING FORM EPIP-09. 5.2.1 Enter the gross Xe-133 equivalent release rate (g) on form EPIP-08 from form EPIP-07. 5.2.2 Enter the expected inhalation period, EIP, in hours (f) on form EPIP-08 from form EPIP-07.
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EPIP 1.4 Page 4 s 5.2.3 Calculate the projected whole body dose on form EPIP-08 by using the equation: (k) (g) 3 3 D(Rem) = X/Q (sec/m ) x Q (Ci/sec) x Kr (Rem m /Ci - Hrs) x EIP (Hrs) where: D = whole body dose (Rem) X/Q = atmospheric dispersion coefficient determined in Step 5.1.5 (sec/m3 ) (k) Q = release rate (Ci/sec) (g) I
" 3 Kr=DoseFactor([*ihs) Attachment 1.4-2 EIP = Expected Inhalation (Exposure) Period (Hours) (f) b 5.3 Thyroid Dose Estimate 5.3.1 Calculate the projected thryoid dose by using the whole body dose calculated in Section 5.2 of this procedure.
5.3.2 Record the projected whole body dose on form EPIP-08 in Section 2. 5.3.3 Choose the appropriate figure based upon the type of accident which has occurred.
- a. Loss of Coolant Accident (LOCA) - Figure 1.4-1.
- b. Gap Activity Accident - Figure 1.4-4.
- c. Fuel Handling Accident - Figure 1.4-4.
- d. Steam Generator Tube Rupture - Figure 1.4-5. ,
NOTE: IF THE TYPE OF ACCIDENT IS UNKNOWN, USE THE LOCA FIGURES. l 5.3.4 Obtain the ratio factor + hat relates the whole body dose to a thyroid dose from the figure chosen with the corresponding appropriate time after the accident and record on form l ,-_s EPIP-08, Section 2. ( ) 5.3.5 Calculate the projected thyroid dose my multiplying the N/ whole body dose by the ratio factor obtained in Step 5.3.4 on form EPIP-08, Section 2.
4 EPIP 1.4 Page 5 V} I 5.4 Radionuclide Ground Deposition Estimation NOTE: FORM EPIP-10 CAN BE COMPLETED ONLY IF IODINE GRAB SAMPLE RESULTS OR PARTICULATE RELEASE RATES ARE AVAILABLE. IF FORM EPIP-10 CANNOT BE COMPLETED, PROCEED WITH STEP 5.4.5 0F THIS SECTION. 5.4.1 Enter the Xe-133 equivalent release rate or the specific particulate release rate on form EPIP-10 from grab sample results or from environmental monitoring results. 5.4.2 Enter the duration of release expected inhalation period (f) from form EPIP-07 on form EPIP-10. 5.4.3 Enter the value of X/Q (k) on form EPIP-10 as determined in Step 5.1.5 5.4.4 Complete Section 2 of form EPIP-10 to calculate the ground deposition using the equation: Dep (pCi/m 2 ) = F x .05 (m/sec) x 3600 (sec/hr) x 10 6 (pci/Ci) x X/Q (sec/m 3) . x Q (Ci/sec) x EIP (hrs) lm i (k) d x X/Q x (g) (f) Dep = F x 1.8 x 10 Q x EIP I Dep = ground deposition (pCi/m3 ) X/Q = atmospheric dispersion factor from Step 5.1.5 (sec/m 3) (k)
- Q = radionuclide release rate (Ci/sec) (g)
EIP = estimated release duration (hrs) (f) F = fraction of isotope subject to deposition (unitless) 3600 = conversion (sec/hr) 106 = conversion (pCi/Ci) 0.05 = assumed deposition velocity (m/sec) 5.4.5 Complete form EPIP-11 from available data and calculations just performed.
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EPIP 1.4 , Page 6 5.4.6 Enter the date and time of these calculations and sign form EPIP-11. 5.4.7 Forward completed attachments to the Technical Support Manager for review. The Technical Support Manager will relay results to the Site Manager. f .I 4 i
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ATTACHMENT 1.4-1 DETERMINATION OF ATMOSPHERIC STABILITY CLASS Surface Wind Speed, Day Night (at 50 meters) Incoming Solar Radiation Thinly Overcast mph Strong Moderate slight > 1/2 low cloud < 1/2 cloud
<4 A A-B B 4-7 A-B B C E F 7-11 B B-C C D E 11-13 C C-D D D D .. >13 C D D D D The neutral class D, should be assumed for overcast conditions during day or night. " Strong" incoming solar radiation corresponds to a solar altitude greater than g- 60* with clear skies; " slight" incoming solar radiation corresponds to a solar s'- g altitude from 15*-35* with clear skies. Cloudiness will decrease incoming solar J radiation and should be considered along with solar altitude when determining solar radiation. Incoming radiation that would be strong with clear skies can be expected to reduce to moderate with broken (5/8 to 7/8 cloud cover) middle clouds and to slight with broken low clouds. Night refers to the period from one hour before sunset to one hour after sunrise.
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n s-l EPIP 1.5 MINOR I Revision 3 04-30-82 PROTECTIVE ACTION EVALUATION f 1.0 PURPOSE The purpose of this pro <:edure is to provide a basic guide to determine ) protective action recommendations to be given to the public authorities and j to provide a method to transmit these recommendations and other essential I data for assessment to the appropriate public authorities.
2.0 REFERENCES
2.1 NUREG-0654, Revision 1, " Criteria for Preparation and Evaluation of Radiological Emergency Response Plans and Preparedness in Support of Nuclear Power Flants," November, 1980. 2.2 NUREG-0654, Appendix 1, " Emergency Action Level Guidelines for Nuclear-Power Plants." f 3.0 PRECAUTIONS AND LIMITATIONS 3.1 Ask for the name and title of the person or agency being contacted i prior to transmitting any information. l 3.2 If unable to contact an individual or agency, continue with the transmi::sions to the other individuals or agencies and then attempt to contact the persons or agencies who have not been contacted. 3.3 All actions and recommendations should be appropriately logged. 3.4 If the radiological release duration is unknown, assume a duration of 8 hours for use during an evaluation of the need for a protective action recommendation. 3.5 When protective action recommendations are made, consider the recom-mendation over a 90* sector centered on the average wind direction and a full 360* area near (2 miles) the plant. 4.0. INITIAL CONDITIONS 4.1 Applicable portions of EPIP 1.2, " Plant Status," completed. 4.2 EPIP 1.3, " Estimation of Source Term," completed. 4.3 EPIP 1.4, " Radiological Dose Evaluation," completed. 4.4 Site Emergency or General Emergency has been declared.
EPIP 1.5 Page 2 5.0 PROCEDURE 5.1 Technical Support Manager 5.1.1 Obtain the completed attachments of EPIP 1.4, " Radiological Dose Evaluation," from the person completing them. 5.1.2 Review the results of the dose projection calculations and deposition calculations for all affected areas. 5.1.3 Review Attachments 1.5-1 and 1.5-2. 5.1,4 Based on actual plant conditions, expected plant conditions in the future, weather conditions, local protection avail-able to the public, evacuation times and any other constraints, determine the most appropriate Protective Actions to reduce exposure to the public and relay the information to the emergency support center. 5.2 Emergency Support Manager NOTE: THE FOLLOWING STEPS MUST BE DONE BY THE EMERGENCY SUPPORT (Q' y MANAGER OR HIS DESIGNATED ALTERNATE. UNTIL HE ARRIVES IN THE EMERGENCY SUPPORT CENTER, THE SITE MANAGER IS ACTING AS EMERGENCY SUPPORT MANAGER. 5.2.1 Review the recommendation of the Technical Support Manager and/or Rad / Con Waste Manager. 5.2.2 Complete Section 2 (follow-up message) of the incident report form contained in the appropriate offsite agency notification procedure, (for example, if the incident is classified as a Site Emergency, then Section 2 of the Inci-dent Report Form of EPIP 4.3, " Site Emergency - Offsite Agency Notification," would be completed). 5.2.3 contact the NRC and the persons and agencies notified on NAWAS of the emergency and provide the information contained in Section 2 of the incident report form to them. 5.2.4 For a General Emergency, form EPIP-16 in EPIP 5.3, " General Emergency - Offsite Agency Notification," shall be used as the basis for followup messages to offsite technical per-sonnel such as NSSS vendor and corporate engineering staff. O O
Cv b v V - ATTACHMENT 1.5-1 Recommended protective actions to reduce whole body and thyroid dose from exposure to a gaseous plume Projected Dose (Rem) to Individual in General Public Recommended Action (# Comments Whole body <1 No planned protective actions.(b) Previously recom.nended or State may issue an advisory to seek protective actions may Thyroid <5 shelter and await further instructions. be reconsidered or terminated. Whole body 1 to <5 Seek shelter as a minimum. If constraints exist, special or Consider evacuation. Evacuate unless consideration should be given Thyroid 5 to <25 constraints make it impractical. for evacuation of children Monitor environmental radiation levels. and pregnant women. Whole body 5 and above Conduct mandatory evacuation. Seeking shelter would be an or Monitor environmental radiation levels alternative if evacuation Thyroid 25 and above and adjust area for mandatory evacuation were not immediately possible. based on these levels.
, Control access.
(a)These actions are recommended for planning purposes. Protective action decisions at the time of the incident must take into consideration existing conditions and the dangers associated with certain protective actions. (b)At the time of the incident, officials may implement low-impact protective actions in keeping with the principle of maintaining radiation exposure as low as reasonable achievable.
Reference:
Abstracted from EPA 520/1-75-001, " Manual of Protactive Actions Guides and Protective Actions for Nuclear Incidents," Table 5.1 (Revised 6/79) l
1 N [] J ~ ATTACHMENT 1.5-2 L) .
SUMMARY
OF PERSONNEL IOSE PATE/PIOJECTED (10SE COMMt1HE7tT ACTION LIMITS De*clare Evacm te R/hr - Dose Rate Action Lireits e , Consider Limited or Plant se an Em Wency R - Projected N se Commitment Action Limits 2 m e l Ev.scuation - Monitor and Control Evacuation s Workers w x o8 ax m
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EPIP 1.7 MINOR Revision 0 04-30-82 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 letdown radiation monitor (R9) may be unusually high or offscale. 3.1.2 The containment radiation monitors (R11 & R12) may be unusually high or offscale. 3.1.3 The containment area monitors (R2 & R7) 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 radiological 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. (v) v
EPIP 1.7 Page 2 O V 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 analysi~s. 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 the following conversion factors. f Time Since Shutdown (Hr) Ci/ml per R/hr (Conversion Factor) (O
\s_,) 1 2.31 x 10"2
_ _2 8 5.93 x 10 _1 ', 24 1.77 x 10 _1 - 148 4.24 x 10 _1 720 2.05 x 10 5.4 Interpolate conversion factors for times between those values listed.
- 55. Enter the conversion factor from Section 5.3 on form EPIP-33.
5.6 Determine the estimated Sample Activity using the equation: Estimated Sample Activity (Ci/ml) = i Sample Bomb Contact Reading * (R/hr) x Conversion Factor Ci/ml R/hr
- Contact reading is on shielded sample bomb which incorporates 3 inches of external solid lead shielding.
I 5.7 Enter the estimated Sample Activity on form EPIP-33. ( ) I l ' l l l 1
e EPIP 1.7 Page 3 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 hours - 0900 hours = 8 hours Ci ml 2.75 R/hr x 5.93 x 10 p
= 1.63 x 10- Ci/ml 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 lf k 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 Spray additive tank: 2,574 gallons 7.1.3 Correction Factor for Time Since Shutdown [CF(t)] hours Omd
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