ML17254A187

Rev 3 to Procedure PC-25.4, Guidelines for Interpreting Post-Accident Sampling Results to Estimate Reactor Core Damage

ML17254A187
Person / Time
Site:	Ginna
Issue date:	01/22/1985
From:	ROCHESTER GAS & ELECTRIC CORP.
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ML17254A186	List: ML17254A185 ML17254A187
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PC-25.4, NUDOCS 8502040207
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ROCHESTER GAS AND ELECTRIC CORPORATION GINNA STATION CONTROLLED COPY NUMBER 8INNA STATION UNIT ~1 COMPLETED DATE:-

TINIE:-

PROCEDURE NO.

\\

PC-25.4 REV.

NO ~

3 GUIDELINES FOR INTERPRETING POST-ACCIDENT SAMPLING RESULTS TO ESTIMATE REACTOR CORE DAMAGE TECHNICAL REVIEW PORC REVI.EW DATE

~ JAN

~

0 EVIEW PLANT SUP INTENDENT JAN 22 IS85 EFFECTIVE DATE QA~

NON-QA CATEGORY 1 ~ 0 REVIEWED BY:

TElS PROCEDURE CONTAlNS 21 PAGES 8502040207 8gpg@9 +;

PDR ADOCK OSOOOann I'

PDR

~~

~ ~

PC-25.4:1 PC-25 4

GUIDELINES FOR INTERPRETING POST-ACCIDENT SAMPLING RESULTS TO ESTIMATE REACTOR CORE DAMAGE PURPOSE:

This procedure provides guidelines for the preliminary assessment of reactor core damage based upon post-accident sampling results or other radiological indications.

1 ~ 2 1

~ 3 Certain. reactor transients could result-in fuel cladding

damage, fuel overheating or potential fuel melting.

One or more of these conditions would involve the release of radionuclides into. the primary coolant, followed by possible transfer.to auxiliary systems and the containment atmospheres By examining the magnitude of radioactivity increases in a post-accident condition, as well as confirming the presence or 'ab'sence of 'c'ertain groups of rad'ionuc'li'd'es; an early assessment of core damage can be performed.

2.8 2 ~ 1 2 ~ 2 2 '

REFERENCES:

Procedure PC-23. 1, PC-23. 2, PC-25. 2, PC-4, PC-5 Procedure P-9 Procedure S-14.2, S-14.3 2.4 2.5 Procedure SC-188 NUREG-8737, II.B.3 2.6 2 '

2 '

Westinghouse Mitigating Core Damage Training M'anual Rogovin Report, Part 2,

Volume II, pp 524 527 WASH-1488 Appendix VII 2 '

Westinghouse Owner's Group Post, Accident Core Damage Assessment Methodology, Rev.

2,

November, 1984

3.8 PRINCIPLE-3 '

Gross indicators of radioactivity concentration (letdown

monitor, containment, area and airborne monitors) will provide early information on the severity of core damage following a transient.

When later primary coolant and containment samples are collected and analyzed, measured concentrations of radionuclides can be examined to determine the extent of total available activity that has been released to the RCS Auxiliary systems.

Certain radionuclides will tend not to be released unless fuel overheating or melting occurs.

Thus the absence of these species will bound the maximum extent of core damage.

The determination of certain isotope ratios may also help distinguish whether released activity

~

originated in the gap or from overheated fuels 3

~ 2 Additional indications of core damage can be made using the data for core exit temperatures and containment radiation monitors.

Fuel damage from overtemperatures and possible fuel melt can be indicated.by increased hydrogen levels in containment atmosphere.

P 4.8 4.1 4.2 4.3 4 '

4.5 PREREOUISITES AND NEEDED EQUIPMENT:

Isotopic analysis of primary liquids.

Data from plant radiation monitors.

Plant operational status including pertinent core data.

Isotopic analysis of containment atmosphere.

Hydrogen analysis of containment atmosphere.

4.6 Data from core exit thermocouples.

5 '

PRECAUTIONS AND LIMIT VALUES:

5.1 Care should be taken to avoid def ining too precisely the extent of core damage based upon initial sampling results.

Other plant indicators will also be available (such as incore temperature indication, containment hydrogen

monitors, etc.)

which should also be considered in arriving at a more refined estimate.

5.2 The time of sampling relative to the suspected transient or core degradation sequence must be considered.

The effects of isotope

decay, sampling equilibrium and progressing core degradation may tend to confound sample interpretation.

F

~

~

PC-25.4:3 5.3 Reactor power history is to be considered in determining whether certain key radionuclides have reached equilibrium.

5.4 Measured concentration of radioactivity may need to be adjusted to account for system dilution (e.g

~ accumulators, safety injection water) prior to es'timating fuel damage.

6 ~ 1 INSTRUCTIONS:

Indication of RCS Abnormal Conditions 6 ~ 1 ~ 1 There are various instruments which will indicate to the plant operator if abnormal conditions exist in the RCS which could contribute to fuel failure.

This instrumentation would include indication of coolant temperature,

pressure, subcooling and water inventory (ifavailable).

Plant operating and

-emergency procedures provide detailed instruct,ion on how the instrumentation is to be used.

6 '

~ 2 Post-accident analysis of these parameters should determine whether v'oidi'ng. of the RCS occuired, or'ore specifically, whether any part of the core was uncovered at any time Special attention should be paid to core exit thermocouple temperatures and core water level indications.

If the core remained covered, fuel damage should be limited to gap releases from cladding defects.

However, if the core was uncovered, it becomes more probable that extensive cladding oxidation could have occurred leading to cladding and fuel pellet fragmentation.

6 '

6. 2.1 6 ~ 2

~ 1.1 Gross Radioactivity Indications Letdown Monitor R-9 If letdown has not been isolated, monitor R-9 will provide a gross indication of current primary coolant activity'efer to procedures P-9 and SC-188.

A reading of 2888 mR/hr on monitor R-9 will correspond to approximately 1% fuel rod cladding defects.

The monitor's upper

range, 18,888 mR/hr, may therefore be assumed to correspond to about 5%

cladding defects.

6'.1.2 Containment Airborne Monitors R-ll, R-12 6.2.1

~ 2 ~ 1 Primary coolant system leakage results in the release of noble gas and radioactive particulates into the containment atmosphere.

An increase in primary coolant activity and/or system leakage will be indicated on containment monitors R-11 and R-12.

Refer to procedures P-9 and S-14.2.

6'.2 Containment Area Monitors R-2, R-29 and R-38

PC-25.4:4 6.2

~ 2 ~ 1 An increase in the direct radiation readings of the containment area-monitors will also indicate abnormal primary coolant activity and/or leakage conditions in the containment building.

Refer to procedures P-9 and S-14.3

~

Procedure S-14.3 provides a series of curves showing monitor dose rate versus time after shutdown.

The curves have been calculated for various accident categories including postulated gap activity release, coolant

release, and fuel inventory release.

The procedure also provides tables of estimated isotopic activities which correspond to the accident categories considered.

6.2.3 Primary Coolant and Containment Atmosphere Samples

' '.3.1 The previous indications may show that a substantial release of radioactivity has occurred from an accident conditions Undiluted samples may read several orders of magnitude higher than under normal conditions.

Extreme precautions are required in sample collection, handling and analysis to minimize potentially severe radiological hazards that may exist.

6"..3 Release of Fuel; Rod Gap Activity'

~ 3

~ 1 Analytical results indicating only noble

gas, iodines and cesium isotopes will suggest that only a release of gap activity has occurred.

Radionuclides such as strontium, barium and tellerium would not be significant contributors to coolant activity in this situation because of much lower escape fractions.

6 '

For a given fraction of failed fuel (e.g.

18%),

the initial degassed activity will be mainly comprised of radioiodines (at T=8 hr).

The initial activity contribution due to cesium may be small compared to the iodine activity.

6.3 '

For perspective, assuming 18% failed fuel defects, the initial gas and degassed activities are calculated to be approximately 184 times higher than normal activity levels.

Activity levels will drop substantially (factor of 2 to 3) during the first day due to the decay of short-lived noble gas and iodine isotopes.

6 '

Evidence of Fuel Overheating 6 ~ 4 ~ 1 Potential fuel overheating may be suggested 'if the actual primary coolant activity cannot be accounted for by assuming 188$ release of gap activity; or, if certain radionuclides are detected which are indicators of fuel pellet overheating'

~ 4.2 Under conditions of fuel overheating, radionuclides such as strontium, barium and tellerium would be released from the fuel to a greater extent than found in gap releases.

PC-25. 4: 5 6.5 6.5

~ 1 Evidence of Fuel Melting Post-accident samples which display extremely high activity or a greater presence of normally less volatile radionuclides, may indicate that partial fuel melting has occurred in addition to clad damage.

6.5

~ 2 For a given fraction of melted fuel (e.g.

18%) the initial degassed activity will be largely due to radioiodine.

However, the presence of strontium activity is likely to be detected following radioiodine

decay, and is calculated to be more predominant than cesium activity under fuel melt conditions.

6.5

~ 3

6. 5.4 For perspective, assuming 18% melted fuel, the initial gross gas is calculated to increase approximately a factor of 18 over normal levels.

The gross degassed activity may increase as much as 18 to 186.

Detailed sample isotopic analysis may also indicate the presence of noble metals (Ru, Rh, Pd, Mo, Tc), rare earths and. act'inides (U,.Pu) and 'other'e'fractor'y materials which would also confirm extensive fuel damage.

6.6 Calculation of released activity.

6.6.1 Record the following plant indications.

The values should be recorded as close as possible to the time at which the samples are taken.

Reactor Coolant System:

Pressure Temperature Reactor Vessel level Pressurizer level PSIG OF Containment Building:

Atmosphere Pressure Atmosphere Temperature Sump level PSIG OF 6.6.2 Obtain and analyze selected samples using approved procedures.

Use reactor shutdown as time zero for all calculations'ecord decay corrected liquid sample data on ATTACHMENT 1.

Record decay corrected containment atmosphere sample on ATTACHMENT 2.

PC-25. 4: 6 6.6.2.1 Record data on ATTACHMENT 2.

Containment, atmosphere samples may need to be corrected for elevated temperatures and pressures using the equation:

Specific Act. = Measured Act x P2 x T~

+ 468 Pl

+ P2 T2

+ 468 where Tl, Pl = Measured sample temperature F and pressure in psigo T2, P2 = Standard temperature, 32oF and standard pressure 14.7 psia.

6.6.3 Record'on ATTACHMENT 3.

Calculate for each sample the following ratios for each noble gas and iodine isotope using the calculated activities.

I Noble Gas Ratio = Noble Gas Isoto e Act.

Xe-133 Specific Act.

Iodine Ratio

= Iodine Isoto e Act.

I-131 Specs.fic Act.

63 '

Determine the possible source of release, either fuel pellet or clad gap, by comparing the results obtained to the predicted ratios listed on Attachment 3.

An accurate comparison i.s not anticipated.

Within the accuracy of this procedure it is appropriate to select as the source that ratio which is closest to the value obtained above.

6.6.4 Calculate the sum of the total quantity of each fission product available for release from sampled sources.

6.6.4.1 Record on ATTACHMENT 4 as curies total.

It is assumed for this discussion that the RCS is at normal level and a

LOCA has not occurred.

If a LOCA has occurred, the determination of the mass in the RCS must be estimated by indications available to operations.

The major activity would probably be i:n the sump.

Calculate the quantity of fission products in the reactor coolant using:

Curies Total activity = Acto (uCi/cc) x RCS Mass x 1E-6 where:

RCS Mass calculated from data in step 6.6.1 and Ratio Correction Factor from Table 2.

RCS Mass

= (6236 C.F.)

x (2.83E4) x (Ratio correction factor) 6.6 '.2 Record on ATTACHMENT 4 as curies total.

Using activity in sump calculated on Attachment 1, calculate the total quantity of fission products in the sump using:

PC-25. 4: 7

6. 6.4. 2 (Cont'd)

Curies Total activity = Acto (uCi/cc) x Sump Volume x lE-6 Sump volume is determined from Table 3.

As an approximate temperature for the liquid in the sump, use the air temperature of containment.

6.6.4.3 Using activity in uCi/cc calculate on Attachment 2, calculate the quantity of fission products in the containment building atmosphere.

Correct.

the volume for pressure and temperature recorded in step 6

~ 6

~ 1 as required using the equation in 6.6.2.1.

Record on ATTACHMENT 4 as curies total.

Use approximate volume of containment at STP as 9.7E5 cubic feet or 2.7E18 cc.'.6.5 Power Correction Factor.

See Attachment 5

6.6.5.1 Steady state power prior to shutdown.

Steady

state, power

~

condition is assumed where the power does not vary by more than

+ 18 0'of rated power level from'time averaged vaiue.

6.6

~ 5 ~ 1.1 Half-life of nuclide

(

1 day Power Correction Factor = Avera e Power Level (Mwt) for rior 4 days Rated Power Level (Mwt) 6 '.5.1.2 Half-life of nuclide 1 day Power Correction Factor = Avera e Power Level (Mwt) for rior 38 da s

Rated Power Level,(Mwt) 6.6.5.1.3 Half-life of nuclide -

1 year Power Correction Factor = Avera e Power Level (Mwt) for rior 1 ear Rated Power Level (Mwt) 6.6.5.2 Transient power history in which the power has not remained constant prior to reactor shutdown.

For the majority of the selected

nuclides, the 38-day power history prior to shutdown is sufficient to calculate a power correction factor.

6.6.5.2.1 Determine the Fission, Product Core Inventory by correcting the Equilibrium Source Inventory using the power correction factor equation:

Pj (1-e 3) e

~

3 Power Correction Factor

=

RP

PI

PC-25. 4: 8 6.6.5 '.1 (Cont'd) where:

P>

= average power level (Mwt) during operating period t>

RP

= rate power level of the core (Mwt) t> = operating period in days at power P>

where power does not vary more than

+

18 percent power of rated power level" from time averaged value (P>)

= decay constant of nuclide i in inverse days I

t

> = time between end of period j and time of,reactor shutdown in days The power correction factor equation above applies to the category of nuclides with half-lives greater than one day.

.6.6.5.2.2 For the few nuclides with half-lives around one year or longer, a power correction factor which 'ratios cycle-burnup in effe'ctive full power days to total calendar days of cycle operation is applied.

Power Correction Factor = Actual o eratin EFPD of e uilibrium c cle Total expected EFPD of equilibrium cycle operation 6.6 '

Determine the percent of release by dividing the Total Quantity available for.Release by the Corrected Source Inventory for the Gas Gap or Fuel Pellet as indicated on Attachment 5 for source of release (either gap or fuel pellet).

6.7 Containment hydrogen concentration NOTE:

The use of this section assumes a

LOCA releasing H2 into containment formed by a zirconium-steam reaction in the core.

6 ~ 7 ~ 1 Obtain a

measurement of containment atmosphere hydrogen concentration.

6 '

Assuming that all hydrogen formed by a zirconium-water reaction is released to the containment atmosphere, either use Table 4

or calculate the fraction of zirconium-water reaction with the equation:

FZWR =

(8 H2)

(V) (correction factor for STP)

ZM H

1 8-H2%

PC-25.4:9 6.7.2 (Cont'd) where:

FZWR Fraction of Zirconium-Water Reaction V =

Containment

volume, SCF, approx.

9.7E5 CF ZM =

Total zirconium mass, approx.

23,988 lbs.

Conversion

factor, 7.92 SCF of H2 per pound of zirconium reacted the above equation becomes FZWR =

(% H2) (5.12)(Corr Factor for STP)

(188-H2%

6 ~ 7

~ 3 6 '

6 ~ 8 ~ 1 The maximum value that 8 H2 approaches is 13.8%

9 STP.

Core exit thermocouples The core exit thermocouples (CETC) will indicate the peak temperatures of the cladding as long as one RCP is running if there has been no uncovery of the reactor core.

If no CETC indicates temperature of superheated steam at RCS pressure, this is a good indication that the core has not been uncovered and no generalized damage has occurred.

6 ~ 8 ~ 1 ~ 1 If a CETC temperature exceeds 1388 F, cladding failure may have occurred.

6

~ 8 ~ 1 ~ 2 If a CETC temperature exceeds 1658 F, comparison to neighboring CETC is necessary as this may indicated the thermocouple has failed.

6.9 Containment Radiation Monitors 6 ~ 9.1 The CV Rad monitors will indicate increased levels of radiation if there is damage to the core. If there is no LOCA, increased radiation from normal leakage into containment, will not be representative of core damage.

6.9 '

Use table from procedure S-14.3 to estimate a possible amount of damage if a LOCA occurs.

6 ~ 18 The collection of all data with the tables of this procedure, will give an indication of'he potential core damage that may have occurred.

The data may be contradictory from different

sources, so engineering judgement of the data collected is necessary to reach a realistic estimate.

Within the Westinghouse Core Damage Assessment Methodology are graphs relating cladding damage to 8 core inventory released for specific isotopes (Figures 2-2 to 2-9 for gap release; figures 2-11 and 2-12 for fuel overtemperature; figures 2-13 to 2-15 for fuel melt).

GUIDELINE IO DATA NEEDEU FISI CCRE UAHAGt ASSESSHEHT tST IHATIUN PI laary Systoa Saa le Sua Saa le Contalnaent San AtuOSPhvru le Collect Saa le Col lect Saa I o Collect Saa le Sou pie lor H2 In bv Ue as Saa le H

Noble Gas Act Dilute Saa Ie Take dlluto saaple for Act Dilute Saa le Take diluted saaplu tor Act.

Correct to Ua I c ~

H2V )r reaction by vq ~ u. ) ~ a or true taulu o

Calculate total actlvlty available In HCS Gal culdto total dct Ivlty available In Rl'S Correct

)or Tuup, ~

d Vo I ~

Ca I c Tot a I act Ivlt In sua Core Inventor y Attachaent

.5 Sua ol total actlvlty keleased lroa Core

~

Correct lan lor Pover 0 oration Cole

> ol Available Core Inventor xuluased Core Exit Theraocou les Est last lon ol Core Oauage dnd lype ol Udadgu using Rddlolsotoplc Act IvltIes dnd svdl ldble ddt d points and Nest lnghousu Coro Oaaa u AsseSSaont Hethoduloa Uuntalnaunt Hav 1st lon

~ unltors IM Ul

~ ~

C)

TABLE 1 Characteristics of Cat pries of Fuel Dama e*

Page 1

NRC Category of Fuel Dama e Mechanism of Release Source of Release Characteristic Release of Characteristic Isotope Expressed as a

Percent of Source Inventor 1.

No Fuel Damage Halogen Spiking Gas Gap Tramp Uranium I 131, Cs 137 Rb 88 Less than 1

2 ~

3 ~

Intermediate Cladding Failure Major Cladding Failure Clad Burst and Gas Gap Diffusion Release Gas Gap Gas Gap Xe 131m, Xe 133 I 131, I 133 8 to 58 Greater than 58 4,

Intermediate Fuel Pellet Overheating Grain Boundary Diffusion Puel Pellet Cs 134, Rb 88, Te 129, Te 132 8 to58 5.

Major Fuel Pellet Diffusional Release Fuel Pellet Overheating From UO2 Grains Greater than 58 6.

Intermediate Fuel Escape from Pellet Melt Molten Fuel Fuel Pellet 8 to 58 7 ~

Ba 148, La 148 Fuel Pellet La 142, Pr 144 Major Fuel Greater than 58 Pellet Melt This table is intended to supplement this procedure and should not be used without referring to Westinghou Core Damage Assessment Methodology and without considerable engineering judgement.

ItV lJl

~t

TABLE 1 Characteristics of Cate pries of Fuel Dama e*

Page 2

Core Damage Indicator Core Damage Cate~os 1.

No Fuel Damage 2.

8-58% clad damage 3.

58-188% clad damage 4.

8-58'8 fuel pellet overtemperature Containment Radiogas Monitor (R/hr)

> 8 hrs after shutdown 28 8 to 258 258 to 588 588 to 78,888 Core Exit Thermocouples Readings (De<<D P)

< 758 758-1388 1388-1658

> 1658 Core Uncovery Indication No uncovery Core uncovery Core uncovery Core uncovery Hydrogen Monitor (Vol g P

) ***

Negligible 8 - 6 6-13 6-13 5.

58-188% fuel pellet overtemperature 78,888 to 158,888

> 1658 Core uncovery 6-13 6.

8-588 fuel melt 158,888 to 228,888

> 1658 Core uncovery 6-13 7.

58-1888 fuel melt

> 228,888

> 1658 Core uncovery 6-13 This table is intended to supplement this procedure and should not be used without referring to Westinghouse Core Damage Assessment Methodology and without considerable engineering judgement.

Values calculated from Westinghouse Methodology.

• • • Ignitors may obviate -these values.

IhJ EJI

~ ~

Table 2

pc-25.4:13 Ratio of H20 Density to H20 Density at STP vs. Terperature I

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500 ~

300 200 100

.SO

.75

1. 00 fAct/f STP

, oooooo

')

Table 3

PC-25.4:14 00000*

OOOO

)

8 000 Gal ons in Sunup, vs

~

Sump Level Indicatik Ihen Con aimanL Bas t

lll QOd CQ LS CC 9

00'0 0

5 10 15 20 25 30 35 40 Sump Level Reading

PC-25.4:15 Table 4

26. 0 24.0 22.0

20. 0

18. 0 HYDROGEiN CONCENTRATION (v/0')

16.0 14.0

12. 0 2-LOOP

10. 0 8.0 6.0 4.0 2.0

'10 20 30 40 50 60 70 80 90 100 ZIRCONIUM WATER REACTION

( o)

CONTAINMEiNT HYDROGEN CONCEiNTRATION BASED ON ZIRCONIUM WATER REACTION

PC-25.4:16 ATTACHMENT 1 RECORD OF SAMPLE SPECIFIC ACTIVITY IN LIQUID SAMPLES Sample Number:

Sample Location:

Time of Analysis:

Temperature, F:

Pressure, PSIG:

Sample Activity, uCi/gm at time Total Mass of liquid Corrected total activity released Kr 87 Xe 131m Xe 133 I

I 131 I 132 I 133 I 135 Cs 134 Rb 88 Te 129 Te 132 Sr 89 Ba 148 La 148 La 142 Pr 144 Cs 137 Use reactor shutdown as To

PC-25 ':17 ATTACHMENT 2 RECORD OF SAMPLE TEMPERATURE AND PRESSURE CORRECTION FOR CONTAINMENT BUILDING ATMOSPHERE SAMPLE SPECIFIC ACTIVITY Sample Number:

Sample Location:

Time of Analysis:

Temperature, oF:

Pressure, PSIG:

~Isot'o e

Kr 87 Xe 131m Measured Sample Activit uCi cc at STP

• 2.75E4 Ci/uCi/cc x Correction Factor for STP"'

Total Activity 9 STP, Ci Xe 133 131 I 132 I 133 I 135 Cs 134 Rb 88 Te 129 Te 132 Sr 89 Ba 148 La 148 La 142 Pr 144 Cs 137 Use reactor shutdown as To

• • Correction factor for STP = Pl

+ ] 4 ~ 7 Tl CV Temp oF Pl

= CV psig x

492 Tl +~

ATTACHMENT'3 RECORD OF FISSION PRODUCT RELEASE SOURCE IDENTIFICATION Sample Number:

Location:

I~ceto e

Kr 87 Xe 131m Xe 133 I 131 I 132 I 133 I 135 Decay Corrected**

Specific Activity uci/cc Calculated 1.8 Fuel Pellet 8,22

8. 884 1.8 1

~ 5 2 '

1'.9 Approximate Activity Ratio In Gas Ga

8. 822

8. 884 1.8

8. 17

8. 71

8. 39 Identified Source Noble Gas Ratio = Deca Corrected Noble Gas S ecific Activit Decay Corrected Xe 133 Specxfxc Actzvxty Iodine Ratio = Deca Corrected Iodine Isoto e

S ecific Activit Decay Corrected I-131 Specz.fic Activity

• • MCA data calculated with reactor shutdown as To

,IhJ Ql

~ ~

CQ

~

v ATZACMKZ4 KMMKTION RECORD OP RELE'ASE QUAHPITY PC

.4: 19 I~so to Kr 87 Reactor Coolant Sample Number Act. x RCS Mass Ci Containment Sump

+

Sample Number Act. x Sump Mass Ci Containment Atmosphere Sample Total Act. x C.V. Vol.

=

Quantity Number

, Ci Ci Xe 131m Xe 133 I 131 I 132 I 133 I 13S Cs, 134 '".

Rb 88 Te 129 Te 132 Sr 89 Ba 148 La 148 La 142 Pr 144 Cs 137

PC-25.4:28 ATTACHMENT 5 CORRECTED SOURCE INVENTORY I~seto e

Equilibrium Source Inventory, Ci Power Correction Fraction Corrected Source Kr 87 Xe 131m Xe 133 I 131 I 132 I 133 I 135 1 '

x 183 3.8 x 182 7.6 x 184 1

~ 2 x 185 2.8 x 184 8.3 x 184 4.2 x 184 Fuel Pellet Inventor Kr 87 "Xe 131m Xe 133 I 131 I 132 I 133 I 135 Cs 134 Rb 88 Te 129 Te 132 Sr 89 Ba 148 La 148 La 142 Pr 144 Cs 137

,1 ~ 7 2 '

8 ~ 5 4.2 6.1 8 '

7 '

1.8 2 '

1 ~ 4 6 '3.4 7 ~ 3 7 ~ 7 6.5 5.3 4.6 x 187 x 185 x 187 x 187 x 187 x 187 x 187 x 187 x 187 x 187 x 187 x 187 x 187 x 187 x 187 x 187 x 186 Additional information is available from Nestinghouse Core Damage Assessment Methodology tables 2-2; 2-3; 2-3-1.

e

~

I PC-25.4:21 ATTACHMENT 6 RECORD OF PERCENT RELEASE I~seto e

From ATTACHMENT 4 Total Quantity Available For Release Ci From ATTACHMENT 5 Corrected Source Percent Inventor Released Kr 87 Xe 131m Xe 133 I 131 I 132 I 133 I 135 Fuel Pellet Inventor Kr 87 Xe '131m Xe 133 I 131 I 132 I 133 I 135 Cs 134 Rb 88 Te 129 Te 132 Sr 89 Ba 148 La 148 La 142 Pr 144 Cs 137