ML20207S284
| ML20207S284 | |
| Person / Time | |
|---|---|
| Site: | Indian Point  |
| Issue date: | 03/13/1987 |
| From: | James Anderson, Quinn D POWER AUTHORITY OF THE STATE OF NEW YORK (NEW YORK |
| To: | |
| Shared Package | |
| ML093440686 | List: |
| References | |
| PROC-870313, NUDOCS 8703190237 | |
| Download: ML20207S284 (66) | |
Text
indian Point 3 Nuclear P1wer Plant P.O. Box 215 Buchanan. NowWrk 10511 914 739.8200 A NewYorkPbwer 4# Authority DtERGENCY PLAN PROCEDURES PROCEDURE NO. IP-1028 REv.
3 TITLE "
CORE DAMAGE ASSESSMENT H
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WRITTEN BY:
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REVIEWED BY:
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Ao DATE PORC REVIEW:
e APPROVED BY: /[M(
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8703190237 870313 ~
PDR ADOCK 05000286 EFFECTIVE DAT :
COPE DAMAGE ASSESSMENT 1.0 PURPOSE To provide a methodology to determine the extent of core damage following an accident. The assessment is based on radionuclide concentrations and other parameters.
2.0 DISCUSSION 2.1 The core damage procedure is based on quantitative and qualitative assessments of various plant parameters, some of which are interrelated.
2.1.1 Radiation monitors in the VC: This is a gross but immediately available measurement of noble gases released.
2.1.2 Radioactivity released from the core:
Measured in RCS, VC atmosphere, sumps in VC.
l.
Correct by power history, decay, etc.
Evaluate versus expected radioactivity released for clad damage, fuel overheat, fuel melt.
2.1.3 Hydrogen in containment: A measure of the amount of the zirconium water reaction from the fuel cladding.
2.1.4 Core exit thermocouples and reactor vessel level instrumentation: When available, used to determine whether the core has been uncovered and what type of fuel damage may have occurred.
2.2 Qualitative Assessment of Core Damage. Qualitative Assessment of Core Damage, should be used in conjunction with the quantitative assessment of core damage which follows in Section 3.0.
2.3 Limitations of this Procedure This procedure is an approximate method and may give some conflicting results. Engineering judgement must be used throughout.
Some areas for potential errors are:
Plateout of samples in containment or in sample lines; Gamma spectroscopy of highly radioactive samples; Estimates of ECCS water volumes or sump volumes; Calculations of core inventories; Effect of multiple precursors in the parent-daughter decay chains and unequal release fractions.
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IP-1028/3 The uncertainties are such that core damage estimates using this methodology are syfficient only to establish rajor categories of fuel damage. This categorization, with confirmation, will require extensive additional analysis for some several days past the accident date.
3.0 PROCEDURE
QUANTITATIVE CALUCLATIONS OF CORE DAMAGE The calculations can be performed by following the attachments as worksheets or by running the computer programs.
Instructions for the program are listed in Attachment 4B.
3.1 Data Collection 3
Using Attachment 2, record all appropriate data concerning RCS, sumps in VC, and VC atmosphere sampling.
Power History - Note if computer is being used, the power history data has already been entered; follow the prompts in the program.
Record EFPD and calendar days of operation on Attachment 4
}
Part 1.
If reactor has been at steady state power (i 10% of average power level) for 4 days or more, record power level on, Part 2.
If reactor has not been at steady state power for at least 30 days, use Attachment 4A to record power history over the last 30 days.
Record sample results from RCS, Recire. Sump, VC atmosphere on Attachments 5 and 6.
A 3.2 Power History Correction Factor (PCF)
The inventories of fission products shown in Attachment 3 are for end of core life 100% power steady state operating conditions. This must be corrected for actual power history.
3.2.1 Steady state power (1 10%) prior to shutdown:
Long-lived nuclide correction is calculated in
- , Part 3.
Short-lived and medium-lived nuclide corrections are calculated in Attachment 4, Part 4.
3.2.2 Transient Power History:
Long-lived nuclides:
transient power history not applicable, use Attachment 4 to calculate power correction factor.
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IP-1028/3 Short and medium-lived nuclides: correct each nuclide separately using Attachment 4A as a calculation sheet.
Computer-calculated power correction factors are also available. They may be accessed by using Attachment 4B.
3.3 Chemistry Sample Corrections Samples of RCS, Recire. Sump, and VC atmos'here must be corrected p
using Attachment 5 (Water Samples) and Attachment 6 (VC Atmosphere).
3 3.3.1 Back-Decay Correction CF (bd)
This factor is used to correct the sample result back to the time of reactor shutdown.
Nuclides that are daughters in a chain must be accounted for by following the calculations in A.
The Chemistry computer has the capability to back-decay nuclides. Ensure that this correction is not applied twice. The daughters as discussed above should not be back-decayed by the Chemistry computer.
3.3.2 Temperature-Pressure Correction CF (tp)
This factor is used to account for the differences in temperature and pressure between 'he sample and the t
sampled system (e.g., RCS, VC air).
Water samples are corrected for temperature only.
Air samples are corrected for both temperature and pressure.
3.4 Calculation of Percent Core Damage The calculation of percent core damage involves 3 basic steps:
Determination of activity released from the core; Determination of the power corrected activity inventory; Comparison of the actual activity released to the expected inventory.
This calculation is performed for clad damage, fuel overheat, and fuel melt using Attachments 7, 7A, 7B, 7C, and 7D.
3.4.1 Calculate total activity released by radionuclide:
Using Attachment 7 as a calculation sheet, add the activity from RCS, containment sumps, and VC atmosphere to determine total activity released from the core.
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u-IP-1028/3 Values for activity concentrations are obtained from Attachments 5 and 6, and should have been previously corrected for decay, dilution, temperature, pressure, etc.
in accordance with Attachments 5 and 6.
3.4.2 Calculate activity normally present in the RCS during operations:
Using Attachment 7A as a calculation sheet, determine the amount of each nuclide present during normal operations.
This activity is subtracted from the total amount released from the core.
This calculation is only used in assessing clad damage.
For other types of fuel damage, it is an insignificant fraction of the activity.
3.4.3 Calculate Percent Fuel Damage:
Use Attachments 7B (Clad Damage), 7C (Fuel Overheat),
and 7D (Fuel Melt) as calculation sheets.
Correct the nuclide inventories from Attachment 3 using the previously developed power correction factors.
Compare the activity released (Attachments 7, 7A) to the corrected inventories to obtain percent fuel damage.
3.5 Assessment of Core Damage using Activity Released Assessment of core damage involves determining:
The type of core damage: clad damage, fuel overheat, fuel melt.
The amount of core damage: 0 to 100% in each of the above categories.
3.5.1 Comparison with expected inventories released: lists the nuclides associated with the 3 types of fuel damage and the amount of activity expected to be released for 100% clad damage, 100% fuel cverheat, and 100%
fuel melt.
3.5.2 The nuclides released are characteristic of the type of damage as are the ratios of nuclides.
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i l
ZP-1028/3 3.5.3 Clad Damage:
Nuclides associated with cladding damage are primarily the medium-lived and long-lived noble gases and iodines.
- B contains the calculated percent clad damage.
The ratios of the noble gases to Xe-133 (and Iodines to I-131) in the gap differ from the ratios in the fuel-itself. The ratios are shown in Attachment 3 and can help to accertain whether the release was from the fuel (fuel overheat or melt) or from the gap (clad damage).
RCS pressure, temperature, and power transients may result in Iodine spiking where the Iodine concentrations in the RCS increase sharply. This is not indicative of cladding failure but should be considered so that it is not confused with clad damage. provides an estimate of the total I-131 release that might be expected during an iodine spike.
Clad rupture is dependent on fuel temperature and RCS pressure where higher RCS pressures will delay clad rupture.
3.5.4 Fuel Overheat:
Moderately volatile fission products are released during fuel overheat conditions, including cesium, ruthenium, and tellurium in addition to the more volatile noble gases and iodines. Lesser amounts of barium and strontium are also released.
- C provides the calculated percent fuel overheat.
The use of the isotopic ratios listed in Attachment 3 can be used to determine the source of the noble gases and iodines.
3.5.5 Fuel Melt:
Fuel pellet melting leads to rapid release of noble gases, iodines, bromines, and cesiums remaining af ter fuel overheat.
Significant release of the strontium, barium-lanthanum chemical groups is the most distinguishing feature of fuel melt conditions.
- D provides the calculated percent fuel melt.
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IP-1028/3 1
The use of isotopic ratios listed in Attachment 3 can be used to determine the source of the noble gases and iodines.
3.5.6 Non-Uniform Core Damage:
The above evaluations address an assumed uniform distribution of core damage. The degree of damage may vary within the core, and this should be considered in explaining any conflicting data.
4.0 AUXILIARY INDICATORS There are plant indicators monitored during an accident which can provide verification of the initial estimate of core damage based on the radionuclide analysis. The plant indicators include containment hydrogen concentration, core exit thermocouple temperatures, reactor vessel water level, and containment radiation level.
4.1 Containment Hydrogen Concentration An accident in which the core is uncovered and the fuel rods are exposed to steam may result in the reaction of the zirconium of the cladding with the steam which produces hydrogen. It is assumed that all of the hydrogen that is produced is released to the containment atmosphere.
The hydrogen dissolved in the primary system during normal operation contributes an insignificant amount of the total hydro ~ gen released to the containment. The hydrogen recombiners will not have a significant effect on a zirconium - steam reaction in the case of severe core degradation.
The percentage of zirconium water reaction does not equal the percentage of clad damaged but it does provide a qualitative verification of the extent of clad damage estimated from the radionuclide analysis. shows the relationship between the hydrogen concentration and the percentage of zirconium water reaction.
4.2 Core Exit Temperatures and Reactor Vessel Rater Levels Core Exit Thermocouples (CETC) and the Reactor Vessel Level Indication System (when available) (RVLIS) readings can be used for verification of core damage estimates in the following ways.
Due to the heat transfer mechanisms between the fuel rods, steam, and thermocouples, the highest clad temperature will be higher than the CETC readings. Therefore, if thermocouples read greater than 1300*F, clad failure may have occurred.
1300*F is the lower limit for cladding failures.
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ZP-1028/3 If any RCPs are running, the CETCs will be good indicators of clad temperatures and no core damage should occur since the forced flow of the steam-water mixture will adequately cool the Core.
No generalized core damage can occur if the core has not uncovered.
So if RVLIS full range indicates that the collapsed liquid level has never been below the top of the core and no CETC has indicated temperatures corresponding to superheated steam at the corresponding RCS pressure, then no generalized core damage has occurred. 0 provides information on types of damage to fuel at increasing temperatures.
4.3 Containment Radiation Levels R-25 and R-26 are located just above the 95' VC and can be used as a gross indication of activity (primarily noble gases) in the containment atmosphere.
R-25 and R-26 would be expected to read approximately the same value if there were noble gases dispersed in containment. 1 provides data on expected radiation levels for clad damage, fuel overheat, and fuel melt conditions.
5.0 REFERENCES
References are listed in Attachment 12.
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IP-1029/3 LIST OF ATTACHMENTS ATTACHMENT TITLE 1
Qualitative Assessment of Core Damage 2.
Sampling Data for Core Damage Calculations 3
Core Release Inventories of Characteristic Fission Products 4
Power Correction for Core Inventories - Steady State 4A Power Correction for Core Inventories - Transient Conditions 4B-Instructions for Use of CORDAM Computer Program 4C Results from Computer Program Test Case 5
Water Sample Data and Calculations 3
5A Parent-Daughter Decay Correction 6
VC Atmosphere Sample Data and Calculations 7
Calculation of Total Activity Release from Core 7A Calculation of Activity Present During Normal Operations 7B Calculation of Percent Clad Damage 7C Calculation of Percent Fuel Overheat 7D Calculation of Percent Fuel Melt 8
Expected Iodine Spike vs. Normal Iodine Activity 9
VC Hydrogen Concentration vs. Zirconium-Water Reaction
~
10 Expected Fuel Damage Correclation with Fuel Rod Temperature 11 Expected Containnent Radiation Levels Post-Accident (R-25/R-26) 12 References e
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~
3-IP-1028 ATTACHMENT 1 QUALITATIVE ASSESSMENT OF CORE DAMAGE NO DAMAGE CLAD DAMAGE FUEL OVERHEAT FUEL MELT Radiation Levels 1 R/hr.
Up to Up to Up to in VC *
(Minimum 32 R/hr.
8.3E4 R/hr.
1.4E5 R/hr.
(R-25 & R-26)
Reading) 3
% Hydrogen 0%
<C Up to 12.6%
in VC **
Core Exit 600*F 1300*F and check temperature vs. pressure Thermocouples for super heated core uncovery.
l RVLIS (if Full Used in conjunction with CETCs to determine available) ***
core uncovery.
Expected Kr. Xe, I Kr. Xe, I Cs, Te Sr, Ba, La, Pr Nuclides Time dependent R-25/R-26 readings can be found in Attachment 11.
Presence of hydrogen is indicative of reaction of the cladding but does not indicate whether fuel overheat or melt has occurred.
No generalized core damage can occur if the core remains covered.
l
IP-1028 ATTACHMENT 2 SAMPLING DATA FOR CORE DAMAGE CALCULATIONS Calculation No.:
Current Date:
Reactor Shutdown:
Date:
Current Time:
Time:
SAMPLE AND MEDIA DATA RCS VC ATMOSPHERE RECIRC SUMP OTHER 3
Sample No.
Date of Sample Time of Sample Sample Temperature (*F)
- F
'F
- F Sample Pressure (psia) psia psia psia System Temperature (*F)
- F
'F
'F System Pressure (psia) or Level psia psia ft l3 (see a. below)
Volume of ECCS Dilution Water:
gallons a.
Level in Recire. Sump:
ft gallons b.
Level in VC Sump:
ft gallons f
c.
Level in Reactor Sump:
inches gallons I
(Note:
This can only be estimated based on the 2)" level alarm, the 6" level alarm, and the reactor sump pump light; Technical Support Center should estimate this volume if necessary.)
1 d.
Level in Containment:
inches gallons above 46 ft, elevation i
Total estimated gallons of water in VC sumps and floor:
gallons e.
i l
l i
IP-102G ATTACHMENT 3 CORE RELEASE INVENTORIES OF CHARACTERISTIC FISSION PRODUCTS FUEL FUEL FUEL DECAY CAP OVERHEAT MELT PELLET ** CAP **
RELEASE RELEASE RELEASE ACTIVITY ACTIVITY CONSTp)
(DAY' (Ci)
(Ci)
(Ci)
RATIO RATIO NUCLIDE HALF-LIFE Clad Failure Nuclides Kr-85*
10.72 yr 1.77E-4 1.6E4 9.0E5 1.5E6
.01
.11 Kr-87 76.3 m 1.31E+1 3.1E3 1.8E7 3.0E7
.22
.022 Kr-88 2.84 h 5.86E0 6.7E3 2.5E7 4.2E7
.29
.045 Xe-131m 11.84 d 5.85E-2 7.5E2 2.8E5 4.7E5
.004
.004 Xe-133 5.245d 1.32E-1 1.5ES 8.8E7 1.5E8 1.0 1.0 I-131 8.04 d 8.62E-2 2.4E5 4.3E7 7.2E7 1.0 1.0 4
I-133 20.8 h 8.00E-1 1.6ES 8.8E7 1.5E8 2.1
.71 I-135 6.61 h 2.52E0 8.3E4 7.9E7 1.3E8 1.9
.39 Fuel Overheat Nuclides Cs-137*
30.17 y 6.3E-5 N/A 4.9E6 8.1E6 Te-129 69.6 m 1.4E+1 N/A 1.5E7 2.4E7 Te-132 78.2 h 2.1E-1 N/A 6.2E7 1.0E8 1
Fuel Melt Nuclides Ba-140 12.79 d 5.4E-2 N/A 2.2E5 3.5E7 La-140 40.22 h 4.1E-1 N/A 2.5ES 3.7E7 La-142 95.4 m 1.1E+1 N/A 1.9ES 3.1E7 Pr-144 17.28 m 5.8E+1 N/A 1.5ES 1.4E6 Long-lived nuclides.
Ratio for Noble Cases is to Xe-133 = NG Isotope /Xe-133.
Ratio for Iodines is to I-131 = I Icotope/I-131.
IP-1028 ATTACHMENT 4 POWER HISTORY CORRECTION FOR STEADY STATE POWER HISTORY 1.
Data for Long-Lived Nuclides Power Correction Factor.
EFPD Calendar Days Start Date of the Current Cycle Oldest Fuel Cycle (A)
Previous Cycle Current Date (B) 2 Cycles Previous Days Between (A)&(B)
=
2.
Data if Plant has been at Steady-State Power (within 10% of average power level).
Steady State Power Level (last 4 days) =
Steady State Power Level (last 30 days) =
3.
Calculation of Long-Lived Power Correction Factor.
Nuclide Half-Life EFPD/ Calendar Days Long-Lived Kr-85 10.72y Nuclides Cs-137 30.17y 4.
Calculation of Short and Medium-Lived Power Correction Factor -
(Steady State Operation).
Steady State Power Level (%)
Nuclide Half-Life (last 4 days): P(4)
P( 4)/100%
Short-Lived Kr-87 76.3 m Nuclides Kr-88 2.84 h I-133 20.8 h I-135 6.61 h Te-129 69.6 m La-142 95.4 m Pr-144 17.28 m Steady State Power Level (%)
Nuclide Half-Life (last 30 days): P(30)
P(30)/100%
Medium-Xe-131m 11.84 d Lived Xe-133 5.245d Nuclides I-131 8.04 d Te-132 78.2 h Ba-140 12.79 d La-140 40.22 h P( 4)/100%
NOTE:
Short-Lived Power Correction Factor (PCF)
=
P(30)/100%
Medium-Lived Power Correction Factor (PCF)
=
EFPD/ Calendar Days Long-Lived Power Correction Factor (PCF)
=
ZP-1028 ATTACHMENT 4A POWER HISTORY CORRECTION FOR NUCLIDE i (Transient Power History)
Nuclide:
Half-Life:
A=
day" (from Attachment 3) j j
j Power Duration (Days)
-A
-A Period Level (%)
(Days)
Decay Time (1-e i j)
(e'A Uj)
P)(1-e i j)(e-U)j 1
2 3
4 5
=.
7 8
9 10 E
(1-e-i j)(e-i j) =
~E I )(1-e-1 j)(e j)
PCF
=
=
1 100%
p) steady reactor power level (percent)
=
~
A decay constant for isotope 1 (day )
=
g T) time at power level P (days)
=
time since end of T to reactor shutdown (days) t)
=
NOTE: Power history should cover the last 30 days or more.
s-IP-1028 ATTACHMENT 4B INSTRUCTIONS 10R USE OF COMPUTER PROGRAM FOR CALCUI.ATIONS 1.
Use a half-duplex, 300 or 1200 baud terminal. Parity setting is irrelevant. Dial one of the following numbers:
(
.)
(
)
3 1.1 Place telephone into acoustic coupler or turn on modem (as appropriate) when tone is heard.
1.2 The computer will send some characters to the terminal. Hit the return key.
1.3 When the terminal requests " SERVICE", type 65 and hit the return key.
1.4 When the terminal requests " USER NUMBER", type P489PFM, (password)*
and hit the return key.
2.
Type -PCF and hit the return key.
(Allow about 30 seconds for the program 3
to run.) A complete Power Correction Factor report will be sent to the Date and time given should be the time of trip or commencement of screen.
reactor shutdown. If this is not the case, inform the Site Reactor Engineer or the Performance and Reliability Supervisor to update the shutdown file.
3.
First you must run a test case of the core damage assessment computer program. Type -CORTEST and hit the return key. The program will automatically execute using the input data specified in ENG-208, Rev. 1,
" Acceptance Test for Core Damage Assessment Computer Program". To ensure the program is functioning correctly, compare the output of the program to the test results listed in Attachment 4C.
4.
Type -CORDAM and input the data requested. Note that the printout obtained in Step 2 will be used as part of the input section to this program.
CORDAM will complete all calculations and output Attachments 5 through 7D.
5.
Type " BYE" and hit the return key.
6.
Hang up the telephone and disconnect the equipment.
Password available from Site Reactor Fngineer, Performance and Reliability' Supervisor, or Control Room.
IP-1028 ATr>OMNT 4C 11Sr CASE OF ATTADMNT 7D CAIIIIATI(N OF PERONT FUEL PELT (A)
(B)
(C)
(D) = (B)x(C)
(E)
(F)
(C)
ICF Corrected (E/D x 100%)
theorrected Ibel < Pbwer Correctics Factor Ibel Melt Activity Released Perrent NG or Icxifne Nuclide Release Inventory (C1)
(Attachamt 4 or 4A)
Inventory (C1)
From Core (C1)*
Phel Melt Ratios **
Kr-85 1.5En 2.430F-OI 3.645E405 1.479E408 4.058E+04 6.464E-01 Kr-87 3.E7 9.915E-Ol
' 2.975 9 07 8.450E+11 2.841E406 3.692803-4.168F407 6.360E+09 1.526E404 2.779E+01 Kr-88 4.2E7 9.923E-OI I
Xe-I31m 4.7ES 9.810E-01 4.611E+05 1.178E408 2.555E+04 5.149E-0? __
Xe-133 I.5FB 9.998E-Ol 1.500F408 2.288s08 1.526E402 1.000END I-131 7.2E7 9.948E-Ol 7.1632107 1.051E+08 1.467E402 1.(YWJF+00
/
I-133 1.5FB 9.993E-01 1.499E408 1.522E+08 1.016E+02 1.449E+00
' T.29'.E402 4.056ERO a-135 1.3is 9.949E-Ol 1.293E408 4.261E+08
~
i l
cs-137 8.11%
2.430E-01 1.968E+06 6.997E+07 3.555E403 N/A.
Te-129 2.4F7 J.000E400 2.w oi'+07 1.094909 4.560E403 N/A 5.812E407 5.812901 N/A Te-132 1.0iB l.000EHOD 1.0COF+08 _
Ba-140 3.5E7 1.tTATr.iOO 3.500E407 4.223E407 1.2Nf402 N/A
- )
la-140 3.7E7 1.0u0E+00 3.700E407 3.153&O7 8.522E401 N/A la-142 3.lE7 1.000&OO 3.100E407 1.878E+10 6.058E404 N/A Pr-144 1.4FE 1.00DR00 1.400F+06 1.785E406 1.275E+02 N/A 1
s..
i j
sg h
p
'd -
Y?~
l2%
v.
L f y' a
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ATTACHMENT 5 1
WATER ShMPLE'0ATA AND CALCULATIONS Water Sampl9.TypI:
Sample No.:
r hRCS l
lRecireSum[> '
l lOther:
, B)
(C)
(b)
(E)
(F)
(G)
(
(A) a ecay Temperature CbnsNnt Reported Correction Correction Corrected Nuclide H51f-Life x(hr_g) uCi/cc Factor Factor uCi/ gram t.
- Kri 5 10.72 y 7.38 E-6
'3
, Kr-87 76.3 m 5.45 E-1
$ Ki 2.84 h 2.44 E-1 Ke-131m* 11.84 d, 2.44 E-3 7e-133*
5.245d 5.51 E-3 (1-131
'S.40 4 3.59 E-3 I-133 20.8 h 3.33 E-2 I-135 6.61 h 1.05 E-1 Cs-137 30.17 y 2.62 E-6 Te-129*
69.6 m 5.98 E-1
/ Te-132 78.2 h 8.86 E-3 l f.
Ba-140-12.79 d 2.26 E-3 La-140*
40.22 h 1.72 E-2 La-142*
95.4 m 4.36 E-1 l
Pr-144*
17.28 m 2.41 E0 A
Column.E: Back-Decay Correction Factor = CF(bd) = 1
=e
,- At
- - NOTE: Nuclides merked with
- are daughters in a decay chain. This must be taken into account in order to back-decay correct. Attachment SA should be followed for those nuclides.
s Column F: Temperature Correction Factor CF(t).
This factor converts uCi/cc to uCi/g.
If temperature of the water is (200*F, CF(t) = 1, and uCi/cc = uCi/g.
If temperature of the water is >200*F, use the Table below to
~
determine'CF(t).
Column G: Corrected uCi/g = reported uCi/cc x CF(bd) x CF(t)
(D) x (E) x (F)
(G)
=
Temperature RCS Water Temperature Correction Factor CF(t)
< 150*F 1.0
~
e' 200*F
.97 300*F
.92 400'F
.86 500*F
.79 600'F
.68 700*F 44 i
.3 ld 4
g
--,e,
-,,,e,
.,en
. ~, - - _., -. _ -
,,,,, -. ~.., _
/~
u 4
Page 1 of 2 ij-4 PARENT-DAUGHTER DECAY CORRECTION O
The Table on Page 2 of this Attachment lists the significant parent-daughter relationships. The decay scheme of the parent-daughter is < described as follows:
Q (U)
- B A B A Where:
100% fuel melt inventory (Ci) of parent
- Q*
=
100% fuel melt inventory (C1) of daughter
- Q'
=
Q (t) = hypothetical daughter activity (Ci) at sample time B
branching factor
- K
=
AA j Parent decay constant, (hr~ )*
~
daughter decay constant (hr
)*
=
AB time period from shutdown to time of sample (hr) t
=
1.
Calculate the hypothetical daughter concentration, Q (t) at the time of
~
B sampling assuming 100% fuel melt release of both parent and daughter activity.
2.
Determine the fraction (Fr) of the decay of the initial inventory of the daughter to the hypothetical daughter activity at sample time.
Q* (e~ B)
~
Q (*
B 3.
Calculate the amount of the measured sample specific activity associated with the decay of the daughter that was released.
Fr x measure specific activity (uCi/gm or uCi/cc)
M
=
B DI Where:
M measured activity of B
=
B 4.
Use this value of M as the reported uCi/cc in Column D of Attachment 5 or B
6 and continue with further corrections as necessary on Attachment 5 or 6.
See Page 2 of this Attachment for data on affected nuclides.
i
IP-1028 ATTACHMENT 5A Page 2 of 2 Parent Daughter O.
9(
Nuclide AA(hr~)
9 A (hr )
B B
Fr B*
~
A K
Nuclide B
I-131 3.59E-3 7.2E7
.008 Xe-131m 2.44E-3 4.7ES I-133 3.33E-2 1.5E8
.976 Xe-133 5.51E-3 1.5E8 Xe-133m 1.28E-2 2.1E7 1.0 Xe-133 5.51E-3 1.5E8 Sb-129
.161 2.3E7
.827 Te-129
.598 2.4E7 Te-129m 8.47E-4 5.8E6
.68 Te-129
.598 2.4E7 Ba-140 2.26E-3 3.5E7 1.0 La-140 1.72E-2 3.7E7 Ba-142 3.78 3.3E7 1.0 La-142
.436 3.1E7 Ce-144 1.02E-4 1.3E6 1.0 Pr-144 2.41 1.4E6
- M should be transferred to Attachment 5 or 6 into Column D, reported uCi/cc.
B l
IP-1028 ATTACHMENT 6 VC ATMOSPHERE SAMPLE DATA AND CALCULATIONS Sample No.:
(A)
(B)
(C)
(D)
(E)
(F)
(G)
Cb8EIlne Back-Decay Temp / Press Reported Correction Correction Corrected Nuclide Half-Life 1 (hr_g) uCi/cc Factor Factor uCi/ gram Kr-85 10.72 y 7.38 E-6 Kr-87 76.3 m 5.45 E-1 Kr-88 2.84 h 2.44 E-1 Xe-131m* 11.84 d 2.44 E-3
- Xe-133*
5.245d 5.51 E-3
. I-131 8.04 d 3.59 E-3 I-133 20.8 h 3.33 E-2 I-135 6.61 h 1.05 E-1 J
Cs-137 30.17 y 2.62 E-6 Te-129*
69.6 m 5.98 E-1 Te-132 78.2 h 8.86 E-3 Ba-140 12.79 d 2.26 E La-140*
40.22 h 1.72 E-2 La-142*
95.4 m 4.36 E-1 Pr-144*
17.28 m 2.41 E0 Column E: Back-Decay Correction Factor = CF(bd) = fe"*
NOTE: Nuclides marked with
- are daughters in a decay chain. This must be taken into account in order to back-decay correct. Attachment 5A should be followed for those nuclides.
Column F: Temperature / Pressure Correction Factor = CF(tp) = P(a)
(T(s) + 460)
P(s)
(T(a) + 460) 4 T(a), P(a) = VC atmosphere temperature *F and pressure (psia)
T(s), P(s) = VC sample temperature 'F and pressure (psia) i Column G: Corrected uCi/cc = reported uCi/cc x CF(bd) x CF(tp)
(D) x (E) x (F)
(G)
=
4'
-w
-.--n..--_,.-.....,.,-,.,,--.e
,n--
,..-----.-...,,e_,
IP-1028 ATDOMNr 7 CAIIIIATI(N OF ALTIVrlY RElIASED HOf mRE Total Activity &
E RCS & Simp RCS Sunp VC W Atmos.
VC Corrected Atmos.
VC wre Corrected RCS 4 = RCS Corrected Simps 4=
(uCf/cc) x 7.39E4 = (C1)
(C1)
(uCf/ gram) x (grams)** x 10 Nuclide (uci/gran) x (grams)* x 10 xr 45 Kr47 Kr-88 Xe-131m Xe-133 I-131 T-133 I-135 Cs-137 Te-129 Te-132 lb-140 la-140 la-142 Pr-144 Normally 90,000 gal. x 3785 cc/ gal. x 1 gran/cc 3.41 EB gras
=
Eter in W (gallms) x 3785 cc/ gal. x 1 gram /cc = VC amps (grams) 1his value should be based on data fran all available level instnmentation (see Attachment 2) ed should be approximately equal to ECW voltane added. VC amp and Recire. amp voltaes can be detennined using Control Roan graph book.
101E:
The activity is determined fran the Recire amp. This is aa====i to be the activity in all unter in the NC simps.
Consider this as a possible source of error.
- uCi/cc x 2.61 E6 cu. ft. x 2.83 E4 cc/cu. ft. x 10 C1/tci = 7.39 E4
d IP-1028 ATDK3MNr 7A
~
CAIINATIm OF ACTIVITY ISESINT IntING NDEMAL OPERATIWS (A)
(B)
(C)
(D)
(E)
(F)
(G)=(F)-(C)-(E)
Activity Released Corrected Ci eratims VC (C1)
Frta Core Released Normal Operations RCS (C1)
Normal Op(tC1/cc)* x 7.4E4***
Normal Ops (Attachment 7)
Prm Core VC Conc Nuclide RG Cone (uC1/cc)* x 320**
Nonnal Ops Kr-85 Kr-87 Kr-88 Xe-131m Xe-133 I-131 I-133 I-135 320
= 3.2E8 cc IES x IE-6 Ci/uci 4tain frcan recent pre-slutdown RCS sample:
- 7.4FA = 7.4E10 cc in VC x IE-6 C1/tCi
- Available from Omistry or Site Reactor Fagineer If unavailable, use the following approximate values as a am of the operation activity:
1 E
l Kr-85 12 Kr-87 12 kr-88 20 Xe-131m 40 Xe-133 200 1-131 8
I-133 10 I-135 10 The results in Coltan G to be used in Attachment 7B.
NUIE: Accatmt for Iodine spiking in accordance with Sectico 3.5.3 and Attaciment 8, If necessary.
IP-1028 ATDOMNT 7B CAIDIATIOi OF PERCENT CUD DAMAE (A)
(B)
(C)
(D) = (B) x (C)
(E)
(F)
(G) thcorrected Clad Pbuer Correction Factor PCF Corrected Activity Released (E/D) x 100%
10 or Iodine 14rlide Ihange Inventory (Attachent 4 or 4A)
Clad % Inventory Fr a Core (C1)*
Perc. Clad hee Ratios **
Kr45 1.6FA Kr-87 3.1D Kr-88 6.7D Xe-131m 7.5E2 Xe-133 1.5ES I-131 2.4E5 I-133 1.6ES I-135 8.3FA i
Prm Attacimmt 7A.
Noble Gas Isotope I
m Is tv
(
re to mtios h At h e 3.)
'Ihe percent fuel daage valtes can only be considered as approxinetions. If the actual age of the fuel assembly (s) danaged M7IE:
and the power region in the core is different frin the core average, (core average was used to develop the inventories in Coltam B) tien the actual inventories in the fuel d= aged could differ by a factor of 2-3. 'Ihe calculated percent damage nust be considered al with the isotopic ratios (Coltan C), presence of other ruclides, and other parameters as dia,mmt elseAere in this pr e.
i I
IP-1028 ATLOMNT 7C CAIIINATIOi OF PHICENT REL OVBGFAT (A)
(B)
(C)
(D) = (B)x(C)
(E)
(F)
(G)
PCF Corrected (E/D x 100%)
Uncorrected Puel Overteat Power Correction Factor Mael Overteat Activity Released Percent NG or Iodine Nuclide Release Inventory (C1)
(Attachment 4 or 4A)
Inventory (C1)
Fra Core (C1)*
Fuel Overkat Ratios **
Kr-85 9.0E5 Kr-87 1.8E7 Kr-88 2.5E7 Xe-131m 2.8E5 Xe-133 8.8E7 I-131 4.3E7 I-133 8.8E7 I-135 7.9E7 N/A Cs-137 4.9E6 N/A Te-129 1.5E7 N/A Te-132 6.2E7 N/A Ib-140 2.2E5 N/A la-140 2.5ES la-142 1.9E3 N/A N/A Pr-144 1.5E5 Fran Attacla nt 7.
Noble Cas Isotope I dine Isotope (Campare to ratio in Attachment 3.)
or Xe-133 I-131 If the actual age of the fuel assedily(s) daaged h percent fuel d= ape values can only be considered as approxiantims.
N[7IE:
arut the power region in the core is different fra the core average (core aver. age ms used to develop the inventories in h calculated percent damage mast le Coltam B), then the actual inventories in the fuel d=agevi could differ by 20-30%.
considered alorg with tie isotopic ratios (Colten G), presence of other ruaclides, and other parmeters as discussed elseulere in this procedure.
IP-1028 ATrX3MhT 7D CAIDIATI(N OF IDCH(r REL MI.T (A)
(B)
(C)
(D) = (B)x(C)
(E)
(19 (G)
FCF Corrected (E/D x 100%)
Uncorrected Ibel Melt Ibwer Correction Factor h 1 Melt Activity Released Percent NC or huline Nuclide Release Inventory (C1)
(Attaciment 4 or 4A)
Inventory (C1)
Fran Core (C1)*
Fuel Melt Ratios **
Kr-85 1.5%
Kr-87 3.0E7 Kr-88 4.2E7 Xe-131m 4.7E5 Xe-133 1.58 I-131 7.2E7 I-133 1.58 I-135 1.3G Cs-137 8.IE6 N/A Te-129 2.4E7 N/A Te-132 1.08 N/A lu-140 3.5E7 N/A 1a-140 3.7E7 N/A 1a-142 3.1E7 N/A N/A Pr-144 1.4m Fran Attactment 7.
- Noble Gas Isotope I dine Is tope (Campare to ratio in Attacisnent 3.)
"Ihe percent fuel damage values can only be considered as approximatims. If the actual age of the fuel assedily(s) dmagal and tie power regim in the core is different frum the core average (core average was used to develop tie inventories in Cohan B), then the actual inventories in the fuel damaged could differ by 3(MO%. 'Ihe calculated percent dange nust be considered along with the isotopic ratios (Cohan C), presence of other ranclides, and other parameters as discussed elsewtere in this procedure.
l
I IP-1028 ATTACRMENT 8 EXPECTED IODINE SPIKE VS. NORMAL IODINE ACTIVITY Average I-131 Maximum I-131 I-131 uCi/ gram
- Release (Curies)
Release (Curies) 0.5 1.0 3400 6500 0.5 380 950 0.1
.01
- 0.1 200 650
.001 -
.01 100 300 l
4.001 2
10 Normal operating I-131 specific activity in RCS.
IP-1028 ATTACHMENT 9 VC HYDROGEN CONCENTRATION VS. ZIRCONIUM-WATER REACTION Percent Zirconium Hydrogen Concentration Water Reaction in VC (Volume %)
10 1.3%
20 2.5%
30 3.8%
40 5.0%
50 6.3%
60 7.5%
70 8.8%
80 10.0%
90 11.3%
100 12.6%
IP-1028 ATTACHMENT 10 EXPECTED FUEL DAMAGE CORRELATION WITH CORE EXIT THERMOCOUPLE READINGS Fuel Damage Temperature *F*
No Damage
< 1300 Clad Damage 1300 - 2000 Ballooning of zircaloy cladding
> 1300 Burst of zircaloy cladding 1300 - 2000 0xidation of cladding and hydrogen generation
> 1600 Fuel Overtemperature 2000 - 3450**
Fission product fuel lattice mobility 2000 - 2550-Grain bouniary diffusion release of fission 2450-3450**
products Fuel Melt
> 3450**
Dissolution and liquefaction of UO in the
> 3450**
2 the zircaloy - Zr0 eutectic 2
Melting of remaining UO 5100**
2 These temperatures are material property characteristics and are non-specific with respect to locations within the fuel and/or fuel cladding.
Core Exit Thermocouple are not valid ever 3000*F.
NOTE:
When narrow range thermocouple readings go offscale (as indicated by an asterisk on the thermocouple map), use the wide range readings.
s-IP-1028 ATTACEMENT 11 EXPECTED CONTAINMENT RADIATION LEVELS POST-ACCIDENT (R-25/R-26)
Time After R/hr for 100%
R/hr for 100%
R/hr for 100%
Shutdown Clad Damage Fuel Overheat Fuel Melt 0
32 8.3 E4 1.4 E5 1 hr.
31 7.4 E4 1.3 E5 2 hrs.
27 5.9 E4 9.8 E4 4 hrs.
21 3.8 E4 6.4 E4 8 hrs.
16 1.9 E4 3.2 E4 12 hrs.
13 1.2 E4 1.9 E4 24 hrs.
11 7.2 E3 1.1 E4 48 hrs.
9.6 5.7 E3 9.0 E3 4 days 7.5 4.3 E3 6.7 E3 7 days 5.4 2.8 E3 4.5 E3 14-days 2.8 1.1 E3 1.8 E3 30 days 1.3 1.4 E2 2.2 E2 3
Radiation levels are taken f rom Ref erence 15.
IP-1028 ATTACHMENT 12 REFERENCES 1.
" Clarification of TMI Action Plan Requirements", NUREG-0737, USNRC, November 1980.
2.
"A Report to the Commission and to Public, NRC Special Inquiry Group",
M. Rogovin, 1980.
3.
"0RIGEN Isotope Generation and Depletion Code", Oak Ridge National Laboratory, CCC-217.
4.
Method of Calculating the Fractional Release of Fission Products from Onide Fuel, ANSI /ANS 5.4 - 1982.
5.
WCAP-9964, Westinghouse Electric Corporation.
6.
" Source Term Specification", ANS 18.1 Standard 1976.
7.
"Radionuclide Release Under Specific LWR Accident Conditions", Draft NUREG-0956, USNRC, January 1983.
8.
" Release of Fission Products from Fuel in Postulated Degraded Accidents",
IDCOR Draft Report, July 1982.
9.
"TMI-2 Accident: Core Heat-up Analysis", NSAC/24, January 1981.
10.
" Light Water Reactor Hydrogen Manual", NUREG/CR-2726, August 1983.
- 11. Westinghouse Emergency Response Guidelines.
12.
Analysis of the Three Mile Island Accident and Alternative Sequences, Prepared for NRC by Battelle, Columbus Laboratories, NUREG/CR-1219.
- 13. Westinghouse Owners Group Core Damage Assessment Methodology, February 1984.
14.
Core Damage Procedure based upon Post-Accident Chemistry and Radiation Sample Analysis, R. W. Bradshaw, R. D. Ivany, Combustion Engineering, Inc.,
November 1983.
15.
"High Range Containment Monitor Response to Post Accident Fission Product Releases", prepared by Sargent and Lundy for the New York Power Authority, SL-7009, May 1986.
indian Point 3 Nuc al215 RE-AD-1.0 p
guenanan. New wrk 10511 ATTACHMENT 6.7 914 739.8200
- > NewYorkPbwer
& Authority EMERCENCY PLAN PROCEDURES PROCEDURE NO. IP-1041 REV.
6 TITLE:
PERSONNEL MONITORING FOR EOF. TSC. OSC. AND CONTROL ROOM PERSONNEL "THIS PROCEDURE HAS BEEN EXTENSIVELY REVISED" 0
b WRITTEN BY:
M, //
20 7
REVIEWED BY:
p
~
SIGN TliRE/DATE PORC REVIEW:
MI' 3/3![ 7
/
DA APPROVED BY:
/[c.s "I
/$IGNATUREDATE EFFECTIVE DATE:
(3 3, S7
/
RES LOG #: hh'
IP-1041 PERSONNEL MONITORING FOR EOF, TSC, OSC, AND CONTROL ROOM PERSONNEL
'0 INTENT To describe the procedure to be used for personnel monitoring of the support forces located in the Emergency Operations Facility, Technical Support Center, Operations Support Center, and the Control Room.
2.0 DISCUSSION Emergency Operations Facility Technical Support Center, Operations Support Center, and the Control Room personnel shall be monitored as follows.
3.0 PROCEDURE 3.1 Non-Radiological Condition 3.1.1 A control point will be established by the H.P. Technician.
Radiological monitoring equipment should be set up and operationally tested.
3.1.2 Facility general radiation levels will be monitored by means of a dosimeter and TLD at each location. The dosimeters and TLDs should be taped to an inside wall at each facility at average chest level.
3.1.3 Personnel should report to their accountability officer prior to leaving the facility.
3.2 Radiological Condition Exists or Is Anticipated 3.2.1 A frisker station will be set up at all entry / exit points in each facility. Any person entering into the facility must frisk prior to entrv.
At the OSC/TSC locations, the interlocking door system will be energized and all entrv/ exit will be made through the OSC Control Point only.
Periodic h and b field measurements shall be made (RO-2, 3.2.2 E-530, or ecuivalent).
(See IP-1040, Habitability of the Emergency Facil.fty" for habitability criteria.)
In the EOF, EP-Forr d42 will be used to record the data.
In the OSC/TSC, and Control Roem, surveys will be documented on facility maps.
I of 2
fP-104L/6 3.2.3 To exit a facility:
EOF - sign in/out using EP-Form #45 wearing dosimetry and/or a.
protective clothing'as per H.P. monitor.
b.
Control Room - sign in/out using H.P. log book wearing dosimetry and/or protective clothing as per H.P. Technician.
OSC/TSC - sign in/out using H.P. log book or OSC Briefing c.
Form (EP-Form #18) wearing dosimetry and/or protective clothing as per H.P. Team Leader.
3.3 Dosimetry will establish exposure tracking system in accordance with approved plant procedures.
3.4 Environmental monitoring teams shall obtain their own dosimeter and TLD from the stock in the emergency vehicles.
2 of 2
indien Point 3
[25 RE-AD-1.0 p
Buchanan, New wk 10511 ATTACHMENT 6.7 914 739.8200
- > NewWrkPbwer 4# Authority EMERCENCY PLAN PROCEDURES PROCEDURE NO. IP-1050 REV.
12 TITLE:
ACCOUNTABILITY "THIS PROCEDURE RAS BEEN EXTENSIVELY REVISED."
WRITTEN BY:
can/ MM L!d b
SIGNA R/ TE
~
~
REVIEWED BY:
t
/
SIGNATliRE/ TE J////7 PORC REVIEW:
m
[
AT
//
A n>
1 5
APPROVED BY:
t I~ NATU E/ITATE G
EFFECTIVE DATE:
C3 3 / PJ RES LOG #:
IP-1650 ACCOUNTABILITY 1.0 INTENT This procedure is used during an emergency to assure that all IP-3
. personnel are accounted for.
2.0 DISCUSSION Accountability will be initiated at the Site Area Emergency classification (or earlier if directed by the Emergency Director).
When directed by the ED, Control Room personnel will sound the Site Assembly Alarm at which time all site personnel will report to their assigned assembly areas.
Initial accountability, resulting in a " missing persons" list, must be completed in about 30 minutes.
3.0 LOCATION OF ASSEMBLY AREAS 3.1 Control Room All Watch, off-watch Operations personnel, contingency, and " spares" will report to the Control Room until requested to report to the OSC.
3.2 All non-watch personnel will report to their assigned assembly areas (or to the nearest assembly area) when the Site Assembly Alarm sounds.
Assembly area locations are denoted by large orange signs with blue letters reading " ASSEMBLY AREA".
The larger assembly areas have been further subdivided into department sections.
All personnel shall report to their respective assembly areas as indicated below:
(See Attachment 8.1 for the site map with NYPA assembly area locations.)
3.2.1 (L) Training (Of fice Area):
Training personnel or persons in the immediate vicinity.
Personnel in formal training closses will be accounted for by the assigned instructor who will report their status to the office area.
3.2.2 (K) Administration Building (2nd Floor Lunch Room):
Administration Building personnel, Floors 2, 3, and 4 (except H.P., Chemistry and other designated personnel who will report to the OSC or TSC).
3.2.3 (C) Machine Shop (15' Elevation):
All non-watch J
maintenance and construction personnel within the security fence.
First Floor (Maintenance) personnel should use this as their assembly area.
3.2.4 (G)
Construction Conference Trailer:
NYPA Construction personnel and all other personnel in the immediate area.
1 of 5
IP-1050/12 3.2.5 (J) Warehouse (Office Area):
Warehouse personnel and personnel in immediate vicinity.
3.2.6 (H) Con Edison Service Center, West Store Room Area:
Personnel may eventually assemble here if evacuation of the site is necessary.
3.2.7 (TSC) Technical Support Center personnel should report to the TSC as their assembly area.
If a TSC designated individual is at another assembly area, they should report in at that area and then go directly to the TSC.
3.2.8 (OSC) Designated Operations Support Center personnel shall report to the Operations Support Center as their assembly area.
If an OSC designated individual is at another assembly area, they should report in at that area and then go directly to the OSC.
3.2.9 (EOF) Designated Emergency Operations Facility personnel shall report to the EOF as their assembly area.
3.2.10 SECURITY Security shall account for their own personnel.
3.2.11 CONTRACTORS Contractor personnel (non-supervisory) not assigned to the protected area do not routinely receive accountability training. These personnel will be accounted for by their supervisors who will report their status to the nearest accountability area.
4.0 LOCATION AND OPERATION OF ACCOUNTABILITY CARD READERS 4.1 Accountability card readers are installed in the following assembly areas within the protected area:
4.1.1 Control Room 4.1.2 OSC 4.1.3 TSC 4.1.4 2nd Floor Lunch Room - Administration Building 4.1.5 15' Elevation Machine Shop 4.2 When arriving at your assembly area, card into the accountability card reader insuring that the red light is lit.
Leave card in the reader until the light turns green, then remove your card.
Report to the Accountability Officer in the area.
4.3 If the red light is not lit, go directly to the Accountability Officer and check in.
4.4 If you are requested to change assembly areas (e.g.,
from the Lunch Room to the OSC). report to the Accountability Officer in the new Do not card out when leaving the assembly area.
area.
2 of 5
m IP-1050/12 5.0 RESPONSIBILITIES 5.1 LEAD ACCOUNTABILITY OFFICER (LAO) - The LAQs during normal work bours are those designated on Attachment 8.2.
During off hours, the Security Shift Coordinator is the LAO.
5.2 The Lead Accountability Officer should call Ext. 8067 (Security C.A.S.
Operator) to insure that the accountability card readers have been activated.
In addition, call all assembly areas to insure that the phones are functional.
If they are not, notify the OSC Manager.
5.2.1 Determine accountability in assembly areas outside the protected area.
5.2.2 Receive list of unaccounted for personnel within the protected area from Security (if the card readers are operable).
5.2.3 Develop missing persons list.
5.2.4 Inform Emergency Director as soon as protected area accountability is complete.
5.2.5 Inform Emergency Director and Area Accountability Officers when site accountability is complete. As necessary:
a.
Discuss Search and Rescue with Emergency Director. Only the ED can authorize a Search and Rescue effort.
b.
Discuss evacuation routes with Emergency Director and transmit that information to Area Accountability Officers.
5.2.6 If evacuation is to the Con Edison Service Center Building, an H.P.
will be dispatched (if radiological conditions warrant) to escort evacuees.
Re-accounting of personnel should be performed upon arrival at the Service Center.
5.2.7 Keep assembly areas informed of plant conditions.
5.2.8 If radiation levels in an assembly area are greater than 10 mR/hr., contact H.P. Team Leader in the OSC for direction.
If available, an H.P. will be dispatched to assembly area to verify radiation levels.
5.3 AREA ACCOUNTABILITY OFFICERS 5.3.1 Report to assigned assembly area.
Identify yourself as the Area Accountability Officer to all personnel.
5.3.2 Ensure the telephone and the P.A. speaker in assembly areas are functional.
5.3.3 If release is in progress or as directed, ensure all personnel frisk prior to entry.
3 of 5
IP-1050/12 5.3.4 Perform accountability as follows:
a.
Assembly areas with card readers:
1.
Insure that the card readers are activated by observing red light indication.
If it is not activated, call the LAO.
2.
Insure all personnel in your assembly area have punched into the card reader.
b.
All Accountability Officers complete the following:
1.
The Area Accountability Officer has a master list of personnel by department who should be reporting to each designated assembly area.
2.
Cross off all accounted for personnel and add any "others" reporting to the assembly area.
3.
Where possible, utilize first line supervisors to account for their personnel.
4.
A person should only be accounted for if they are visibly present or known to be off site.
This information should be given to the Accountability Officer by employees' supervisors.
If the person is
" thought" to be off site, it should be so noted.
5.
Upon completion, call the LA0 informing him/her of the status of accountability, c.
Account for " late reporters" as necissary to LAO.
5.3.5 Control area access by sign-in/ sign-out sheets at single access point.
5.3.6 Update assembly area personnel as to general plant conditions and developments as informed by the LAO.
5.3.7 Continuously monitor Ludlum 300/E-530 for changing radiological conditions.
If greater than 10 mR/hr.,
immediately contact the LA0 for possible evacuation.
5.3.8 Contact LA0 as necessary to resolve problems requiring immediate attention.
5.4 SECURITY 5.4.1 Activate the assembly area accountability card readers when the ALERT emergency is announced or when notified to do so by the Control Room or Lead Accountability officer.
4 of 5
IP-1050/12 5.4.2 Perform accountability in the protected area in accor' ance d
with the following:
a.
Account for all Security personnel.
b.
Generate an unaccounted for personnel list as soon as possible. Cross off all Security personnel, Utilize Security personnel (radio communications), the page c.
system, and call assembly areas within the protected area to identify / locate missing personnel.
d.
Notify the lead Accountability Officer in less than 30 minutes of any missing personnel.
5.4.3 If the accountability card readers are inoperable:
a.
Send the Visitor's List to the LAO, b.
Account for all Security personnel.
c.
Call the LA0 informing him/her. of the status of accountability.
6.0 0FF HOURS 6.1 All non-watch personnel should assemble at the Machine Shop, IS' Elevation.
6.2 TSC and OSC personnel should report to the TSC and OSC respectively.
6.3 All Watch, contingencv, and " spares" should report to the Control Room.
6.4 Responsibility for assembly area habitability surveys shall be assumed by the Shift Supervisor and the Watch Health Physics Technician or other individuals designated by the Shift Supervisor.
7.0 PERSONNEL CONTAMINATION CHECK (IP-1060) 7.1 If personnel are relocated from an assembly area to the Con Edison Service Center due to radiation levels, they should be checked for contamination.
7.2 Prior to leaving the Con Edison Service Center, personnel and vehicles should be re-checked for contamination.
8.0 ATTACHMENTS 8.1 Assembly Area and Evacuation Route Map 8.2 Accountability Telephone Listing 8.3 Lead Accountability Checklist 8.4 Accountability Officer Checklist 8.5 Security Accountability Checklist 5 of 5
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IP-1050 ATTAC': MENT 8. 2 Page I of 2 ACCOUNTABILITY ASSEMBLY ACCOUNTABILITY OFFICE AREA ACCOUNTABILITY AREA 0FFICER EXTENSION EXTENSION L
Training Bobby Martin Steve Smith Doug Ames Bill Swirdell J
Warehouse David Dicioccio Lou Tiberi K
Administration Jill Chema Jim Reagan George Nikolatos Marianne Tansky C
Machine Shop Chuck Alphin Mike Devlin Bruce Witherall G
Construction William Eichert Conference Ronald Mackowiak Trailer Marie Campanaro Control Room Gail Ruh TSC Jean Moretti Ed Noel Al Froebrich OSC Anthony Vitale Cliff Marks Marv Ellen Mastregiacomo EOF Laura Earens Pan Walsh Lori Gierloff Nancy Povie H
Con Edison Service Center (West Store Room Area)
FtM1 131987
IP-1030 ATTACHMENT 8.2 Page 2 of 2 TELEPHONE EXTENSIONS Unit 3 Control Room and Page Unit 2 Control Room (Con Edison)
Shift Supervisor's Office Operations Superintendent Office Security Shift Coordinator Security Building Extensions Con Edison LA0 Westinghouse (Ray Heisey)
OSC TSC EOF (Emergency Director)
LEAD ACCOL'NTABILITY OFFICERS OFFICE EXTENSIONS (Normal Working Hours)
Ruthanne Bowman Sal Colemi Christine Metzger Mary Ann Petrillo SECURITY ACCOUNTABILITY OFFICE EXTENSIONS (Protected Area during Normal Working Hours)
Security Shift Coordinator L. J. Malaspina Harry Bain LEAD ACCOUNTABILITY OFFICER (Off Hours)
Security j_g ',3 tv3I
IP-1050 ATTACFMENT 8.3 LEAD ACC0l'NTABILITY OFFICER CHECKLIST 1.
Upon hearing the Site Assembly Alarm, call NYPA Construction Office
(
) and Training (X
) to notify them of the Site Assembly Alarm in the event the alarm cannot be heard.
2.
Verify activation of the assembly area accountability card readers by
~
calling the Security C.A.S. Operator (X
).
3.
Verify operability of phones in all assembly areas.
If they are not functional, call the OSC Manager.
4.
Await calls from the Area Accountability Officers located outside the protected area for the status of accountability in those areas.
If an assembly area outside the protected area has not reported their status within approximately 15 minutes, call that area for accountability status.
Phone numbers for those areas are as follows:
a.
NYPA Construction b.
EOF c.
Warehouse d.
Training 5.
If a page is required to locate missing personnel, call the OSC at and ask for the Accountability Officer.
6.
In order to run a badge check, call Security at asking for the C.A.S. Supervisor or asking for the C.A.S. Operator.
7.
When " protected area" accountability is complete, Security will notify the LA0 of any unaccounted for personnel.
8.
When protected area accountabilf tv is complete, notify the Emergency Director and name all unaccounted for personnel.
9.
When " site" accountability is complete, discuss Search and Rescue operations with the Emergency Director, if necessary.
10.
Discuss evacuation routes with the Emergency Director and coordinate evacuation with Area Accountability Of ficers and H.P. Team Leader in the OSC as necessary.
- 11. Keep all assembly areas informed on plant conditions.
12.
If Area Accountability Officers report radiation levels of greater than 10 mR/hr., contact the H.P. Team leader in the OSC for direction.
If available, an H.P.
will he dispatched to the assembly area to verify radiation levels.
3 m
IP-1050 ATTAC'>f ENT 8. f.
e AREA ACCOUNTABILITY OFFICER CHECKLIST 1.
Report to assigned assembly area.
Identify ~ yourself as the ' Area Accountability Officer to all personnel.
2.
Ensure the telephone and the P.A.
speaker in assee.bly areas are.
functional.
3.
If release is in progress or as directed, ensure all personnel frisk -
prior to entry.
...s j
4.
Perform accountability as follows!
a.
Assembly areas with card readetst g
Insure that the card readers are activated by observina red light indication.
If it is not activated, call the LAO.
Insure all personnel in your assembly area have punched'into the card reader.
b.
All Accountability Officers complete t'he-fellowing:
1 On the master list of personnel by department, which shows who should be reporting to each dssignatrid assembly area, cross off all accounted for peconn'e' and add any "others" reporting to_the assembly area q
Were possib&e, utilize first line supervisors te see.ount for their personnel.
A person shou'Id only be accounted for if they are visibly present'or knavn to be off site. This informatien s!.ould be given to the Accountability Officer by en loyees' supervisers.
If the person is " thought" to be off site, it i
should be so noted s
cowf ecion, call the LA0 informing hie /her of the l
Upon status of accountability.
c.
Account for " late reporters" as necessary to LAO.
5.
Control area access by sign-in/ sign-out sheets at sind e access point.
6.
Update < assembly area personnel as to general plant condi: lend and developments as informed by the LAO.
7.
Continuously monitor Ludium 300/E-530 for charging' radiologi' cal conditions.
If 3reater than 10 mR/hr., immediately tentact the LA0 for possible ' evacuation.
1 i
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Contact LA0 as necessary to t er'olve problems' t w iring ircediote l
attention.
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IP-1050 ATTACHMENT 8.5
/
SECURITY ACCOUNTABILITY CHECKLIST 3
1.
Activate the assembly area accountability card readers when the ALERT Emergency is announced or when notified to do so by the Control ~ Room or lead Accountability Of ficer.
r 2.-
When ' the Site Assembly Alarm is sounded, insure card readers are activated and account for all personnel in the protected area.
31 Parfona saccountability in the protected area in accordance with the
~
'following:
1 vi t
/'
a.
Account for all Security personnel.
b.
Generate an unaccounted for personnel list as soon as possible Cross off all Security personnel.
c.
Parform a badge check to determine if they are in the protected area and, if so, what their-last location was.
1 l,
-d.
Utilize Security personnel (radio communications), the page system, and by calling assembly areas within the protected area to identify / locate missing personnel.
i,
, 4. e Provide the Lead Accountability Officer with the names of any f
unaccounted for personnel as soon as possible.
- y~
As requested by Lead Accountability Officer, perform badge checks of i
5 '. -
unaccounted for personnel to determine if they are within the protected area. Advise Lead Accountability Officer.
6.;
During off hours, the Security Shift Coordinator is the Lead Accountab'ility Officer and must keep the Emergency Director advised of 1
tie status of accountability for the entire site.
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indien Point 3 f$*[215 RE-AD-1.0 Benanan. New Wrk 10511 ATTACHMENT 6.7 914 739 8200
- b NewYorkPower
& Authority I
EHERGENCY PLAN PROCEDURES PROCEDURE NO. IP-1055 REV.
4 TITLE:
FIRE EMERGENCY WRITTEN BY 814/
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REVIEWED BY:
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SIGNATURE /DATE
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DATE
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APPROVED BY:
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EFFECTIVE DATE:
RESLOG#l bh-7
IP-1055 FIRE EMERGENCY 1.0 DISCUSSION The following are fire hazards which may occur which would require activation of the Emergency Plan and subsequent notifications.
Notification of Unusual Event An uncontrolled fire in the protected area not affecting safety systems.
Alert Fire of a magnitude that may significantly affect safety systems.
Site Fire of a magnitude that causes safety systems to become inoperable such that the ability to reach a safe condition could not be guaranteed.
2.0 REFERENCES
2.1 AP-27.3 - IP-3 Site Fire Protection Procedures 2.2 FP Fire Notification Guidelines 2.3 IP-1056 -' Directing Fire Fighting Personnel in Controlled Area 3.0 PROCEDURE 3.1 Person (s) should call the Control Room (
or page 4L party phone) immediately to report a fire. The location, size, and type of the fire, and callers' name should be given.
3.2 Control Room shall sound the fire siren.
3.3 Control Room shall anr.cunce over the P.A. System for the fire brigade to respond. Message shall be repeated twice and the siren resounded.
3.4 All unnecessary personnel shall be evacuated from the fire area.
3.5 When requested by the Fire Brigade Leader, the SRO, under the direction of the Shift Supervisor, shall call Verplanck Fire Department for assistance.
3.6 See IP-1056, Directing Fire Fighting Personnel in Controlled Area, if applicable.
1 of 1
indian Point 3 Cuclear Power Plant PO. Bos 215 RE-AD-1.0 Buchanan, New York 10511 ATTACHMENT 6.7 914 739 8200
- > NewYorkPbwer 4# Authority EMERGENCY PLAN PROCEDURES PROCEDURE NO. IP-REV.
TITLE:
Personnel Radiological Check and Decontamination WRITTEN BY:
[],l/0X4f MM 8
Of
/ SIGNAT
/D E 3 J 89 REVIEWED BY:
SIGNATURE /DATE
}!/ 2[/ 7 PORC REVIEW:
/$/
.Ao DATE /
APPROVED BY:
/#
3 4 E/[
SIGN E/D E d?[ A k 7 EFFECTIVE DATE:
U RES LOG f:
b-
IP-1060 PERSONNEL RADIOLOGICAL CHECK AND DECONTAMINATION 1.0 INTENT To describe the emergency condition methods of checking personnel for contamination and their subsequent decontamination when required.
2.0 DISCUSSION Determination of personnel contamination levels, supervision of personnel decontamination and subsequent checkout will be performed by members of the Health Physics staff.
Resolution of problem cases will be handled by the Radiological Assessment Team Leader (RATL).
3.0 DECONTAMINATION FACILITIES 3.1 Decontamination facilities available include:
a.
Decon facility on the 4th floor of the Administration Building located at the HP Control Point;
-D b.
Con Edison Service Center Building.
3.2 Decontamination supplies are available at each location.
4.0 PRECAUTIONS 4.1 Decontamination will be performed in accordance with RE-HPI-6.41.
4.2 Chemical decontamination should only be performed with medical supervision or under direction of a knowledgeable individual.
4.3 Clean is considered less than 100 cpm above background.
5.0 PROCEDURE 5.1 Personnel will be monitored for contamination:
a.
when leaving restricted areas; b.
when leaving areas in the plant suspected to be contaminated; c.
assembly areas if suspected to be contaminated; d.
re-assembly areas.
5.2 Records of personnel monitoring will be maintained on the Personnel Contamination Check Form, EP-Form #14 (Attachment 6.1).
S'in k
5.3 Records of personnel decontamination will be maintained on the Decontamination Record Form, EP-Form #15 (Attachment 6.2).
NOTE:
EP-Forms #14 and #15 are to be returned to the Watch H.P. or H.P.
Team Leader in the OSC as applicable.
1 of 2
IP-1060/3 5.4 Documentation of all monitoring and decontamination activities will be directed to the Health Physics Team Leader in the OSC for evaluation and' retention.
5.5 H.P.
Control Point and Decon Facility decontamination will be performed in accordance with RE-HPI-6.41.
5.6 Decontamination at the Con Edison Service Center:
5.6.1 Determine the contamination level category by using a frisker with an HP-210 G.M. tube to check the individual.
The categories are as follows:
a.
Clean
- less than 100 cpm above background; b.
Low level - less than 10,000 cpm above background; High level - 10,000 cpm above background or greater.
c.
5.6.2 For individuals contaminated in the Low Level category, use the Service Center locker room shower. This amount will not exceed the limits specified in 10CFR 20.303 for discharge into a sanitary sewage system, a.
Shower using non-alkaline soap such as Phisoderm, if available, and lukewarm water. Keep contamination away from 1
non-contaminated parts of the body.
If practical, wash off higher levels of contamination first.
b.
Recheck individual af ter shower.
Levels less than 100 cpm above background are considered uncontaminated.
c.
If the levels are still greater than 100 cpm above back-ground, have the individual re-shower and then re-check.
d.
If the individual remains contaminated after three (3) showers (over 100 cpm above background),
consult the Radiological Assessme.it Team Leader.
5.6.3 For individuals contaminated in the High Level category, use the portable sample kit, decon kit, and transportation kit located in the Medical Bureau Office at the Service Center.
D The instructions for their use are included with the kits.
The key is located with the Service Center Guard.
5.6.4 When an individual is decontaminated to a level less than 10,000 cpm above background using a "frisker" with an HP-210 C.M. tube or eauivalent, he may be referred to the shower room where further decontamination may be continued.
6.0 ATTACHMENTS 6.1 EP-Form #14 - Personnel Contamination Check Form 6.2 EP-Form #15 - Skin Decontamination Record Form 2 of 2
IP-1060 ATTACHMENT 6.1 EP-FORM #14 PERSONNEL CONTAMINATION CHECK DATE:
INSTRUMENT MODEL:
H.P.:
INSTRUMENT SERIAL NO.:
MAXIMUM l
FRISKER DESCRIPTION OF AREA DISPOSITION INDIVIDUAL'S NAME (CPM)
WITH READING > 100 CPM OF INDIVIDUAL i
l f
NOTE:
All personnel leaving restricted areas or other areas suspected to be contaminated should be surveyed. Record on this form whether '
contaminated or not.
Return this form to the Watch H.P. or H.P. Team Leader in the OSC as applicable.
FEB 0 31987
I IP-1060 ATTACHMENT 6.2 Page 1 of 2 EP-FORM #15 SKIN DECONTAMINATION RECORD SOCIAL NAME:
SECURITY NO :
(LAST)
(FIRST)
(INITIAL)
DATE:
TIME OF CONTAMINATION:
H.P. TECHNICIAN:
How and where it occurred:
Max. initial contamination levels: With Anti-C:
W/0 Anti-C:
Body Orifices / Swabs or Smears / Counting Instrument Used (S/N):
1 2
3 4
5 TIME CPM TIME CPM TIME CPM TIME CPM TIME CPM EYE EAR I
NOSE MOUTT OTHER TIME DECON CONTAMINATION DECON STEP SKIN AREA AGENTS LEVEL AFTER SKIN BEGINS CONCERNED USED DECONTAMINATION CONDITION l
TIME DECONTAMINATION COMPLETED:
DECONTAMINATION DONE BY:
Return this form to the Watch H.P. or H.P. Team Leader in the OSC as applicable.
FEB 0 319871 i
IP-1060 EP-FORM #15 ATTACHMENT 6.2 (continued)
Page 2 of 2 NAME OF SURVEYED INDIVIDUAL:
BY:
SURVEY DIRECTIONS:
INDICATE LEVELS OF TIME:
DATE:
CONTAMINATION ON THE METER CORRESPONDING BODY PART.
TYPE:
SERIAL CALIBR.
NO:
DUE:
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indian Point 3 Nuclear Power Plant h,df;** *"
IN^AENE0T6.7 AfD NewYorkPbwer 4# Authority 9 8 EMERGENCY PLAN PROCEDURES PROCEDURE NO. IP-1063 REV.
5 TITLE:
VEHICLE / EQUIPMENT RADIOLOGICAL CHECK AND DECONTAMINATION WRITTEN BY h)1/ iff 8
7 REVIEWED BY:
SIGNATURE /DATE
# [
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APPROVED BY:
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UR DATE EFFECTIVE DATE:
63.3!99 l
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RES LOG f:
IP-1063 VEFICLE/EOUIPMENT RADIOLOGICAL CFECK AND DECONTAMINATION 1.0 INTENT To describe the methods of checking vehicles and equipment for contamination and their subsequent decontamination at the Con Edison Service Center when required.
2.0 DISCUSSION During a Site Area or General Emergency, vehicle access to or departure from the Indian Point Site is stopped with the exception of emergency vehicles.
Permission for vehicles or equipment to leave the site is obtained from the Emergency Director after evaluation of contamination (potential or actual).
The criteria presented in this procedure shall be used to determine the status of site personnel private vehicles in the event of a site evacuation that requires contamination checks (see IP-1053).
The responsibility for performing contamination checks and supervision of any decontamination is that of the Onsite Monitoring Team which is made up of Health Physics personnel.
Vehicle and equipment decontamination will be performed at the Service Center parking area (or other designated area) which will be accessed through the south gate located near the gasoline pumps.
This gate will be opened by a member of the Security Force.
No vehicles will be allowed to leave the Broadway entrance unless authorized by the Emergency Director.
3.0 PROCEDURE 3.1 Check for removable (loose) contamination by making smear checks (gauze pads or paper disks),
a.
A gauze pad smear is made up of the major surface area of the outside of the car and tires.
The pad is then placed against an RM-14/HP-210 probe or equivalent.
No rise above background is considered uncontaminated.
Any indication above background will g
require paper disk smears to quantify activity.
b.
A paper disk smear is made of a number of representative areas (100 cm in size) and counted with a GM or proportional counter 2
or the RM-14/HP-210 or equivalent.
To be considered uncontaminated.., the removable contamination must be less than 1000 dpm/100 cm' or less than 100 cpm above background using the RM-14/HP-210 or equivalent.
{
3.2 Check for fixed contamination by slowly passing the E-530/166 probe or equivalent as close as possible to the surface.
To be considered uncontaminated, the fixed contamination, as seen by the instrument, must be less than 0.1 mR/hr. above background.
I of 2
IP-1063/5 3.3 If the vehicle or equipment is contaminated, have it moved to the decontanination location in the northeast corner of the Service Center North parking lot.
3.4 Position the vehicle or large equipment close to the corner water run-off opening.
This will allow contamination to run off into a small depression where it will be contained and concentrated by the land contour.
Isolate and post the run off area, as necessary.
3.5.Using hoses hooked up to the nearest fire. hydrant or utilizing a Fire Deparment pumper, wash the vehicle or equipment with the detergent and water.
3.6 If the vehicle or equipment is still contaminated, rewash and recheck until it checks out uncontaminated.
3.7 Record all contamination checks and washes along with the vehicle license plate number on EP-Form #16 and equipment information on EP-Form #17 (Attachments 4.1 and 4.2).
Return these forms to the Watch H.P. or H.P. Team Leader in the OSC as applicable.
3.8 Vehicles (not including environmental monitoring vehicles) that are contaminated will not be permitted to leave the site.
NOTE:
Release of vehicles and equipment at contamination levels greater than the 1000 dpm limit may be required under certain conditions (high offsite contamination levels, emergency medical treatment).
Permission to release vehicles and equipment in such circumstances must be obtained from the Emergency Director.
4.0 ATTACHMENTS 4.1 EP-Form #16 - Vehicle Contamination Check 4.2 EP-Form #17 - Equipment Contamination Check 2 of 2
YP-1063 ATTACHMENT 4.1 EP-FORM #16 VEHICLE CONTAMINATION CHECK DATE:
H.P.:
COUNTER SERIAL NO.:
COUNTER CALIBR. DUE DATE:
FRISKER SERIAL NO. :
FRISKER CALIBR. DUE DATE:
LOOSE VEHICLE CONTAMINATION FIXED DESCRIPTION 2
LICENSE DPM/100 CM CONTAMINATION OF AREA DISPOSITION NUMBER OR CPM > BKGD CPM > BKGD CONTAMINATED OF VEHICLE RETURN THIS FORM TO THE WATCH H.P. OR H.P. TEAM LEADER IN THE OSC AS APPLICABLE.
FEB 0 3 09871
IP-1063 ATTACHMENT 4.2 EP-FORM #17 EQUIPMENT CONTAMINATION CHECK DATE:
H.P.:
COUNTER SERIAL NO.:
COUNTER CALIBR. DUE DATE:
FRISKER SERIAL NO.:
FRISKER CALIBR. DUE DATE:
LOOSE CONTAMINATION FIXED 2
EQUIPMENT DPM/100 CM CONTAMINATION LOCATION OF DISPOSITION DESCRIPTION OR CPM > BKGD CPM > BKGD CONTAMINATION OF EQUIPMENT
'l RETURN THIS FORM TO THE WATCH H.P. OR H.P. TEAM LEADER IN THE OSC AS APPLICABLE.
FEB 0 31987
... -... ~ _ _ _ - _,,.
indian Point 3 Nuclear Power Plant PO. Boo 315 RE-AD-I.0 Buchanan, New York 10511 ATTACPJfENT 6.7 914 739 8200
- > NewWrkPower tv Authority EMERGENCY PLAN PROCEDURES PROCEDURE NO. IP-1076 REV.
7 TITLE:
BEEPERS WRITTEN BY:
'i,1aa 03 A!S 7 e
m SIGNATURE T
REVIEWED BY:
!O7 SIGN TURE/DATE
~
3 2h7 PORC REVIEW:
-f DA E J 2 2
APPROVED BY:
.N UR DATE
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83!3!d7 EFFECTIVE DATE: V RES LOG #:
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s IP-1076 BEEPERS 1.0 INTENT To outline the procedure for beeper page ' units used to contact NYPA personnel for emergency notification.
2.0 DISCUSSION The paging service used at IP-3 is Metromedia Telecommunications Paging Services.
A VHF band is used.co transmit signals to the paging units via towers located throughout the service area.
The IP-3 pagers have digital readout as well as tone. This allows the person being paged to know where to call. back (i.e.,
directly to the paging party).
Use of a touch-tone telephone is required for the digital message. ~Should rotary telephones be I
used, the pager will. beep but no digital display will be available..
i Therefore, the paged person will not know who to call.
It is best to use Security (
), as an intermediary if dialing from a rotary phone and they will be able to page personnel from the plant touch-tone-phones; or if the " air" is busy, to dial (
)
and ask Metromedia to page manually.
Again only a beep will be heard and a message must be left with Security on who to call back.
NOTE:
Site telephones are both the touch-tone dial and rotary dial types.
The multi-extension console model is rotary dial and the single user desk model is touch-tone.
See Section 4.0 for instructions on paging from these different types of telephones.
3.0 PRECAUTIONS 3.1 Beepers are fragile and should be handled carefully.
I 3.2 Beepers should be worn and in the "on" position when you are away from your home telephone, office telephone, or plant paging system.
3.3 Missing or defective beepers should be reported immediately to the R.E.S. Department. (
)
4.0 PROCEDURE 4.1 Establish the necessity to use beeper paging (i.e., no answer-at home or office phones or by plant paging).
4.2 There are four (4) different ways to page individuals with these digital display pagers.
In order of preference:
e a.
Using a touch-tone phone, enter the telephone number for call back as per the following instructions:
4 i
1 of 3
IP-1076/7 DATA PAGE SERVICE - How to send a message to:
NOTE: You must use a touch-tone phone or adapter.
1.
Dial the Beeper No.:
2.
Wait until you hear 3 beep cones: beep / beep / beep.
3.
Enter the phone number where you wish to be q
called.
(You can enter up to 24 digits.)
4.
Press the number sign button (#).
5.
Hang up.
b.
If you are using a rotary phone, you can call Security (
)
and have them use the plant touch-tone phones so they can enter the number for callback (see Section 4.2.'a.
for instructions).
c.
If you are paging someone using a rotary phone, the paged individual will only hear a beep.
In this instance, Security must be called (
) leaving a message with them so that when the person being paged calls in. Security can relay the message.
d.
If a busy signal is obtained, you can call Security (
' end request them to call Metromedia at (
)
to have chem do the paging manually.
No digital display will appear, only a beep.
In this instance, Security must be called leaving a message with them so that when the person being paged calls in.
Security can relay the message.
4.3 Instructions when you are paged:
7 Call back telephone number which appears on your pager printout.
a.
b.
If only a beep is heard and/or a 1
appears in the center of the screen, this means you were:
- 1) paged with a rotary phone; or 2) paged manually by Metromedia; or 3) the telephone code entered was not signaled to you properly.
Should you receive only a beep, call Security (
) to ask for your message, if one has been left for you.
g 5.0 TESTING 5.1 Beepers will be tested every other month to insure their operability and use in accordance with 3PT-TM03.
5.2 A letter will be distributed to all beeper holders notifying them of the test and the date of the testing period..4 will be used to record the results of the beeper test.
5.5 Follbw instructions in Section 4.2 to test each beeper.
)
2 of 3
IP-1076/7 6.0 ATTACHMENTS 6.1 Beeper Use Instructions 6.2 Metromedia Paging Network 6.3 Beeper Holders 6.4 Beeper Test Record 3 of 3
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I The boundaries of the Company's service area are not precisely drawn and are, as shown. subject to minor vanations. In addition, signal strength may f
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Metromedia Telecommunications Paging Services 429 Sylvan Avenue Englewood Chffs, NJ 07632
I IP-1076 ATTACEMENT 6.3 BEEPER HOLDERS Beeper Holders Pager No.
Alb r igh t, Ma r t y................................ (
)
Brons, Jack....................................
Carano, B111...................................
Deschamps, Bob.................................
DiChiara, Joe..................................
Dube, Joe......................................
Forte, Todd....................................
Gillen, Jim....................................
Gullick, Jerry.................................
- Hahn, John.....................................
'7
- Hamlin, B111...................................
- Heady, B111....................................
Josiger, Bil1..................................
- Kelly, Larry...................................
l I
- Munoz, Steve...................................
7 Perrotta, Joe..................................
Russell, Joe...................................
- Russell, Pat...................................
Quinn, Dennis..................................
l7
- Smith, Steve...................................
Tagliamonte, Ed................................
Vignola, Joe...................................
Chemistry Supervisors..........................
Nuclear Generation Duty Officer................(
)
?
DATA PAGE SERVICE How to send a message to:
NOTE: With this type of paging, NOTE: You must use a touch-tone phone the individual will see or adapter.
the number you entered on 1.
Dial the Beeper No.:
his pager and will know 2.
Wait until you hear 3 beep cones:
where to call back.
beep - beep - been.
3.
Enter the phone number where you wish to be called.
(You can enter up to 24 digits.)
4.
Press the number sign button (#).
5.
Hang up.
If using a rotary phone, follow the instructions in Section 4.0 of this l
procedure.
i I
9
IP-!O 4 ATTAC10fENT 6,1 DATE:
BEEPER TEST RECORD
-Individual Pager No.
Result of Test Comment Albright, Marty Brons, Jack Carano, Bill Deschamps, Bob DiChiara, Joe Dube, Joe Forte, Todd Gillen, Jim Gullick, Jerry 7
Hahn, John Hamlin, Bill Heady, Bill Josiger, Bill Kelly, Larry Munoz, Steve
'r*
Perrotta, Joe Russell, Jce Russell, Pat Quinn,- Dennis Smith, Steve Tagliamonte. Ed Vignola, Joe Chemistry Supvs.
Nuclear Generation Duty Officer 4
For Service or Manual Paging: Metromedia Telecommunications Signature of A.R.E.S.S.
Signature of Test Coordinator /Date Signature of Performance and Reliability Supervisor /Date