ML20126B674

Emergency Plan Implementing Procedure EC-13, Reactor Core Damage Estimation

ML20126B674
Person / Time
Site:	Clinton
Issue date:	04/20/1985
From:	Evans K ILLINOIS POWER CO.
To:
Shared Package
ML20126B666	List: ML20126B674 U-600069, Forwards Emergency Plan Implementing Procedure EC-13, Reactor Core Damage Estimation. Procedure Revised to Include Third Core Damage Class,Referred to as Fuel Overheat Per TMI Item II.B.3
References
EC-13, NUDOCS 8506140179
Download: ML20126B674 (39)

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, i ILLINOIS POWER COMPANY PROCEDURE: EC-13 1 CLINTON POWER STATION REVISION: 0 i

EMERGENCY PLAN IMPLEMENTING PROCEDURE DATE: 4/20/85 PAGE: 1 of 14 TITLE: REACTOR CORE DAMAGE ESTIMATION i, -

AUTHORITY Function Signature Date Prepared By ib O/Il[I Supervisor-Emergency Planning #

l 47/f[ Pf [

Concurrence -

Concurrence ilt /

4A d-ff jSj 5

,gg Concurrence Quality Assurance x MAY 2 31985

( Facility Review Group '

,d 0 ,

Power Plant Manager Final Approval K* 4'

'N / '.

f'b ['

PROCEDURE HISTORY Revision Date Revision Date Revision Date fDR D D F

ILLINOIS POWER COMPANY PROCEDURE: EC-13 i t * *CLINTON POWER STATION REVISION: 0 EMERGENCY PLAN IMPLEMENTING PROCEDURE DATE: 4/20/85 PAGE: 2 of 14 TITLE: REACTOR CORE DAMAGE ESTIMATION I I

CONTENTS

1.0 INTRODUCTION

2.0 RESPONSIBILITY 3.0 DEFINITIONS 4.0 INSTRUCTIONS 4.1 Core _ Damage Estimate from PASS 4.2 Hydrogen Analysis 4.3 Drywell Radiation Analysis 4.4 Reactor Core Uncovery Time 4.5 Identification of Release of Source By Determination of Fission Product Ratios 4.6 Analysis For Ba, Sr, La, Ru

5.0 REFERENCES

6.0 ATTACHMENTS Appendix A, Core Inventory of Major Fission Products in a Reference Plant Operated at 3651 MWt For Three Years Appendix B, Fission Product Concentrations in Reactor Water and Drywell Gas Space During Reactor Shutdown Under Normal Conditions Appendix C, Ratios of Isotopes in Core Inventory and Fuel Gap Appendix D, Relationship Between 1-131 Concentration in the Primary Coolant (Reactor Water + Pool Water) and the Extent of Core Damage in Reference Plant Appendix E, Relationship Between Cs-137 Concentration in the Primary Coolant (Reactor Water +

Pool Water) and the Extent of Core Damage in Reference Plant

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~ Appendix F, Relationship Between Xe-133 Concentration in the Containment Gas (Drywell + Primary Containment) and the Extent of Core Damage in Reference Plant Appendix G, Relationship between Kr-85 Concentration in the Containment Gas (Drywell + Primary Containment) and the Extent of Core Damage in Reference Plant Appendix H, Sample Most Representative of Core Conditions During an Accident for the Estimation of Core Damage Appendix I, Plant Parameters Appendix J, Sample Calculation of Fission Product Inventory Correction Factor Appendix K, Percent of Fuel Inventory Airborne in the Drywell Appendix L, Maximum Acceptable Core Uncovery Time Vs Time After Reactor Shutdown

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1.0 INTRODUCTION

1.1 The purpose of this procedure is to determine the degree of reactor core damage based on water and gas samples taken from the primary system, and drywell radiation levels during accident conditions.

1.2 There are four general classes of fuel damage and three degrees of damage within each of the classes except for the "No Fuel Damage" class.

Class of Minor Intenmediate Maj or Fuel Damage (<l0%) (10%-50%) (>50% )

No Fuel Damage 1 1 1 Cladding Failure 2 3 4 Fuel Overheat 5 6 7 Fuel Melt 8 9 10 The objective of this procedure is to narrow down, to the maximum extent possible, those categories which apply to an actual inplant situation.

1.3 In determining the extent of core damage, an initial core damage assessment will be made based on radionuclide measurement, j This initial assessment consists of:

l a) Obtaining samples from the Post-Accident Sampling System (PASS).

b) Analyzing the samples for major fission product concentrations by gamma ray spectrometry.

c) Decay correcting samples to the time of reactor shutdown.

d) Normalizing the sample concentrations with reference plant data from a BWR-6/238 with a Mark 3 Containment.

I

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., ~ EMERGENCY PLAN IMPLEMENTING PROCEDURE DATE: 4/20/85 PAGE: 5 of 14 TITLE: REACTOR CORE DAMAGE ESTIMATION e) Comparing the normalized concentrations to Reference Plant Concentrations Vs. Core Damage Graphs developed by General Electric to estimate the amount of core damage.

This initial core damage assessment will provide one or more candidate categories of possible core damage which will most-likely represent the actual in-plant condition.

After the initial assessment is made other parameters should then be evaluated to corroborate and further refine the initial estimate. These parameters should include:

a) Containment hydrogen levels which provide a.

measure of the extent of metal water reaction which, in turn, can be used to estimate the degree of -clad damage.

b) Drywell radiation levels, which measure core

-damage by an indication of the inventory of airborne

-fission products (i.e., noble gases, a fraction of the halogens, and smaller fraction of the particulates) released from the fuel to the drywell.

c) ' Reactor Vessel water level, which is used to establish if there has been an interruption of adequate core cooling. Significant periods of core uncovery, as evidenced by reactor vessel water level readings, would be an indicator of a-situation where core. damage is likely.

, d) Some shorter lived isotope concentrations can be i' measured from the reactor water and containment gas samples. The ratios of these isotopes can be used to determine the source of the release n (fuel or gap).

,. e) Less volatile fission product concentrations such as isotopes of Sr, Ba, La, and Ru can be measured. If unusually high concentrations in the water samples are found, some degree of fuel I~ melting may be inferred.

The flow diagram in Attachment 1, indicates how the analysis based on radionuclide measurements, and the analysis of other significant parameters relates to the

estimation of core damage. l

p'-

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+ E . -- EMERGENCY PLAN IMPLEMENTING PROCEDURE DATE: 4/20/85 PAGE: 6 of 14 TITLE: REACTOR CORE DAMAGE ESTIMATION 2.0 RESPONSIBILITY 2.1 The Station Emergency Director is responsible for the implementation of this procedure.

2.2 The Supervisor-Emergency Planning is responsible for review of this procedure.

2.3 The Station Emergency Director with advice of his staff will determine sample locations and sample times.

2.4 The' Technical Assessment Supervisor is responsible for performing the calculations in this procedure leading to the. determination of core damage.

2.5 The_ Chemistry Department is responsible for obtaining samples from the PASS System and analyzing samples to determine fission product concentrations.

2.6 Radiation Protection is responsible for setting up radiological controls needed to obtain PASS Samples.

3.0 DEFINITIONS e

None 4.0 INSTRUCTIONS 4.l Core Damage Estimate From PASS 4.1.1 Obtain samples, consistent with Appendix H, from the Post Accident Sampling System, per RA-09, POST ACCIDENT SAMPLING. Record the Sample location, clock time, date, Drywell pressure,-

Containment pressure, Containment temp, Drywell temp, gas sample pressure, and gas sample temp on CORE DAMAGE ESTIMATE FROM PASS DATA SHEET (Attachment 2).

NOTE Data for this procedure may be obtained from the status boards or computer operator.

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. EMERGENCY PLAN IMPLEMENTING PROCEDURE DATE: 4/20/85 PAGE: 7 of 14 TITLE: REACTOR CORE DAMAGE ESTIMATION 4.1.2 Perform gamma spectroscopy per CPS No. 6103.01, GMKMA SPECTROMETER - GELI to determine the I-131 and Cs-137 concentrations in the water samples and Xe-133 and Kr-85 concentrations in the gas samples. Record the isotope concentrations on CORE DAMAGE ESTIMATE FROM PASS DATA SHEET (Attachment 2).

NOTE If the Gamma Spectroscopy System is setup to give concentrations of isotopes in liquids in units of/4Ci/ml, convert al to grams by Iml=lgm.

NOTE Measurements of Cs-137 and Kr-85 activities may not be possible until the reactor has been shut down for longer than a few weeks and most of the shorter lived isotopes have decayed.

4.1.3 Correct the measured gaseous activity concentration for temperature and pressure by:

T 2y C

gi = C .g1 (Vial) x Where:

l l C gt = Containment /Drywell isotopic l

concentration (gCi/cc)

C 1 (Vial) = Sample Vial isotopic Concentration (pCi/cc) l (Py ,T )

1

= Sample Vial pressure and i temperature on absolute scales

(*K, psia)

(P,,T )

' 2

= Containment /Drywell pressure and temperature on absolute scales

(*K, psia)

Record the calculated value for C on CORE -

DAMAGE ESTIMATE FROM PASS DATA SHEkT 1' (Attachment 2).

e 6

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] -1 EMERGENCY PLAN IMPLEMENTING PROCEDURE DATE: 4/20/85 PAGE: 8 of 14 TITLE: REACTOR CORE DAMAGE ESTDIATION 4.1.4 If the fission product concentrations are measured separately for the reactor water and suppression pool water or the drywell gas and the containment gas, the measured concentrations C or C should be averaged fromtheseparatemeasureMkntsbh Uyg = Average fission product concentration in CPS Coolant pCi/g)

Isotope Conc. 8 Isotope Conc.

U,1 = In Rx WTR (2.13 x 10 ) + In Supp Pool (4.09 x 10 9)

- (4.3 x 10')

U = Average fission product concentration in Containment gas ( Ci/cc) 8i l Isotope Conc 9 Isotope Conc.

-U = In Drywell (6.98 x 10 ) + ,In CNMT (3.7 x 1010) l 8 (4.398 x 10)

4 l and U on CORE

~

RecordthecalculatedvaluesU"bHEET,Ykttachment DAMAGE ESTRIATE FROM PASS DATA 2.)

4.1.5 Calculate the fission product inventory

, correction factor F 71 for I-131, Cs-137, Xe-133,

, and Kr-85 by:

F71 = Imentory in reference plant

Inventory in operating plant F ig = 3651 (1-e -1095AD i

i

- h iT) ) e- A1 1),

P)(1-e _

i Where:

P) = Steady period jreactor(MW )* power operated in t T) = Duration of operating period j (day)*  ;

T* = Time between the end of operating i 3 period j and time of last reactor l

shutdown (day) i h1 = Decay Constant for nuclide 1 (Day ~I)

• In each period, the variation of steady power should be limited to *20%.

, , ILLINOIS POWER COMPANY PROCEDURE: EC-13 CLINTON POWER STATION REVISION: 0 EMERGENCY PLAN IMPLEMENTING PROCEDURE DATE: 4/20/85 PAGE: 9 of 14 TITLE: REACTOR CORE DAMAGE ESTIMATION Record the calculated fission product inventory correction factors on CORE DAMAGE ESTIMATE FROM PASS DATA SHEET, Attachment 2.

NOTE See Appendix J for a sample calculation R

4.1.6 Calculatethenormalizeg*goncentrationCef for I-131 and Cs-137, and C g for Xe-133 anSiKr-85 by:

(*f =U yt x F 77 x (1.10)

C f = Ug '" i xF g x (1.10)

Where:

(*f=normalizedconcentrationofisotope i for reference plant coolant (pCi/g)

R C = normalized concentration of isotope 8 i for reference plant containment gas (pCi/cc)

Up = average concentration of isotope i in CPS coolant at time, t (pCi/g)

U = average concentration of isotope i 81 in CPS containment gas at time, t (mci /cc) hi = decay constant of isotope 1 (day-1) t = time between the reactor shutdown and the sample analysis time (day)

F yt = inventory correction factor for isotope i Record calculated normalized concentrations for i reference olant coolant and containment gas on CORE DAMAGt ESTIMATE FROM PASS DATA SHEET (Attachment 2),

l

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Upper Limit concentrations shown in Appendix B the extent of fuel or cladding damage can be ,

determined directly from A Record the estimated fuel /ppendix D through clad damage G.

on CORE DAMAGE ESTIMATE FROM PASS DATA SHEET (Attachment 2).

4.2 Hydrogen Analysis 4.2.1 Obtain a containment hydrogen and oxygen gas concentration reading from the Containment H 7 /0 Atmospheric Monitoring (CAM) system (%). The 2 reading to be selected should be based on engineering judgement. Record readings on HYDROGEN ANALYSIS DATA SHEET (Attachment 3).

NOTE The calculation for percent metal-water reaction is based on perfect hydrogen mixing in the containment. Gas concentration readings should be taken in the drywell and containment to verify that this is the case.

4.2.2 Calculate the Decimal Equivalent Metal-Water Reaction (MWR) Per: .

MWR = 2.422 KH2- (0 2 + 1.284 1

~ )(H2 -)(02 Where: TH2 = Percent hydro ~ gen concentration (decimal equivalent)

)(0 2 - Percent oxygen concentration (decimal equivalent)

Record calculated value for MWR on HYDROGEN ANALYSIS DATA SHEET (Attachment 3).

4.3 Drywell Radiation Analysis 4.3.1 Obtain Drywell Atmosphere Monitoring readings,

[R) in R/hr from 1RIX-CM059(1H13-P638) and IRIX-CM060(1H13-P639). Record readings on DRYWELL RADIATION ANALYSIS DATA SHEET (Attachment 4).

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l TITLE: REACTOR CORE DAMAGE ESTIMATION l

NOTE  :

The two drywell Monitor Readings should be averaged for use in Appendix K calculations. If the two readings are in disagreement by more than an order of magnitude a determination

! should be made as to the validity of the data.

4.3.2 Determine elaased time [T] in hours fron plant shutdown to the Drywell radiation monitor reading and. record on DRYWELL RADIATION ANALYSIS DATA SHEET (Attachment 4).

L 4.3.3 From Appendix K, determine the fuel inventory-release for the reference plant [I ref) %, and l record on DRYWELL RADIATION ANALYSIS DATA SHEET, (Attachment 4).

4.3.4 Determine the inventory release [I] to the CPS drywell using the following formular l

I = 1.3 X I ref .

' Record inventory release to the drywell on DRYWELL RADIATION ANALYSIS SHEET (Attachment 4).

4.4 Reactor Core Uncovery Time NOTE The graph of Maximum Acceptable Core Uncovery i Time Vs. Time After Reactor Shutdown in Appendix (L) was based on the time required for a com-pletely uncovered core to heat up from equi- ,

librium at 545'F to a peak clad temperature of '

2,200*F with no spray or steam cooling. If only L partial core uncovery or spray cooling occurred,  !

longer maximum acceptable core uncovery times are likely.

4.4.1 From the control room reactor water level instrumentation determine the length of time the reactor core was completely uncovered. Record this data on REACTOR CORE UNC0VERY TIME DATA 3 SHEET (Attachment 5). '

l

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• EMERGENCY PLAN IMPLEMENTING PROCEDURE DATE: 4/20/85 PAGE: 12 of 14 TITLE: REACTOR CORE DAMAGE ESTIMATION 4.4.2 Compare the actual core uncovery time determined in step 4.4.1 to the maximum acceptable core uncovery time obtained from Appendix (L) to determine if core damage is likely. Record the maximum acceptable core uncovery time from Appendix (L) on REACTOR CORE UNC0VERY TIMZ DATA SHEET (Attachment 5).

4.5 Identification of Release Source by Determination of Fission Product Ratios.

4.5.1 From samples obtained from PASS, determine the concentrations of the following short-lived isotopes by gamma ray spectroscopy per CPS No.

6103.01, GAMMA SPECTROMETER - GELI.

Liquid samples (ACi/g) Gas sample (yCi/cc)

! /

I-131 Kr-85m I-132 Kr-87 I-133 Kr-88 I-134 Xe-133 1-135 Record isotopic concentrations on FISSION PRODUCT RATIOS DATA SHEET (Attachment 6).

4.5.2 Correct the measured fission products to the time of reactor shutdown by C g,g =

>L' C 1,t

• Where:

C i,

= concentration of isotope i at shutdown. (nCi/g) or (/Ci/cc)

C 1,t = measured concentration of isotope i at time t. (pCi/g) or (Ati/cc) 2Ng = decay constant of isotope 1 (day'1).

t = time between reactor shutdown and sample analysis (day).

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.. EMERGENCY PLAN IMPLEMENTING PROCEDURE DATE: 4/20/85 PAGE: 13 of 14 TITLE: REACTOR CORE DAMAGE ESTIMATION Record the Corrected fission product concen-trations on FISSION PRODUCT RATIO DATA SHEET (Attachment 6).

4.5.3 Calculate the isotopic ratios by:

Noble gas ratio = Noble gas isotopic concentration Xe-133 concentration Iodine ratio = Iodine isotopic concentration 1-131 concentration Record the fission product ratios on FISSION PRODUCT RATIO DATA SHEET (Attachment 6).

4.5.4 Determine the release source by comparing the isotopic ratios from step 4.5.3 to the ratios supplied in Appendix C. Record the release source determined by each ratio on FISSION PRODUCT RATIO DATA SHEET (Attachment 0).

NOTE Generally, lower fission product activity ratios are found in the fuel gap, so lower fission product ratios measured in CPS coolant or containment atmosphere is indicative of fuel cladding failure. Higher fission product activity ratios are found in the core fuel, and h*.gher fission product activity ratios are indicative of fuel melt.

4.6 Analysis for Ba, Sr, La, Ru.

4.6.1 From samples obtained from PASS determine the concentrations of the following short-lived isotopes by gamma ray spectroscopy:

Sr-91 Ci/g)

Sr-92 Ci/g)

Ba-140 Ci/g)

La-140 Ci/g)

Ru-103 QNCi/g)

Record the isotope concentrations on ANALYSIS FOR Ba, Sr, La and Ru DATA SHEET (Attachment 7).

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. . EMERGENCY PLAN IMPLEMENTING PROCEDURE DATE: 4/20/85 PAGE: 14 of 14 TITLE: REACTOR CORE DAMAGE ESTIMATION 4.6.2 Compare the isotopic concentrations obtained in step 4.6.1. to the baseline concentration data maintained by the Chemistry Department. If unusually high concentrations of Sr, Ba, La and Ru are found in the water samples (i.e., greater than 100% above baseline), some degree of fuel melting may be inferred.

5.0 REFERENCES

1. NED0-22215 Procedures For The Determination of the Extent of Core Damage Under Accident Conditions.

Dated: August 1982

2. NEDO-22215 Attachment 8D, Procedures for Estimating Core Damage Based on Plant Parameters Other than Post-Accident Sampling System Measurements.

3. BWROG Emergency Procedures Guidelines, Rev. 2 Appendix C: Calculational Procedures, C23.0 Maximum Acceptable Core Uncovery Time.

4. RA-09, POST ACCIDENT SAMPLING

5. CPS No. 6103.01, GAMMA SPECTROMETER-GELI

6. FSAR APPENDIX, TABLE D-1, RADIOACTIVE SOURCE ASSUMPTIONS, (AMENDMENT 14).

6.0 ATTACHMENTS

1. SEQUENCE OF ANALYSIS FOR ESTIMATION OF CORE DAMAGE

2. CORE DAMAGE ESTIMATE FROM PASS DATA SHEET

3. HYDROGEN ANALYSIS DATA SHEET

4. DRYWELL RADIATION ANALYSIS DATA SHEET

5. REACTOR CORE UNC0VERY TIME DATA SHEET  ;

6. FISSION PRODUCT RATIO DATA SHEET

7. ANALYSIS FOR Ba, Sr, La, and Ru

--_,,_-m --

. , . _ __ ,_ y 5

_ m m

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EMERGENCY PLAN IMPLEMENTING PROCEDURES . ATTACHMENT: 1-PAGE: 1 of 1 SEQUENCE OF ANALYSIS FOR ESTIMATION OF CORE DAMAGE ~

I

, Hydrogen Yes m Drywell Yen h Water Yes~ m NORMAL OPERATION

- Analysis , F Radiation , Level F MINOR CLAD DAMAGE 3

o (Confirm)p (Confirm) N (Confirm'J A

o Core Damage *

}

termine timum H Estimate W, '

j ample From PASS  ? n n )[

oint g

w E S O

:c

!i Hydrogen Yan 6 Drywell Yes i Water Yes , Analysis for

Analysis . Radiation a "

Level v Ba, Sr, La, Ru l (Confirm) '

(Confirm)y (Confirm) i 1F l MAJOR CLAD DAMAGE Determination i FUEL OVERHEAT *Yes Of Fission FUEL MELT Product Ratios i 0 z

! 1 r CLAD DAPAGE i

POSSIBLE FUEL OVERHEAT NO-CORE MELT l

l E _____ -

,, ILLINOIS POWER COMPANY PROCEDURE: EC-13 CLINTON POWER COMPANY REVISION: 0 EMERGENCY PLAN IMPLEMENTING PROCEDURE ATTACHMENT: 2 PAGE: 1 of 3 CORE DAMAGE ESTIMATE FROM PASS DATA SHEET Sample Location Check Appropriate Sample Analysis (Remarks)

Location Time Time Jet Pump RWCU Supp. Pool Liquid Containment

• Atmosphere RHR Drywell Drywell Pressure (psia)

Drywell Temp. (*K)

Containment Pressure (psia)

Containment Temp (*K)

Containment Atmos Sample Vial Pressure (psia)

Containment Atmos Sample Vial Temp (*K)

Drywell Sample Vial Pressure (psia)

Drywell Sample Vial Temp (*K)

Isotopic Concentration of Sample Samp12 Location Isotope Concentration Isotope Concentration Jet Pump I-131 (J/Ci/g) Cs-137 (j/Ci/g)

RWCU I-131 9/Ci/g) Cs-137 (//Ci/g)

Supp. Pool Liquid I-131 p/Ci/g) Cs-137 (uCi/g)

Containment Atmos Xe-133 JJCi/ce) Kr-85 (pci/cc) l RHR I-131 y/Ci/g) Cs-137 (//Ci/g)

Drywell Xe-133 (j/ci/cc) Kr-85 (t/Ci/cc)

L

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C g g

Containment Atmos Sample (ACi/cc) (pCi/cc)

C gt: Drywell Atmos Sample (pCi/cc) (pCi/cc)

Average Fission Product Concentrations J

twI-131 : (Averngo I-131 Concentration in CPS Coolant)= gCi/g) twCs-137 :(Average Cs-137 Concentration in CPS Coolant)= (pCi/g)

UgXe-133 :(Average Xe-133 Concentration in CNMT gas)= @Ci/cc)

U gKr-85 : (Average Kr-85 Concentration in CNMT gas)= (pCi/cc)

Fission Product Correction Factor FII-131: (I-131 Fission product Correction Factor) =

FICs-137: (CS-137 Fission product Correction Factor) =

FIXE-133: (Xe-133 Fission product Correction Factor) =

FIKr-85: (Kr-85 Fission product Correction Factor) =

Normalized Fission Product Concentrations for Reference Plant Ch{f131: (I-131 Conc. in reference plant coolant) = (pci/g)

C ,R - 37: (Cs-137 Conc. in reference plant coolant) = WCi/g)

Ref pCi/cc)

C e-133: (XE-133 Conc. in reference plant CNMT gas) =

(gCi/cc)

Cf -85: (Kr-85 Conc. in reference plant CNMT gas) =

Estimated Fuel / Clad Damage Cladding Failure Fuel Meltdown Appendix D estimate  %  %

d Appendix E estimate  %  %

, Appendix F estimate  %  %

Appendix G estimate  %  %

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Time /Date Completed:

1 L-

,, ,IELINOIS POWER COMPANY PROCEDURE: EC-13 CLINTON POWER COMPANY REVISION: 0 EMERGENCY PLAN IMPLEMENTING PROCEDURE ATTACHMENT: 3 PAGE: 1 of 1 HYDROGEN ANALYSIS DATA SHEET Containment Hydrogen and Oxygen Gas Concentration Containment Drywell 0

2 Conc. =  %  %

H Conc =  %  %

2 Metal - Water Reaction MWR =

Remarks Performed By:

Time /Date:

,, . ILLINOIS POWER COMPANY PROCEDURE: EC-13 CLINTON POWER COMPANY REVISION: 0 EMERGENCY PLAN IMPLEMENTING PROCEDURE ATTACHMENT: 4 PAGE: 1 of 1 DRYWELL RADIATION ANALYSIS DATA SHEET ,

Drywell Radiation Radiation 7 Monitor Detector IDf Reading (R/hr) 1 RIX-CM059(1H13-P638) 1 RIX-CM060(1H13-P639)

I Elapsed time from plant Shutdown to Drywell radiation monitor reading (hours)

Inventory release for reference Plant determined from App. K (%)

Calculated inventory release to CPS Drywell (%)

Remarks Performed By:

Time /Date:

r i t

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, EMERGENCY PLAN IMPLEMENTING PROCEDURE ATTACHMENTS 5 PAGE: 1 of 1 REACTOR CORE UNC0VERY TIME DATA SHEET Control Room Reactor Core Instrument IDI Uncovery Time Maximum Acceptable Core Uncovery Time from App. (L).

Remarks Performed By:

Time /Date

• IbLINOIS POWER COMPANY PROCEDURE: EC-13

'CLINTON POWER COMPANY REVISION: 0 EMERGENCY PLAN IMPLEMENTING PROCEDURE ATTACHMENT: 6 PAGE: 1 of 2 l FISSION PRODUCT RATIO DATA SHEET Isotopic Concentrations Liquid Sample Concentration Gas Sample Concentration (Isotope) (mci /g) (Isotope) (aCi/cc)

I-131 Kr-85m I-132 Kr-87 I-133 Kr-88 1-134 Xe-133 I-135 i

Corrected Fission Product Concentrations Liquid Sample Concentration Gas Sample Concentration (Isotope) (uCi/g) (Isotope) (aci/cc) 1-131 Kr-85m I-132 Kr-87 I-133 Kr-88 I-134 Xe-133 I-135 Isotopic Ratios Noble Gas Ratios Iodine Ratios ,

Kr-85m . I-132 ,

Xe-133 1-131 Kr-87 I-133 ,

Xc-133 , 1-131 Kr-88 , I-134 ,

Xe-133 '

I-131 1-135 I-131 "

,' . ILLINOIS POWER COMPANY PROCEDURE: EC-13 CLINTON POWER COMPANY REVISION: 0 EMERGENCY PLAN IMPLEMENTING PROCEDURE ATTACHMENT: 6 PAGE: 2 of 2 Release Source Release Source Release Source Ratio (Core Inventory / Fuel Gap) Ratio (Core Inventory / Fuel Gap)

Kr-85m . I-132 ,  !

Xe-133 I-131 i Kr-87 , I-133 ,

Xe-133 1-131 Kr-88 , I-134 ,

Xe-133 1-131 t-I-135 ,

1-131 Performed By:

Time /Date: <

t 1

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Sr-91

( Sr-92 Ba-140 La-140 Ru-103 Performed By:

Time / Dater

, ILLINOIS POWER COMPANY PROCEDURE 8 EC-13 CLINTON POWER STATION REVISION: 0 EMERGENCY PLAN IMPLEMENTING PROCEDURE APPENDIXs A PAGE: 1 of 1 TITLE: REACTOR CORE DAMAGE ESTIMATION CORE INVENTORY OF MAJOR FISSION PRODUCTS IN A REFERENCE PLANT OPERATED AT 3651 MW FOR THREE YEARS

, Major Gamma Ingentory** Ray Energy)

(Intensity Chemical Group Isotope

• Half-Life 10 Ci kev (T/d) 1 Noble gases Kr-85m 4.48h 24.6 151(0.753) i Kr-85 10.72y 1.1 514(0.0044) i Kr-87 76.3m 47.1 403(0.495)

Kr-88 2.84h 66.8 196(0.26),1530(0.109)

Xe-133 5.25d 202.0 81(0.365) ,

Xe-135 9.11h 26.1 250(0.899) l Halogens I-131 8.04d 96.0 364(0.812) 1-132 2.3h 140 668(0.99,773(0.762) 1-133 20.8h 201 530(0.86) .

1-134 52.6m 221 847(0.954),884(0.653)

I-135 6.61h 189 1132(0.225),1260(0.286) 1 Alkali Metals Cs-134 2.06y 19.6 605(0.98),796(0.85)

Cs-137 30.17y 12.1 662(0.85)

Cs-138 32.2m 178.0 463(0.307),1436(0.76)

Tellurium Group Te-132 78.2h 138 228(0.88) ,

i Noble Metals Mo-99 66.02h 183 740(0.128)

Ru-103 39.4d 155 497(0.89)

Alkaline Earths Sr-91 9.5h 115 750(0.23),1024(0.325)

Sr-92 2.71h 123 1388(0.9)

Ba-140 12.8d 173 537(0.254)

Rare Earths Y-92 3.54h 124 934(0.139)

La-140 40.2h 184 487(0.455),1597(0.955)

Ce-141 32.5d 161 145(0.48)

Ce-144 284.3d 129 134(0.108)

Refractories Zr-95 64.0d 161 724(0.437,757(0.553) -

Zr-97 ,

16.9h 166 743(0.928)  !

l

• 0nly the representative isotopes which have relatively large inventory and considered to be easy to measure are listed here.

• • At the time of reactor shutdown.

, , ILLINOIS POWER COMPANY PROCEDURE: EC-13 CLINTON POWER STATION REVISION: 0

. EMERGENCY PLAN IMPLEMENTING PROCEDURE APPENDIX: B PAGE: 1 of 1 TITLE: REACTOR CORE DAMAGE ESTIMATION FISSION PRODUCT CONCENTRATIONS IN REACTOR WATER AND DRYWELL GAS SPACE DURING REACTOR SHUTDOWN UNDER NORMAL CONDITIONS Reactor Water, (WCi/g) Drywell Gas (yCi/cc)

Isotope Upper Limit Nominal Upper Limit Nominal I-131 29 0.7 --- ---

Cs-137 c 0.3 a 0.03 b ___ _ _ ,

-0" -5b Xc-133 --- --- 1x10 1x10 Kr-85 --- --- 4x10-5a 4x10-6b C

0bserved experimentally, in an operating BWR-3 with MK I containment data obtained from GE unpublished document, DRF 268-DEV-0009.

b Assuming 10% of the upper limit values, c

Release of Cs-137 activity would strongly depend on the core inventory which is a function of fuel burnup.

l

ILLINOIS POWER COMPANY PROCEDUREe EC-13

' 'CLINTON POWER STATION REVISION: 0 EMERGENCY PLAN IMPLEMENTING PROCEDURE APPENDIX: C PAGE: 1 of 1 TITLE: REACTOR CORE DAMAGE ESTIMATION RATIOS OF ISOTOPES IN CORE INuENTORY AND FUEL CAP Activity Ratio in Activity Ratio in Isotope Half-Life Core Inventory Fuel Gap Kr-87 76.3m 0.233 0.0234 Kr-88 2.84h 0.33 0.0495 Kr-85m 4.48h 0.122 0.023 Xe-133 5.25d 1.0 1.0 1-134 52.6m 2.3 0.155 I-132- 2.3h 1.46 0.127 I-135 6.61h 1.97 0.364 I-133 20.8h 2.09 0.685 I-131 8.04d 1.0 1.0 Ratio (for noble gases) , noble gas isotope concentration Xe-133 concentration Ratio (for iodines) = Iodine isotope concentration I-131 concentration

ILLINOIS POWER COMPANY PROCEDURE: EC-13

'CLINTON POWER STATION REVISION: 0 EMERGENCY PLAN IMPLEMENTING PROCEDURE APPENDIX: D PAGE: 1 of 1 TITLE: REACTOR CORE DAMAGE ESTIMATION 4

RELATIONSHIP BETWEEN I-131 CONCENTRATION IN THE PRIMARY COOLANT (REACTOR WATER + POOL WATER) AND THE EXTENT OF CORE DAMAGE IN REFERENCE PLANT tsp ',

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' PROCEDURE: EC-13

. 'CLINTON POWER STATION REVISION: 0 EMERGENCY PLAN IMPLEMENTING PROCEDURE APPENDIX: E PAGE: 1 of 1 TITLE: REACTOR CORE DAMAGE ESTIMATION RELATIONSHIP BETWEEN Cs-137 CONCENTRATION IN THE PRIMARY COOLANT (REACTOR WATER + POOL WATER) AND THE EXTENT OF CORE DAMAGE IN REFERENCE PLANT i W.

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• ILLINOIS POWER COMPANY PROCEDURE: EC-13 l CLINTON POWER STATION REVISION: 0 l EMERGENCY PLAN IMPLEMENTING PROCEDURE APPENDIX: F , -

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REACTOR CORE DAMAGE ESTIMATION l

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, ILLIN0IS POWER COMPANY PROCEDURE: Er 13

'CLINTON POWER STATION REVISION: 0 EMERGENCY PLAN IMPLEMENTING PROCEDURE APPENDIX: G PAGE: 1 of 1 TITLE: REACTOR CORE DAMAGE ESTIMATION RELATIONSHIP BETWEEN Kr-85 CONCENTRATION IN THE CONTAINMENT GAS (DRYWELL + PRIMARY CONTAINMENT) AND THE EXTENT OF CORE DAMAGE IN REFERENCE PLANT tei" ,

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. -1 ILLINOIS POWER COMPANY PROCEDURE: EC-13 .--

.CLINTON POWER STATION REVISION: 0 . 1 EMERGENCY PLAN IMPLEMENTING PROCEDURES APPENDIX: H. ,

PAGE:- 1 of 2 i

TITLE: REACTOR CORE DAMAGE ESTIMATION SAMPLES MOST REPRESENTATIVE OF CORE CONDITIONS DURING AN ACCIDENT FOR THE ESTIMATION OF CORE DAMAGE Break CategoryISystem Conditions Sample Location Other Instructions Supp. Supp.

Jet Pool Pool Pump Liquid Atmos. RHR Drywell Small Liquid Line Break, Reactor 1 2 Power hlI Yes ---

Yes --- Yes small .iquid Line Break, Keactor Power 411 --- ---

Yes Yes Yes 2 RHR must be in shutdown cooling mode.

Reactor water level must be raised and flow from moisture separators.

Small Steam Line Break, Reactor Power >1% Yes ---

fes l ---

Yes 2 Small stean Line Break, Keactor 1 Yes Yes 2 RHR must be in shutdown cooling '

Power <1% --- ---

fes mode.

Reactor water level must be raised and flow from moisture separators.

Large Liquid Line Break, Reactor Power >1I. Yes 3 Yes 4

Yes 1 --- Yes 2 Suppression pool must be in suppression cooling mode.

Large Liquid Line Break, Reactor 4 2 Power (1Z --- Yes Yes 1 Yes 3 Yes RHR must be in shutdown cooling mode.

i Suppression pool must be in l suppression cooling mode.

l Reactor water level must be l raised and flow from moisture j sepatators.

l ILLINOIS POWER COMPANY PROCEDURE: EC-13 CLINTON POWER STATION REVISION: 0 -

EMERGENCY PLAN IMPLEMENTING PROCEDURES APPENDIX: H PAGE: 2 of 2 TITLE: REACTOR CORE DAMAGE ESTIMATION SAMPLES MOST REPRESENTATIVE OF CORE CONDITIONS DURING AN ACCIDENT FOR THE ESTIMATION OF CORE DAMAGE Break Category / System Conditions' Sample Location Other Instructions Supp. Supp.

Jet Pool Pool Pump Liquid Atmos. RHR Drywell Large Steam Line Break, Reactor Power 21% Yes 3 Yes 4 --- ---

Yes Large Steam Line Break, Keactor Power <1% --- ---

Yes 1 Yes Yes 2 RHR must be in shutdown cooling mode.

Reactor water level must be raised and flow from moisture separators.

1. Use if SRV's are vented to suppression pool.

2. Use if SRV's are not vented to suppression pool.

3. Use if makeup water flow is 4 50% of core flow present.

4. Useifmakeupwaterflowis])50%ofcoreflowpresent.

"' . ILLINOIS POWER COMPANY PROCEDURE: EC-13 CLINTON POWER STATION REVISION: 0 EMERGENCY PLAN INPLEMENTING PROCEDURE APPENDIX: I PAGE: 1 of 1 TITLE: REACTOR CORE DAMAGE ESTIMATION PLANT PARAMETERS PRIMARY COOLANT

• CONTAINMENT CAS**

PRIMARY RATED REACTOR SUPPRESSION DRYWELL CONTAINMENT REACTOR TYPE / CON- POWER WATER MASS 8 POOLgATER CAS9 M. CAS y M PLANT TAINMENT DESIGN (MW) (10 g) (10 g) (10 cc) (10 cc)

CPS BWR6/MKIII 2894 2.13 4.09 6.98 37.00

• Total Primary Coolant Mass = Reactor Water + Suppression Pool Water

• • Total Containment Gas Volume = Drywell Gas + Primary Containment

'*,,, ILLINOIS POWER COMPANY PROCEDURE: EC-13 CLINTON POWER STATION REVISION: 0

~

EMERGENCY PLAN INPLEMENTING PROCEDURE APPENDIX: J PAGE: 1 of 2 TITLE: REACTOR CORE DAMAGE ESTIMATION SAMPLE CALCULATION OF FISSION PRODUCT INVENTORY CORRECTION FACTOR F = Inventory of nuclide i in reference plant Ii Inventory of nuclide i in operating plant

= 3651 (1-e -1095ht )

T '"

- htT))e->t j P)(1-e where P) = steady reactor power operated in period j (MWt) ht = decay constant of nuclide 1 (day-1)

Tj = duration of operating period j (day)

T = time between the end of operating period j and time of last reactor shutdown (day) 3651 = ave, operation power (in MWt) for the reference Station.

1095 = continuous operation time (in day) for the reference Station.

Assuming a reactor has the following power operation history:

Operation Time Average Power Operation T' P Period Days since startup T3 (day) 3 3 ggy )

1A 1 - 60 60 254 1000 IB 61 - 70 --- --- 0 2A 71 - 270 200 44 2000 2B 271 - 300 --- --- 0 3 301 - 314 14 0 3000 e '

k_ _

i , ILLINOIS POWER COMPANY PROCEDURE: EC-13

,' CLINTON POWER STATION REVISION: 0 EMERGENCY PLAN INPLEMENTING PROCEDURE APPENDIX: J PAGE: 2 of 2 TITLE: REACTOR CORE DAMAGE ESTIMATION For I-131 ("A = 0.0862 day "I) 3651(1-e-0.0862x1095)

FI(I-131) = 1000(1-e-0.0562x60),-0.0562xzs4+2000(1-e-0.0562xzoo) e -0.0562x44 + 3000(1-e-0.0562x14),-0.0562x0

= 3651 = 1.7 0 + .45 + 2103 For Cs-137 (A = 6.29 x 10-5 day ~1) 3651(1-e-6.29x10-5x1095) ,

yI(Cs-137 , 1000(1-e-6.29x10 'x60)e-6.29x10 'x254

+ 2000(1-e-6.29x10~5x200),-6.29x10~5 x44

+3000(1-e-6.29x10-5x14),-6.29x10-5 x0

= 243.16 " 7*77 3.74 + 24.93 + 2.64

3 g 6 e ILLINOIS POWER COMPANY PROCEDURE: EC-13 CLINTON POUER STATION REVISION: 0 EMERGENCY PLAN INPLEMENTING PROCEDURE APPENDIX: K PAGE: 1 of 2 TITLE: REACTOR CORE DAMAGE ESTIMATION PERCENT OF FUEL INVENTORY AIRBORNE IN THE DRYWELL t

10- "

• ' l00t, F UJ L INVENTORY = 10JD) NOB LE S '

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TIM E A F T E R. ' 5NUTDOWN (MK) io' _

Inventory Approximate Source and Damage Estimate l Released 100. 100% TID-14844, 100% fuel damage, potential core melt.

50. 50% TID noble gases, TMI source.

10. 10% TID, 100% NRC gap activity, total clad failure, t

partial core uncovered.

3. 3% TID,100% WASH-1400 gap activity, major clad  !

failure.

1. 1% TID, 10% NRC gap, Max. 10% clad failure.

! .1 .1% TID, 1% NRC gap, 1% clad failure, local heating of 5-10 fuel assemblics.

_ _ _ - - -----.--m

?. ILLINOIS POWER COMPANY PROCEDURE: EC-13 y

. CLINTON POWER STATION REVISION: 0 EMERGENCY PLAN INPLEMENTING PROCEDURE APPENDIX: K PAGE: 2 of 2 TITLE: REACTOR CORE DAMAGE ESTIMATION

.01 .01% TID, 1% NRC gap, clad failure of 3/4 fuel element (36 rods).

- 10-3 .01% NRC gap, clad failure of a few rods.

10~4 100% coolant release with spiking.

4 5x10-6 100% coolant inventory release.

10-6 Upper range of normal airborne noble gas activity in containment.

}

r ILLINOIS POWER COMPANY PROCEDURE: EC-13

, #* g yCLINTON POWER STATION REVISION: 0

. EMERGENCY PLAN INPLEMENTING PROCEDURE APPENDIX: L PAGE: 1 of 1 TITLE: REACTOR CORE DAMAGE ESTIMATION MAXIMUM ACCEPTABLE CORE UNC0VERY TIME VS. TIME AFTER REACTOR SHUTDOWN 90 _q _ .

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