Text

Rev 1 to Station Procedure OPEP02-ZG-0007, Coolant Activity & Radionuclide Trend for Failed Fuel
	ML20211Q256
Person / Time
Site:	South Texas
Issue date:	12/01/1986
From:	; HOUSTON LIGHTING & POWER CO.
To:	;
Shared Package
ML20211Q176	List: ML20211Q256; ST-HL-AE-1838, Forwards Rev 1 to Station Procedure OPEP02-ZG-0007 Re nonsafety-related Coolant Activity & Radionuclide Trend for Failed Fuel,Per SER Confirmatory Item 18 & NUREG-0737,Item II.B.3 Re post-accident Sampling Sys;
References
	NUDOCS 8612190228
	Download: ML20211Q256 (55)
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3 HOUSTON LIGHTING AND POWER COMPANY SOUTH TEXAS PROJECT ELECTRIC GENERATING STATION PLANT PROCEDURES MANUAL STATION PROCEDURE NON SAFETY-RELATED (0)

Coolant Activity and Radionuclide Trend for OPEP02-ZG-0007 Failed Fuel Rev. 1(General)

Page 1 of 55 APPROVED:

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PLANT MANAGER h

DATE APPROVED DATE EFFECTIVE Table of Contents Earn

1.0 Purpose and Scope

4 2.0 Prerequisites 4

3.0 Precautions 4

4.0 RCS Radionuclide/ Failed Fuel Trend 5

5.0 Normal Operations Failed Fuel Estimates 7

6.0 Estimation of Failed Fuel During Post-Accident Conditions 7

7.0 Acceptance Criteria 9

8.0 Documentation 10 9.0 References 10 10.0 Support Documents 11 10.1 Addendum 1 - Determination of Power Correction 13 Factor, X 10.2 Addendum 2 - Determination of Pressure - Temperature 16 Correction Factor, Z 8612190228 861215 PDR ADOCK 05000498 E

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Coolant Activity and Radionuclide Trend for OPEP02-ZG-0007 Failed Fuel Rev. 1 Page 2 of 55 10.3 Addendum 3 - Determination of Sump Water Volume 17 10.4 Addendum 4 - Estimation of Percent Fuel 18 Overtemperature and Percent Fuel Melt 10.5 Addendum 5 - Estimation of Percent Clad Damage and 22 Percent Fuel Overtemperature 10.6 Addendum 6 - Post-Accident Data Retrieval Locations 24 10.7 Addendum 7 - Auxiliary Indicators of Core Damage 25 10.8 Figure 1 - Power Correction Factor for I-131 Based 27 on Average Power During Operation 10.9 Figure 2 - Relationship of Percent Clad Damage With 28 Percent Core Inventory Released of I-131 10.10 Figure 3 - Relationship of Percent Clad Damage With 29 Percent Core Inventory Released of I-131 With Spiking 10.11 Figure 4 - Relationship of Percent Clad Damage With 30 Percent Core Inventory Released of I-132 10.12 Figure 5 - Relationship of Percent Clad Damage With 31 Percent Core Inventory Released of I-133 10.13 Figure 6 - Relationship of Percent Clad Damage With 32 Percent Core Inventory Released of I-135 10.14 Figure 7 - Relationship of Percent Clad Damage With 33 Percent Core Inventory Released of Kr-87 10.15 Figure 8 - Relationship of Percent Clad Damage With 34 Percent Core Inventory Released of Xe-131m 10.16 Figure 9 - Relationship of Percent Clad Damage With 35 Percent Core Inventory Released of Xe-133 10.17 Figure 10 - Relationship of Percent Fuel 36 Overtemperature With Percent Core Inventory Released of Xe, Kr, I, or Cs 10.18 Figure 11 - Relationship of Percent Fuel 37 Overtemperature With Percent Core Inventory Released of Ba-140

Coolant Activity and Radionuclide Trend for OPEP02-ZG-0007 Failed Fuel Rev. 1 Page 3 of 55 10.19 Figure 12 - Percent Zirc-Water Reaction Based on Hydrogen Concentration 38 10.20 Figure 13 - Percent Noble Gas Released 39 10.21 Data Sheet 1 - Normal Operations Radionuclide Trend 40 10.22 Data Sheet 2 - Normal Operations Failed Fuel Estimate 42 Based on I-131 Activity 10.23 Data Sheet 3 - Post-Accident Specific Activity 43 Determination 10.24 Data Sheet 4 - Post-Accident Gross Activity 48 Determination 10.25 Data Sheet 5 - Percent Fuel Overtemperature and 51 Percent Fuel Melt Estimate 10.26 Data Sheet 6 - Percent Clad Damage and Percent Fuel 53 Overtemperature Estimates 10.27 Data Sheet 7 - Evaluation of Auxiliary Indicators 54 l

Coolant Activity and Radionuclide Trend for OPEP02-ZG-0007 Failed Fuel Rev. 1 Page 4 of 55

1.0 Purpose and Scope

The purpose of this procedure is to provide instructions and guidelines for the following:

1.1 Obtaining the required data for the determination of defective fuel.

1.2 Trending of RCS radionuclide activities during normal at-power conditions for the prediction of failed fuel occurrences.

1.3 Estimating the amount of failed fuel during normal operating conditions.

1.4 Estimating the amount of failed fuel during post-accident conditions.

2.0 Prerequisites 2.1 Plant power history for at least the previous 30 days is available.

3.0 Precautions 3.1 The numbers obtained by using this procedure are rough estimates, at best.

Results of this procedure cannot be used to describe or infer the exact type and amount of fuel damage.

3.2 All equations quoted in this procedure are based on equilibrium full power iodine.

Iodine spiking phenomena following transient operations may provide false fuel damage estimates.

3.3 Sample collection, preparation, and analysis will be performed under applicable Radiation Work Permits.

NOTE If this procedure is being used to trend for failed fuel under normal operations, perform section 4.0.

If this procedure is being used to estimate the extent of failed fuel under post-accident conditions, perform section 6.0.

1 A

Coolant Activity and Radionuclide Trend for OPEP02-ZG-0007 Failed Fuel Rev. 1 Page 5 of 55 4.0 ECS Radionuclide/ Failed Fuel Trend 4.1 Perform the following, ensuring that the sample required in Step 4.1.1.1 is taken at approximately the same time that the data required in Step 4.1.2 is recorded.

4.1.1 Request Chemical Analysis to perform the following:

4.1.1.1 Obtain a liquid sample of the Reactor Coolant System (RCS) in accordance with appropriate sampling procedures.

4.1.1.2 Enter the following in Section 1.0 of Data Sheet 1

(-1):

a.

Sample I.D.

b.

Sample type c.

Date and time of sample d.

Reactor power at time of sample e.

Initials 4.1.1.3 Determine the dissolved noble gas activity of a reactor coolant sample in accordance with OPCP09-ZR-0004 (Determination of Radionuclides by Gamma Spectroscopy).

-4.1.1.4 Determine the gamma activity of a degassed reactor-coolant sample in accordance with OPCP09-ZR-0004 (Determination of Radionuclides by Gamma Spectroscopy).

4.1.1.5 Enter the decay corrected activities and the date and time of analysis in Section 1.0 of Data Sheet 1 (-1).

4.1.1.6 Enter Counting Instrument used/ID # in Section 1.0 of Data Sheet 1 (-1).

4.1.1.7 Sign and date Section 1.0 of Data Sheet 1 (-1).

4.1.1.8 Forward a copy of Data Sheet 1 (-1) to the Lead i

chemical Technician for review.

4.1.2 Complete Section 2.0 of Data Sheet 1 (-1).

Coolant Activity and Radionuclide Trend for OPEP02-ZG-0007 Failed Fuel Rev. 1 Page 6 of 55 4.2 Transcribe the decay-corrected specific activities from Section 1.0 of Data Sheet 1 (-1) to Section 3.0 of Data Sheet 1 (-1).

4.3 Using Addendum 1 and available plant power history data, determine the Power Correction Factor, X, for each isotope listed on Data Sheet 1 (-1).

Record these values in Section 3.0 of Data Sheet 1

(-1).

4.4 Using the equation presented in Section 3.0 of Data Sheet 1 (-1),

calculate the Adjusted Specific Activity for each isotope listed on Data Sheet 1 (-1).

Record these values on Data Sheet 1 (-1).

4.5 Using the equation presented in Section 3.0 of Data Sheet 1 (-1),

calculate the I-131/I-133 Ratio. Record this value on Data Sheet 1 (-1).

NOTE For the purposes of this procedure, " Base-line Specific Activity" is defined as the value determined by Reactor Performance Section to be the current normal operating specific activity present in the RCS for a given isotope.

4.6 Obtain from Reactor Performance Section and record in Section 3.0 of Data Sheet 1 (-1) the latest base-line specific activity data for each isotope listed on Data Sheet 1 (-1).

l I

NOTE If reactor power has varied greater than + 5%

from the time-average power value within the past 60 hours6.944444e-4 days <br />0.0167 hours <br />9.920635e-5 weeks <br />2.283e-5 months <br />, iodine spiking phenomena may occur which will provide false fuel failure indication for iodine isotopes.

l 4.7 If any Adjusted Specific Activity exceeds its base-line value, l

fuel failure may have occurred. Notify the Reactor Performance Supervisor and the Chemical Analysis Supervisor and proceed to Step 5.0.

If no Adjusted Specific Activity exceeds its base-line value, proceed to Step 7.0.

Coolant Activity and Radionuclide Trend for OPEP02-ZG-0007 Failed Fuel Rev. 1 Page 7 of 55 5.0 Normal Operations Failed Fuel Estimates NOTE The failed fuel estimate equations used in this determination are based on operational experience data from M PWR plants. The results obtained from this section are at best rough estimates only, but should provide information concerning the general trends of fuel failure.

5.1 Using the Adjusted Specific Activity for I-131 recorded on Data Uheet 1 (-1), complete Data Sheet 2 (-2).

5.2 Inform the Reactor Performance Supervisor of the failed fuel estimates determined in Step 5.1.

5.3 Proceed to Step 7.0.

6.0 Estimation of Failed Fuel Durine Post-Accident Conditions NOTE The estimation of failed fuel during post-accident conditions is based on a radionuclide analysis of RCS, containment atmosphere, and containment sump samples and the evaluation of core conditions using auxiliary indicators (core exit thermocouples, reactor vessel water level, containment hydrogen concentration, and containment radiation level).

The auxiliary indicators cannot by themselves provide a useful estimate of the amount of core damage but can determine if conditions are consistent with those expected to be associated with core damage. While auxiliary indicators are intended to provide confirmation of the failed fuel estimate based on radionuclide analysis, the auxiliary indicators may be used l

to obtain an initial assessment of whether or not core damage is likely to have occurred.

=

3 e

Coolant Activity and Radionuclide Trend for OPEP02-ZG-0007 Failed Fuel Rev. 1 Page 8 of 55 NOTE Step 6.1 is not required to be completed prior to performing the remainder of section 6.0.

6.1 Use the guidelines contained in Addendum 7 to evaluate the possibility of core damage based on auxiliary indicators and complete Data Sheet 7.

6.2 Perform the following, ensuring that the samples. required in Step 6.2.1 are taken at approximately the same time that the data required in Step 6.2.2 is recorded.

6.2.1 Request Chemical Analysis to perform the following:

6.2.1.1 Obtain a sample of the Reactor Coolant System (RCS), the Containment Sump, and the Containment Atmosphere in accordance with OPCP08-AP-0003 (Sampling and Analysis of RCS, RHR, and RCB Sump -

Post Accident) and OPCP08-AP-0004 (Sampling and Analysis of the RCB Atmosphere - Post Accident).

6.2.1.2 Enter the following in Section 1.0 of Data Sheet 3

(-3):

a.

Sample I.D.

b.

Sample types c.

Date and time of samples d.

Reactor power at time of samples e.

Initials 6.2.1.3 Determine and enter the decay-corrected activitics of the isotopes listed in Section 1.0 of Data Sheet 3 (-3) in accordance with.0PCP08-AP-0005' (Determination of Radionuclides \\- Post-Accident)..

\\.

Enter the decay-corrected activities obtained in Section 1.0 of Data Sheet 3 (-3).

6.2.1.4 Enter Counting Instrument used/ID # in Section 1.0 of Data Sheet 3 (-3).

U s

6.2.1.5 Sign and date Section 1.0 of Data Ebeet 3 (-3).

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i

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Coolant Activity and Radionuclide Trend for OPEP02-ZG-0007 Failed Fuel Rev. 1 Page 9 of 55 g

6.2.1.6 Forward a copy of Data Sheet 1 (-1) to the' Lead Chemical Technician for review.

6.2.2 Complete Section 2.0 of Data Sheet 3 (-3).

Addendum 6 provides appropriate data retrieval locations.

\\

i 6.3 Transcribe the decay-corrected specific activities from Section l'

l.0 of Data Sheet 3 (-3) to Section'3.0 of Data Sheet 3 (-3).

6.4 Using Addendum 1 and available plant power history data, determine the Power Correction Factor, X, for each of the' isotope listed on s

Data Sheet 3 (-3).

Record these values in Section 3.0 of Data Sheet 3 (-3).

Using Addendum 2 and the data recorded in Step 6.2.2, de'termine 6.5 the Pressure-Temperature Correction Factor, Z,' for the Containment Atmosphere isotopes listed on Data Sheet 3 (-3).

Record these values on Data Sheet 3 (-3), Section 3.0.

6.6 Using the appropriate equations presented in Data lheet 3 (-3),

Section 3.0, calculate the Adjusted Specific Activity for each isotope listed on Data Sheet 3 (-3).

Record these values on Data Sheet 3 (-3).

Using the information on Addendum 3 and the appropriate d' at.a 6.7 recorded on Data Sheet 3 (-3), complete Data Sheet 4.

6.8 If QDPS indications have shown core uncovery (Reactor Yessel Level

< 0) or if core uncovery is suspected to have occurred, then complete Data Sheet 5 using the data recorded on Data Sheet 4 and the guidelines in Addendum 4.

If no core uncovery is suspected, proceed to step 6.9 since Data Sheet 5 is used to estimate percent of fuel melt and fuel overtemperature.

6.9 Using the data recorded on Data Sheet 4 and the guidelines in Addendum 5 complete Data Sheet 6 which will provide percent fuel overtemperature and percent clad damage estimates.

6.10 Ensure that the applicable Data Sheets are signed, dated, and forwarded to the Reactor Performance Supervisor.

7.0 Acceotance Criteria None

________________________________r__________________

4 j[

Coolant ' Activity and Radionuclide Trend for OPEP02-ZG-0007 Emiled Fuel Rev. 1 Page 10 of 55 8.0 Dgrc.1muentation NOTE If this procedure was performed under normal operating conditions, only the documentation in Steps 8.1 and 8.2;is required. Otherwise, only the documentation in Steps 8.3 through 8.7 is required.

8.1 OPEP02-ZA-0007-1 Normal Operations Radionuclide Trend Data Sheet.

8.2 OPEP02-ZG-0007-2 Normal Operations Failed Fuel Estimate Based on I-131 Activity.

8.3 OPEP02 "G-0007-3 Post-Accident Specific Activity Determination.

8.4 OPEPU2-ZG-0007-4 Post-Accident Gross Activity Determination.

8.5 OPEP02-ZG-0007-5 Percent Fuel Overtemperature and Percent Fuel Melt Estimate.

8.6 OPEP02-ZG-0007-6 Percent Clad Damage and Percent Fuel Overtemperature Estimates.

t 8.7 QPEP02-ZG-0007-7 Evaluation of Auxiliary Indicators 9.0 References j

9.1 Westinghouse Owners Group "Postaccident Core Damage Assessment Methodology", Rev. 2 November 1984.

9.2 Bechtel Calculation No. 2N109MC5410," NPSH For Containment Spray Pumps During Recirculation", 2/23/83.

9.3 Bechtel Calculation, No. 2N129MC5037," RWST Verification & Level Setpoints", 9/4/82.

9.4 -

STPEGS FSAR Section 6.2.2.2.3, Amendment 38.

t-9.5 STPEGS FSAR Table 6.3-1, Amendment 55.

9.6 STPEGS FS AR Section 6.3.2.2, Amsndment 49.

9.7 STPEGS P&ID 5H129F05013, Rev.

6, " Safety Injection System".

1

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Coolant Activity and Radionuclide Trend for OPEP02-ZG-0007 Failed Fuel Rev. 1 Page 11 of 55 9.8 STPEGS P&ID SN129F05014, Rev. 6, " Safety Injection System".

9.9 STPEGS PEID 5N129F05015, Rev. 6, " Safety Injection System".

9.10 STPEGS P&ID 5N129F05016, Rev. 5, " Safety Injection System".

9.11 STPEGS P&ID SR149F05001, Rev. 6, "RCS Primary Coolant Loop".

9.12 STPEGS P&ID SR349705003, Rev. 4, "RCS Pressurizer".

9.13 STPEGS P&ID 52329Z00045, Rev. 3, " Primary Sampling System".

9.14 OPCP09-ZR-0004 (Determination of Radionuclides by Gamma Spectroscopy), Rev. 0 9.15 OPCP08-AP-0003 (Sampling and Analyais of RCS, RHR, and RCB Sump -

Post Accident), Rev. 0 9.16 OECP08-AP-0004 (Sampling and Analysis of the RCB Atmosphere - Post Accident), Rev. 0 OPCP08-AP-0005 (Determination of Radionuclides - Post-Accident),

9.17 Rev. 0 10.0 SuDDort Documents 10.1 Addendum 1 - Determination of Power Correction Factor, X.

Addendum 2 - Determination of Pressure - Temperature Correction 10.2 Factor, Z.

10.3 Addendum 3 - Determination of Sump Water Volume.

Addendum 4 - Estimation of Percent Fuel Overtemperature and 10.4 Percent Fuel Melt.

Addendum 5 - Estimation of Percent Clad Damage and Percent Fuel 10.5 Overtemperature.

Addendum 6 - Post-Accident Data Retrieval Locations.

10.6 10.7 Addendum 7 - Auxiliary Indicators of Core Damage Figure 1 - Power Correction Factor for I-131 Based on Average 10.8 Power During Operation.

Figure 2 - Relationship of Percent Clad Damage With Percent Core 10.9 Inventory Released of I-131.

J

o Coolant Activity and Radionuclide Trend for OPEP02-ZG-0007 Failed Fuel Rev. 1 Page 12 of 55 10.10 Figure 3 - Relationship of Percent Clad Damage With Percent Core Inventory Released of I-131 With Spixing.

10.11 Figure 4 - Relationship of Percent Clad Damage With Percent Core Inventory Released of I-132.

10.12 Figure 5 - Relationship of Percent Clad Damage With Percent Core Inventory Released of I-133.

10.13 Figure 6 - Relationship of Percent Clad Damage With Percent Core Inventory Released of I-135.

10.14 Figure 7 - Relationship of Percent Clad Damage With Percent Core Inventory Released of Kr-87.

10.15 Figure 8 - Relationship of Percent Clad Damage With Percent Core Inventory Released of Xe-131m.

10.16 Figure 9 - Relationship of Percent Clad Damage With Percent Core Inventory Released of Xe-133.

10.17 Figure 10 - Relationship of Percent Fuel Overtemperature With Percent Core Inventory Released of Xe, Kr, I, or Cs.

10.18 Figure 11 - Relationship of Percent Fuel overtemperature With Percent Core Inventory Released of Ba-140.

10.19 Figure 12 - Percent Zirc-Water Reaction Based on Hydrogen Concentration 10.20 Figure 13 - Percent Noble Gas Released 10.21 Data Sheet 1 - Normal Operations Radionuclide Trend Data Sheet.

10.22 Data Sheet 2 - Normal Operations Failed Fuel Estimate Based on I-131 Activity.

10.23 Data Sheet 3 - Post-Accident Specific Activity Determination.

10.24 Data Sheet 4 - Post-Accident Gross Activity Determination.

10.25 Data Sheet 5 - Percent Fuel Overtemperature and Percent Fuel Melt Estimate.

10.26 Data Sheet 6 - Percent Clad Damage and Percent Fuel Overtemperature Estimates.

10.27 Data Sheet 7 - Evaluation of Auxiliary Indicators

Coolant Activity and Radionuclide Trend for OPEP02-ZG-0007 Failed Fuel Rev. 1 Page 13 of 55 ADDENDUM 1 DETEEMINATION OF POWER CORRECTION FACTOR. X (Page 1 of 3)

The Power Correction Factor, X, may be determined by one of the following methods, whichever is more appropriate.

Isotopic decay constants may be found at the end of this addendum.

Method 1 Case It For isotopes Rb-88, I-132, I-133, I-135 If reactor power has not varied by more than + 10% from the time-average value for at least 4 days prior to sampling, use the following equationt X = % Steady-State (time-average) power for prior 4 days 100 Case II: For isotopes I-131, Xe-131m, Xe-133, Ba-140 If reactor power has not varied by more than + 10% from the time-average value for at least 30 days prior to sampling, use the following equationt X = % Steady-State (time-average) Dower for prior 30 days 100 t

l

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8 B

Coolant Activity and Radionuclide Trend for OPEPO2-ZG-0007 Failed Fuel Rev. 1 Page 14 of 55 ADDENDUM 1 DETERMINATION OF POWER CORRECTION FACTOR. X (Page 2 of 3)

Method 2 For all isotopes except Cs-134 Cs-137 For cases with transient power history in which the power has varied greater than i 10% of the time-average value prior to shutdown, use the following formula:

X=

P (1 - e t )e ~At 100

where, P = Percent power during operating period t t = Operating period (in days) at power P where power does not vary greaterthanf10%fromthetime-averdgevalue(P).

t = Time between end of period j and time of sampling (or shutdown, as appropriate) in days.

A = Decay constant of isotope i, in inverse days. See chart below for values of l

Isotope l Decay Constant A l

l l

l (Inverse days, d' )l I

I l

l I-131 1

0.0862 l

l I-132 l

7.2652 l

l I-133 1

0.7998 l

l I-135 l

2.5263 l

l Xe-131m l

0.0581 l

l Xe-133 1

0.1320 l

l Cs-134 9.2097 x 10" l

l l

l Cs-137 l

6.2944 x 10 l

l 1

l l

Ba-140 l

O.0542 l

l Rb-88 l

0.0123 l

Coolant Activity and Radionuclide Trend for OPEP02-ZG-0007 Failed Fuel Rev. l' Page 15 of 55 ADDENDUM 1 DETERMINATION OF POWER CORRECTION FACTOR. X (Page 3 of 3)

Method 3 For Cs-134 only:

Determine X for Cs-134 using the information on Figure 1 -

For Cs-137 only:

X = Total current Cumulative core Burnup (EFPD)

Total Design Expected Cumulative Core Burnup (EFPD)

For Present Cycle Example:

'The core has three regions of fuel with burnups as follows:

Region 1 (Thrice burned) - 400 EFPD current cumulative Region 2 (Twice burned) - 250 EFPD current cumulative Region 3 (Once burned) - 100 EFPD current cumulative The design expected burnup for each region is as follows:

Region 1 - 450 EFPD Regi_on 2 - 300 EFPD Region 3 - 150 EFPD Therefore i

400 + 250 + 100

= 0.83 450 + 300 + 150 i

4 4

r 1-.--

Coolant Activity and Radionuclide Trend for OPEP02-ZG-0007 Failed Fuel Rev. 1 Page 16 of 55 ADDENDUM 2 DETERMINATION OF PRESSURE - TEMPERATURE CORRECTION FACTOR. Z (Page 1 of 1) 4 The Pressure-Temperature-Correction Factor, Z, may be determined using the following equation.

P (T + 460) 2 P (T2 + 460) where, = P = Containment Atmosphere Sample Pressure, psig T = Containment Atmosphere Sample Temperature, F P = Containment t.tmosphere Pressure, psig 2

T = Containment Atmosphere Temperature, F 2

I i

Coolant Activity and Radionuclide Trend for CPEP02-ZG-0007 Failed Fuel Rev. 1 Page 17 of 55 ADDENDUM 3 DETERMINATION OF SUMP WATER VOLUME (Page 1 of 1) 1.0 Estimation of Containment Sump Water Mass The mass of water in the containment sump may be approximated by totaling the water mass released by the Accumulators and the RWST.

Use the equations below to determine these values.

% Water level prior

% water level 7

to accident at time of sample

.(4.1165 x 10 ) gm (M

ACC i 100

where, M

= mass (in grams) of water released by Accumulator i.

'%waterlevelprior) to accident are values for Accumulator

% water level at i recorded cn Data Sheet 3 (-3) i time of sample j

1-

% Water level prior

% water level to accident 9

RWST '

j - at time of sample i

.(1.2482 x 10 ) gm 100

where, M

=

mass (in grams) of water released by RWST RWST

!h

~

% water level prior to accident 3

p are values for RWST

% water level at recorded on Data Sheet 3 (-3) time of sample

Coolant Activity and Eadionuclide Trend for OPEP02-ZG-0007 Failed Fuel Rev. 1 Page 18 of 55 ADDENDUM 4 ESTIMATION OF PERCENT FUEL OVERTEMPEEATUEE AND PERCENT FUEL MELT (Page 1 of 4)

An estimation of Percent Fuel Overtemperature and Percent Fuel Melt may be made by comparing the relative activity ratios of Cs-137 and Ba-140.

Using the guidelines below will develop a plot of Percent Fuel Overtemperature vs. Percent Fuel Melt for the two isotopes. The intersection of the plots for each isotope will provide a best estimate percentage of each type of damage with the area underneath the curves representing all other possible combinations of damage (based on available data).

1.0 Calculation of Bounding Values Cesium 137 Total Gross Activity (Ci)

D

=

0%

137 7

1.2 x 10 (Ci)

where, D

= % of total core inventory of Cs-137 released 37

[ Total Gross Activity] = value for Cs-137 recorded on Data Sheet 4 (-4) i l

I l

1 I

i I

Coolant Activity and Radionuclide Trend for OPEP02-ZG-0007 Failed Fuel Rev. 1 Page 19 of 55 ADDENDUM 4 ESTJMATION OF PERCENT FUEL OVERTEMPERATURE AND PERCENT FUEL MELT (Page 2 of 4) 137 FOT 137 "

52.0 137 FM 137 "

89.0

where, FOT

= % Fuel which would have experienced overtemperature 137 conditions assuming all Cs-137 release was due to fuel overtemperature FM33, = % Fuel which would have experienced fuel melt assuming all Cs-137 release was due to fuel melt Barium 140 Total Gross Activity

~

(Ci)

~

D

=

x 100 2.0 x 10

where, D

= % of total cose inventory of Ba-140 released

[

40 (Total Gross Activity] = value for Ba-140 recorded on Data Sheet 4 FOT140 "

0.15 l

=

Coolant Activity and Radionuclide Trend for OPEP02-ZG-0007 Failed Fuel Rev. 1 Pag 9 20 of 55 ADDENDUM 4 STIMATION'OF PERCENT FUEL OVERTRNDERATURE AND PERCENT FUEL MELT (Page 3 of 4) 140 x 100 140 24.0

where, FOT

= % Fuel which would have experienced fuel overtemperature 140 assuming all Ba-140 release was due to Fuel overtemperature FM

= % Fuel which would have experienced fuel melt assuming 140 all Ba-140 release was due to fuel melt 2.0 Plotting of Data 2.1 Cesium Data a.

Plot, on Data Sheet 5, the FOT value determined in Section1.0abovealongthe"%huelOvertemperature" axis.

1.0abovealongthe"%FuelMef["valuedeterminedinSection Plot, on Data Sheet 5, the FM b.

axis.

i i

c.

Draw a straight line between both points plotted above and label the line "Cs-137".

2.2 Barium Data a.

Plot, on Data Sheet 5, the FOT value determined in Section1.0abovealongthe"%hhelOvertemperature" axis.

1.0abovealongthe"%FueiMeAk"valuedeterminedinSection Plot, on Data Sheet 5, the FM b.

axis.

c.

Draw a straight line between both points plotted above and label the line "Ba-140".

l 3.0 Data Interpretation 3.1 Draw a straight horizontal line fror the intersection of the "Ba-140" and "Cs-137" curves on Data Sheet 5 to the "% Fuel Overtemperature" axis. The intersection of this line and the axis yields the best estimate % fuel overtemperature value.

I

Coolant Activity and Radionuclide Trend for OPEP02-ZG-0007 Failed Fuel Rev. 1 Page 21 of 55 ADDENDUM 4 ESTIMATION OF PERCENT FUEL OVERTEMPERATURE AND PERCENT FUEL MELT (Page 4 of 4) 3.3 The area underneath the "Ba-140" and "Cs-137" curves represents all the possible combinations of fuel melt and overtemperature based on available data.

3.2 Draw a straight vertical line from the intersection of the "Ba-140" and "Cs-137" curves on Data Sheet 5 to the "% Fuel Melt" axis. The intersection of this line and the axis yields the best estimate % fuel melt value.

J t

Coolant Activity and Radionuclide Trend for OPEP02-ZG-0007 Failed Fuel Rev. 1 Page 22 of 55 ADDENDUM 5 ESTIMATION OF PERCENT CLAD DAMAGE AND PERCENl FUEL OVERTEMPERATURE (Page 1 of 2)

The following methodology is used to estimate Percent Clad Damage and Percent Fuel Overtemperature.

1.0 Core Inventory Releases (Total Gross Activity, Ci] i g7 x 100 (Core Inventory)

% Core Inventory Released for isotope i where, CIR =

(Total Gross Activity, Ci]

= value recorded on Data Sheet 4 (Core Inventory]

= Total available inventory (Ci) of isotope i.

Refer to table below.

l Isotope l

[ Core Inventory. Ci]

l l

l I-131 1

1.1f8) l l

I-132 l

1.7(8) l l

I-133 l

2.3(8) l I-135 l

2.1(8) l l

Kv-87 l

4.7(7) l L Xe-131m l

7.4f5) l Xe-133 l

2.3(8) l

_l Rb-88 l

6.7(7) l l Cs-134 l

2.7(7) l l

Cs-137 l

1.2(7) l l

Ba-140 l

2.0f8) l l

t s

6

Coolent Activity and Radionuclide Trend for OPEP02-ZG-0007 Failed Fuel Rev. 1 Page 23 of 55 ADDENDUM 5 ESTIMATION OF PERCENT CLAD DAMAGE AND PERCENT FUEL OVERTEMPERATURE (Page 2 of 2) 2.0 Comparison of the CIR values calculated in the previous section to the information presented in Figures 2 through 12 will result in % fuel failure estimates based on the activity of each isotope. The chart below related the appropriate Figures to be used for each isotope.

Isotope l % Clad Damage l

% Fuel Overtemperature l

Estimate l

l l

Finure No.

l Finure No.

l l

I-131 l

2*

l 10 l

l I-132 l

4 l

10 l

l I-133 l

5 l

10 l

l I-135 l

6 l

10 l

l Kr-87 l

7 l

10 l

l Xe-131m l

8 l

10 l

l Cs-134 l

10 l

l Cs-137 l

l 10 l

l Bu-140 l

l 11 l

If iodine spiking phenomena are suspected to have occurred, use Figure 3.

l l

)

I Coolant Activity and Radionuclide Trend for OPEP02-ZG-0007 Failed Fuel Rev. 1 Page 24 of 55 ADDENDUM 6 POST-ACCIDENT DATA RETRIEVAL LOCATIONS (Page 1 of 1)

Parameter l

Location (s)/ Description Containment Atmosphere l

Pressure l QDPS, CP018 (PR-0934)

Temperature l ODPS.

CPOO2 (TI-9681)

Containment Atmosphere l

Samples l

Pressure l Obtain from Chemical Analysis personnel Temperature l Obtain from Chemical AnalVEis Dersonnel RCS:

l Tavg l QDPS, CP005 (TI-0412A/0422A/0432A/0442A)

Pressure l ODPS.

CP004 (TI-0445/0456/0457/0458)

RWST Level l QDPS.

CP001 (LI-0931/0932)

Accumulator Levels l

A l ERF/ DADS, CP001 (LI-0950/0951)

B l ERF/ DADS, CP001 (LI-0952/0953)

C l ReF/ DADS.

CP001 (LI-0954/0955)

Containment Emergency l

Water Level (Wide emnae) lODPS. CP018 (LE-3925)

6 Coolant Activity and Radionuclide Trend for OPEP02-ZG-0007 Failed Fuel Rev. 1 Page 25 of 55 ADDENDUM 7 AUXILIARY INDICATORS OF CORE DAMAGE (Page 1 of 2) 1.0 Core Exit Thermocouples (CETC) and Reactor Vessel Water Level (RVWL).

CETC and RVWL indications can be used for verification of core damage estimates in the following ways a.

Due to the heat transfer mechanisms between the fuel rods, steam, and thermocouples, the highest clad temperature will be higher than the CETC readings. 1300 F is considered the lower limit for cladding failures. Therefore if CETC's read higher than 1300 F, clad failure may have occurred.

b.

If any RCP's are running, the CETC's will provide good indication of clad temperature and no core damage is expected since the forced flow of the steam / water mixture should adquately cool the core.

c.

The following guidelines apply if no RCP's are running:

1) No generalized core damage can occur if the core has not uncovered. Therefore if RVWL indications are such that the water 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 dainage has occurred.

2) If RVWL indicates that the water level has decreased below the top of the core and CETC readings indicate superheated. steam temperatures, then core damage may have occurred depending on the time after reactor trip, length, and depth of uncovery.

e Coolant Activity and Radionuclide Trend for OPEP02-ZG-0007 Failed Fuel Rev. 1 Page 26 of 55 ADDENDUM 7 AUXILIARY INDICATORS OF CORE DAMAGE (Page 2 of 2) 2.0 Containment Hydrogen Concentration Hydrogen is generated during accident conditions by the zirconium-water reaction between the clad and steam. The percentage of zirconium-water reaction can be estimated based ont he hydrogen concentration in containment using Figure 12.

Although the percentage of zirconium-water reaction does not equal the percentage of clad damaged, it does provide a qualitative verification of the extent of clad damage estimated from the radionuclide analysis.

3.0 containment Radiation Containment radiation levels can be used as an indication of core damage by calculating the exposure rate using the equation below and correlating the exposure rate and time after accident to the percent of noble gases released using Figure 13.

In general, values below 0.3% releases are indicative of clad failure, values between 0.3% and 52% release are in the fuel overtemperature region, and values between 52% and 100% are in the core melt regime.

However, the methodology for using containment radiation readings to predict core damage is based on several assumptions which may not be applicable to a given scenario. Thus the estimation of core damage using containment radiation readings has inherent uncertainties which limit its usefulness to confirmation of other indicators of core damage.

Exposure Rate (R/hr-MW ) = 5x10" (MW ~ ) x RMR (R/hr)

Where RMR is the radiation monitor reading from the wide range containment atmosphere monitors (A1RA-RE-8050 and ClRA-RE-8051).

Coolant Activity and Radionuclide Trend for OPEP02-ZG-0007 Failed Fuel Rev. 1 Page 27 of 55 FIGURE 1 POWER CODDRCTION FACTOR FOR CS-134 BASED ON AVERAGE POWER DURING OPERATION (Page 1 of 1)

I 1.0

.905PdWER 09

~7 0.8 POW ORRECTION 0.1

_W 155 POWER

/

0.6 0.5 EM PNFR 0.4

.g O.3 1

0.2 0.1 0.0 0

200 400 600 800 1000 i

t j

I i

I

Coolant Activity and Radionuclide Trend for OPEP02-ZG-0007 Failed Fuel Rev. 1 Page 28 of 55 FIGURE _Z RELATIONSHIP OF PERCENT CLAD DAMAGE WITH PERCENT CORE INVENTORY PFLFASED OF I-131 (Page 1 of 1)

I g,

0.7 55

'~

p 0.5

/..

/

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02

/

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7 0-5 5.0-5 3.0-5 2 0-5 i.0-5 I

a r; a; ad d

5 5d d

~

a n n ~

o o

o o o Clad Damage (%)

l

\\

Coolant Activity and Radionuclide Trend for OPEP02-ZG-0007 Failed Fuel Rev. 1 Page 29 of 55 FIGURE 3 RELATIONSHIP OF PERCENT CLAD DAMAGE WITH PERCENT CORE INVENTORY PRLFASrn OF I-131 WITH SPIKING (Page 1 of 1) 1 3,

0.7 s..

05 s..

/

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~~

3.0-5 2 0-5 1 0-5 9

4 A aai d

d di d tv n

n ~ o a

o o

o o Clad Damace (%)

Coolant Activity and Radionuclide Trend for OPEP02-ZG-0007 Failed Fuel Rev. 1 Page 30 of 55 FIGURE 4 m ATIONSHIP OF PERCENT CLAD DAMAGE WITH PERCENT CORE INVENTORY RELEASED OF I-132 (Page 1 of 1) g,3

.07 4:

/.

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4 A A4&

5 a5 &

~

n n a ~

a a

a a CladDamage(%)

Coolant Activity and Radionuclide Trend for Failed Fuel OPEP02-ZG-0007 Rev. 1 Page 31 of 55 FIGURE 5_

RELATIONSHIP OF PERCENT CLAD DAMAGE WITH PERCENT CORE INVENTORY pptras:ED OF I-133 (Page1of1}

1.

0.7 05 03 02 s

/

01

0I

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==

01

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n n m n a o

o o o o o o o o a

n n %

o

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t

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- P 5

0

Coolant Activity and Radionuclide Trend for OPEP02-ZG-0007 Failed Fuel Rev. 1 Page 32 of 55 FIGURE 6 nur.ATIONSHIP OF PERCENT CLAD DAMAGE WITH PERCENT CORE INVENTORY DFTWASED OF I-135 (Page 1 of 1) 1

.5 0.7

'E 0.5 03 02

/

01

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n.

m. u.

n n n u o o o o o o N

n n

A o

.o o o o o CladDamage(%)

, - =,, - - - -

. ~ - -. - -.,.

Coolant Activity and Radionuclide Trend for OPEP02-ZG-0007 Failed Fuel Rev. 1 Page 33 of 55 FIGURE 7 DFf.ATIONSHIP OF PERCENT CLAD DAMAGE VITH PERCENT CORE INVENTORY DFf_FARFn OF Kr-87 (Page 1 of 1) 01

.07 8

.05

.03

.02

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Q

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3 0-5 2 0-5 l.0-5 n.

n.

M.

N M

n A O

O O

O O O

o.

a a

a a a n n m ~

Clad Damage (%)

4

__._.,..__m

,..-,,.--_,.y-,

..m

Coolant Activity and Radionuclide Trend for OPEP02-ZG-0007 Failed Fuel Rev. 1 Page 34 of 55 FIGURE 8 DFf-ATIONSHIP OF PERCENT CLAD DAMAGE WITH PERCENT CORE INVENTORY prt.FASED OF Xe-131m (Page 1 of 1)

I 1

o.7 0.5

/

0.3 f.

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d di a

~

a n n - o o

o o e o Clad Damage (%)

-r-e

-m-

,n,-

an-,-,.

--r

-A

Coolant Activity and Radionutu de Trend for 01EP02-ZG-0007 Failed Fuel Rev. 1 Page 35 of 55 FIGURE 9 DFf ATIONSHIP OF PERCENT CLAD DANAGE WITH PERCENT CORE INVENTORY DFTFASED OF Xe-133 (Page 1 of 1) 1

[t 0.7 t

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n. m.

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n n m n o o

o o o o o

o o

o o n n m n o Clad Damage (".)

,f g

l Coolant Activity and Radionuclide Trend for OPEP02-ZG-0007 Failed Fuel Rev. 1 Page 36 of 55

(

FIGURE 10 DFfATIONSHIP OF PERCENT FUEL OVERTEMPERATURE WITH PERCENT CORE INVENTORY PFr.FASED OF Xe. Kr.

I. or Cs (Page 1 of 1)

(

f r

100-i k

702.

/

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)

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n n

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~

n n

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o i

l Fuel Overtemperature (%)

i I

t

._ _,\\

Coolant Activity and Radionuclide Trend for OPEP02-ZG-0007 Failed Fuel Rev. 1 Page 37 of 55 FIGURE 11 DFr-ATIONSHIP OF PERCENT FUEL OVERTEMPERATURE WITH PERCENT CORE INVENTORY DFrRASED OF Ba-140 (Page 1 of 1)

I i.

'O. 7:

0.5:

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2 0-4.

1.0-4 A

A A

E b

k Fuel Overtemperature (%)

s

.e Cuolant Activity and Radionuclide Trend for OPEP02-ZG-0007 Failed Fuel Rev. 1 Page 38 of 55

~

FIGURE 12

% ZIRC-kATER REACTION BASED

~

pN HYOROGEN CONCENTRATION (Page 1 of 1)3 26.0 24.0 22.0 20.0 r.

18.0 16.0 HYDR 0 Gell C0flCENTRATION (v/o) 14.0 12.0 10.0 8.0 6.0

~

4.0 2.0 10 20 30 40

.50 60 70 80 90 100

ZIRCONIUM WATER REACTION (".)

)

Coolant Activity and Radionuclide Trend for OPEP02-ZG-0007 Failed Fuel Rev. 1 Page 39 of 55 FIGURE 13 PERCENT NOBLE GAS RELEASED (Page 1 of 1)

I 3

1000.0 !

?

4 100% Noble Gas Release 100.0; 52% Noble 10.0 Gas Release H

C

?

E.

E 2

1.0:

=

I w

E 1.0-1 0.3% Noble Gas i

W Release 5?

?-

25 1.0-2p 3

1.0-3; Nomal Operating u

~

Noble Gas Release

'1.0-4!

(i s

~

~~

'1.0-5 1.0 10.0 100.0 1000.0 TIME AFTER ACCIDENT (HOURS)

Coolant Activity and Radionuclide Trend for OPEP02-ZG-0007 Failed Fuel Rev. 1 Page 40 of 55 DATA SHEET 1 NORMAL OPERATIONS RADIONUCLIDE TREND OPEP02-ZG-0007-1 (Page 1 of 2) 1.0 Isotopic Activities l

I I

I Sample I.D.

I Sample type I

I I

I Date of l

Month l

Date I

Year l Time ihrs) l Initials I

I I

I l Collection l l

l l

l l Analysis l

I I

I I Radionuclides l

Decav Corrected Activities (uCi/nm) l I

I I

I I-131 l

l l

I-?32 l

I I

I-133 l

l l

l I-135 l

l l

l Xe-131m l

l l

l Ke-133 l

l l

l Rb-88 l

l ETactor power level at time of sample:

Counting Inst. used/ID#2

' PERFORMED BY:

DATE/ TIME:

CHEMICAL TECHNICIAN REVIEWED BY:

DATE/ TIME:

LEAD CHEMICAL TECilNICIAN This form, when completed, shall be retained for 5 (five) years.

Coolant Activity and Radionuclide Trend for OPEP02-ZG-0007 Failed Fuel Rev. 1 Page 41 of 55 DATA SHEET 1 NORMAL OPERATIONS RADIONUCLIDE TDRED OPEP02-LG-0007-1 (Page 2 of 2) 2.0 Procedure Data Sample Time /Date Time of Sample Analysis Reactor Power at Sample Time (CP005, NI-0041B/0042B/0043B/0044B)

RCS Pressure at Sample Time psig (CP004, PI-0455/0456/0457/0458)

RCS Tavg at Sample Time F

(CP005, TI-0412A/0422A/0432A/0442A) 3.0 Calculations l Isotope l Specific l Power Correction l Adjusted Specific l Base-line l

l l Activity l Factor, X l

Activity l Specific Activity l l

If uCi/rul I

( uCi/em) l

( uCi/nm) l l I-131 l l

l l

l l I-132 l l

l l

l l I-133 l l

l l

l l I-135 l l

l l

l lXe-131ml l

l l

l lXe-133 l l

l l

l l Rb-88 l l

l l

l

[ Adjusted Specific Activity] = [ Specific Activity). X I-131/I-133 Ratio = II-131 Adiusted Specific Activitvl

[I-133 Adjusted Specific Activity]

I-131/I-133 Ratio =

Completed by Test Coordinator Time /Date Vzrified by Time /Date

Coolant Activity and Radionuclide Trend for OPEP02-ZG-0007 Failed Fuel Rev. 1 Page 42 of 55 DATA SHEET 2 NODMAL OPERATIONS FAltan FUEL ESTIKATE RAMED ON I-131 ACTIVITY OPEP02-ZG-0007-2 (Page 1 of 1)

Using the Adjusted Specific Activity value for I-131 recorded on Data Sheet 1 (-1), the following equations can be used to estimate the extent of failed fuel damage:

1.0 Number of Failed Fuel Pins (Maximum Expected and Best Estimate)

Adjusted Specific Activity ( uCi/ga) for I-131

=

3.5 x

10~

uCi/gm pins

=

2.0 Number of Failed Fuel Pins (Minimum Expected)

Adjusted Specific Activity ( uCi/ga) for I-131

=

4.9 x

10~

uCi/gm pins

=

3.0 Percent Failed Fuel (Maximum Expected and Best Estimate)

Adjusted Specific Activity ( uCi/gm) for I-131 1.8 uCi/gm

% Failed Fuel

=

4.0 Percent Failed Fuel (Minimum Expected)

Adjusted Specific Activity ( uCi/gm) for I-131 2.5 uCi/gm

% Failed Fuel

=

Completed by Test Coordinator Time /Date Verified by Time /Date This form, when completed, shall be retained for 5 (five) years.

4 Coolant Activity and Radionuclide Trend for-OPEP02-ZG-0007 Failed Fuel Rev. 1 Page 43 of 55 DATA SHEET 3 POST-ACCIDENT SPECIFIC ACTIVITY DETERMINATION OPEP02-ZG-0007-3 (Page 1 of 5) 1.0 Isotopic Activities RCS Samnle I

i l

l Sample I.D.

I Sample type I

I I

l Date of l

Month l

Date l Year l Time (hrs) l Initials l

I I

I l Collection i I

I I

Analysis l

I I

Radionuclides l

Decay Corrected Activities fuci/nm) l l

l l

l I-131 l

l l

l I-132 l

l l

l I-133 l

l l

l I-135 l

l 1

I I

l Cs-134 l

l l

l Cs-137 l

l l

l Ba-140 l

l l

l Eb-88 l

l Counting Inst, used/ID#:

This form, when completed, shall be retained for the life of the plant.

s Coolant Activity and Radionuclide Trend for OPEP02-ZG-0007 Failed Fuel Rev. 1 Page 44 of 55 DATA SHEET 3 POST-ACCIDENT SPECIFIC ACTIVITY DETERMINATION OPEP02-ZG-0007-3 (Page 2 of 5)

Containment Sump Sample l

I I

l Sample I.D.:

l Sample type:

l l

1 1

I I

I l Date of I

Month I

Date l Year l Time (hrs) l Initials l l

l l

l Collection l l

l l

1 1

I I

I l Analysis l

l l

l I

I l

l I

I I

I l Radionuclides l Decav Corrected Activities fuci/nm) l l

l.

l l

I-131 l

l l

l i

l I-132 l

l l

l I-133 l

l l

l 1

l I-135 l

l l

l Cs-134 l

l l

l 1

l Cs-137 l

l l

1 l

l Ba-140 l

l l

Rb-88 I

l l

Counting Inst. used/ID#:

i I

i l

l l

Coolant Activity and Radionuclide Trend for OPEP02-ZG-0007 Failed Fuel Rev. 1 Page 45 of 55 DATA SHEET 3 POST-ACCIDENT SPECIFIC ACTIVITY DETERMINATION OPEP02-ZG-0007-3 (Page 3 of 5)

Containment Atmosphere Sample l

I I

l Sample I.D.;

l Sample type:

l

+

l 1

1 I

I I

l Date of I

Month l

Date I Year l Time (hrs) l Initials l

l l

l l Collection l l

l l

1 I

l l Analysis l

l l

I I

I l Radionuclides l

Decav Corrected Activities (uci/cc) l l

I i

l Kr-87 l

l l

Xe-131m l

l l

Xe-133 l

l Countirg Inst. used/_ID#:

PERFORMED BY:

DATE/ TIME:

CHEMICAL TECHNICIAN REVIEWED BY:

DATE/ TIME:

LEAD CHEMICAL TECHNICIAN l

o Coolant Activity and Radionuclide Trend for OPEP02-ZG-0007 Failed Fuel Rev. 1 Page 46 of 55 DATA SHEET 3 EpST-ACCIDENT SPECIFIC ACTIVITY DETERMINATION OPEP02-ZG-0007-3 (Page 4 of 5) 2.0 Sampling Data Approximate Time /Date of Accident

/

RCS Conditions at Time of Sh.ple:

Tavg F

Pressure psig Containment Atmosphere Conditions at Time of Sample Pressure psig Temperature F

RWST and Accumulators at Time of Sample:

l l

RWST l Accumulator l

l l

l A

l B l C l

l % Water Level prior to l

l l

l accident l

l l

I l

l sample l

l l

l % Water Level prior to l

l l

Containment Emergency Water Level (Wide Range) at Time of Sample:

l feet i

i I

l l

l

Coolant Activity and Radionuclide Trend for OPEP02-ZG-0007 Failed Fuel Rev. 1 Page 47 of 55 DATA SHEET 3 POST-ACCIDENT SPECIFIC ACTIVITY DETERMINATLQX OPEP02-ZG-0007-3 (Page 5 of 5) 3.0 Calculations RCS Samnle l Isotope l Specific l Power-Correction l Adjusted Specific l l

l Activity l Factor, X l

Activity l

l If uCi/eml I

( uCi/em) l l I-131 l l

l l

l I-132 l l

l l

l I-133 l l

l l

l I-135 l l

l l

l Rb-88 l l

l l

lCs-134 l l

l l

lgs-137 l l

l l

lBa-140 l l

l l

Adjusted Specific Activity = Specific Activity X

( uCi/gn)

Containmen't Sump Sample l Isotope l Specific l Power Correction l Adjusted Specific l l

l Activity l Factor, X l

Activity l

l( uC1/nml l

( uCi/rm) l I-131 l l

l l

l I-132 i l

l l

l I-133 l l

l l

l I-135 l l

l l

l Rb-88 l l

l l

lCs-134 l l

l l

lCs-137 l l

l l

lBa-140 l l

l l

Adjusted Specific Activity = Specific Activity X

( uCi/gm)

Containment Atmosphere Sample l Isotope l Specific l Power Correction l Press-Temp. l Adjusted Specific l l

l Activity l Factor, X l Correction l Activity (uCi/cc) l l

l (uCi/cc l

l Factor. Z l

l l Kr-87 l

l l

l Xe-131M l l

l l

l l Xe-133 l

l l

Adjusted Specific Activity = Specific Activity X.

Y

( uCi/gm)

Completed by Test Coordinator Time /Date Verified by Time /Date

., _ =

Coolant Activity and Radionuclide Trend for OPEP02-ZG-0007 failed Fuel Rev. 1 Page 48 of 55 DATA SHEET 4 POST-ACCIDENT GROSS ACTIVITY DETERMINATION OPEP02-ZG-0007-4 (Page 1 of 3) t 1.0 Containment Sump Water Mass (M

I" ACC A E"

ACC B

[M3cc]c

=

gm N"

RWST

[ Containment Sump Water Mass) = [M

] + (MACC B ACC C RWST

[ Containment Sump Water Mass] =

gm 2.0 Gross Activity (Use the equations given below the table to complete the table) l l

RCS l

Containment Sump I Containment Atmosphere l

l l Adjusted l Gross l Adjusted l Gross l

Adjusted l

Gross l

l Isotope l Specific l Activity l Specific l Activity l Specific l Activity l l

l ( uCi/emil (C1) l

( uCi/em) I (C1)

'l Activity l

l l

l Activity l l Activity l

( uCi/cc)

(C1) l I-1311 l

l l

N l

I-1321 l

l l

I-1331 l

l l

I-1351 l

l l

l Rb-881 l

l l

,l l

l l

l Cs-1341 l

l l

l l Cs-137l l

l l

l l Ba-1401 l

l l

L l Kr-87l j\\

l l

l lXe-131ml l

l l

l l Xe-131 l

l l

This form, when completed, shall be retained for the life of the plant.

i

o Coolant Activity and Radionuclide Trend for OPEP02-ZG-0007 Failed Fuel Rev. 1 Page 49 of 55 DATA SHEET 4 EOST-ACCIDENT GROSS ACTIVITY DETERMINATION OPEP02-ZG-0007-1 (Page 2 of 3)

ECES Adjusted Specific (Gross Activity)

Activity ( uCi/ge) x 259.61 Ci

=

gp, p

r Containment Sumpt kdjustsSpecific Containment Sump Activity ( uCi/gn) x Water Mass (gn) isotope (Gross Activity],

p,=

Ci Containment Atmospheret 5

djusted Specific x

(1.008 x 10 ]

Ci (Gross Activity) isotope

=

Activity ( uCi/ce)i,,,,,

3.0 Total Gross Activity Use the table and corresponding equation below to determine the Total Gross Activity for each isotope.

l Isotope I

Total Gross Activity. Ci l'

l I-131 l

l l

I-132 l

l I

l I-133 l

l l

I-135 l

l l

Rb-88 l

l l

Cs-134 l

l l

Cs-137 l

l l

Ba-140 l

l l

Kr-87 l

l l

Xe-131m l l

l Xe-133 l

l

Coolant Activity and Radionuclide Trend for OPEP02-ZG-0007 Failed Fuel Rev. 1 Page 50 of 55 DATA SHEET 4 POST-ACCIDENT GROSS ACTIVITY DETERMINATION OPEP02-ZG-0007-4 (Page 3 of 3)

RCS Gross Containment Sump

[ Total Gross Activity (Ci)] isotope Activity

+

Gross Activity

=

isotope isotope

~

q

+

Containment Atmosphere Gross Activity

~

Completed by Test Coordinator Time /Date Verifiec by Time /Date l

l i

l

Coolant Activity and Radionuclide Trend for OPEP02-ZG-0001 Failed Fuel Rev. 1 Page 51 of 55 DATA SHEET 5 PERCENT FUEL OVERTEMPERATURE AND PERCENT FUEL MELT ESTIMATE OPEP02-ZG-0007-5 (Page 1 of 2) 1.0 Bounding Values

% of Cs-137 core inventory released.

D

=

37

% Fuel which would have experienced FOT

=

137 overtemperature conditions assuming al Cs-137 release was due to fuel overtemperature.

FM

=

ue w wu a e experienced melt 137 conditions all Cs-137 was due to fuel melt.

D

% of Ba-140 core inventory released.

40 FOT

% Fuel which would have experienced 40 overtemperature conditions assuming all Ba-140 release was due to fuel overtemperature.

% Fuel which would have experienced melt FM

=

140 conditions assuming all Ba-140 release was due to fuel melt.

This form, when completed, shall be retained for the life of the plant.

e.

p.

-..w.

7-r.

~

Coolant Activity and Radionuclide Trend for OPEP02-ZG-0007 Failed Fuel Rev. 1 Page 52 of 55 DATA SHEET 5 PERCENT FUEL OVERTEMPERATURE AND PERCENT FUEL MELT ESTIMATE OPEP02-ZG-0007-5 (Page 2 of 2) 2.0 Plotting of Data ll l l l l l l l l l l l l l l l l l l l 90l l l l l l l l l l l l l l l l l l l l l ll l l l l l l l l l l l lIIIIIll 80l l l l l l l l l l 11IIIIIlIll ll l l l l l l 11IIIIIIIII1l l l 1IliIIIIIIIIIIIIIl 70ll ll l l l l l l l l l 1IlIIIIIl l l l l l l l l l 1I IIIIl l l l 60l l;

; ; ; ; ; ; ; ;;;;;;;;,g

% Fuel l llIlIIll IIIIll IIlIl 50l l overtemperature l

l l l l l l l l l 1 I I I I I I I l_ l l l l 11IIIIIIIIIIIIIIl 40ll ll l l l l l l l l l l l l l IIIl l l l l IIIIIIIIilill IIl 30ll ll l 1IIIIIIIIIIIIIl l l l l l l l l l 11IIIIIIIIIIl 20l ;l

; ; ; ; ; ;;;;;;i,,;;;;

10l11IIIIIIIIIIIIIIIIIl ll l l l l l l l l l 1IIIIIIIIl ll l l l l l l l l l 1IIIIIIIIl 10 20 30 40 50 60 70 80 90 100

% Fuel Melt 3.0 Data Interpretation From intersection of Cs-137 and Ba-140 curves:

f Best Estimate % Fuel Overtemperature =

Best Estimate % Fuel Melt =

Completed by Test Coordinator Time /Date Verified by Time /Date

.o 6

Coolant Activity and Radionuclide Trend for OPEP02-ZG-0007 Failed Fuel Rev. 1 Page 53 of 55 DATA SHEET 6 PERCENT CLAD DAMAGE AND PERCENT FUEL OVERTEMPERATURE ESTIMATES OPEP02-ZG-0007-6 (Page 1 of 1) l Isotope l

CIR I

% Clad Failure l

% Fuel Overtemperature l

l l

(%)

i Estimate l

Estimate l

l l

1 l

l l

l I-131 l

l l

l I-132 l

l l

l I-133 l

l l

l I-135 l

l l

1 I

l Kr-87 l

l l

l Xe-131m l I

l l

Xe-133 l

l l

l Cs-134 l

l l

l Cs-137 l

l l

l Ba-140 l

I I

l i

Completed by Test Coordinator Time /Date Verified by Time /Date I

This form, when completed, shall be retained for the life of the plant.

L

J 4

s <, e s..

Coolant Activity and Radionuclide Trend for OPEP02-ZG-0007 Failed Fuel Rev. 1 Page 54 of 55 DATA SHEET 7 EVALUATION OF AUKILIARY INDICATORS OPEP02-ZG-0007-7 (Page 1 of 2) 1.0 Based CETC and RVWL indications, it apprears that:

(check one) l((l fuel damage may have occurred.

l l fuel damage has probably not occurred.

Summary of CETC and RVWL indications:

2.0 Based on Containment Hydrogen Concentration, it appears that:

(check one) l l fuel damage may have occurred.

l l fuel damage has probably not occurred.

Summary of Hydrogen Concentration indicatons:

_ =.

l This form, when completed, shall be retained for the life of plant.

l l

N

(.

)

s a.

+

~

Coolant Activity and Radionuclide Trend for OPEP02-ZG-0007 Failed Fuel Rev. 1 Page 55 of 55 DATA SHEET 7 EVALUATION OF AUXILIARY INDICATORS OPEP02-ZG-0007-7 (Page 2 of 2) 2.0 Based on Containment Radiation levels, it appears that:

(chect one) l[l fuel damage may have occurred.

l l

fuel damage has probably not occurred.

Summary of containment radiation indications:

Completed by:

Date Reviewed by:

Date

,

ML20211Q256

Text

1.0 Purpose and Scope

1.0 Purpose and Scope

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