0CAN028513, Forwards Methodology for Estimating Core Damage Based Upon post-accident Sampling Sys Readings,Per NRC 840827 Request Re TMI Item II.B.3.Estimate Will Not Be Used as Input to Emergency Planning or Other Procedures

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Forwards Methodology for Estimating Core Damage Based Upon post-accident Sampling Sys Readings,Per NRC 840827 Request Re TMI Item II.B.3.Estimate Will Not Be Used as Input to Emergency Planning or Other Procedures
ML20099C135
Person / Time
Site: Arkansas Nuclear  Entergy icon.png
Issue date: 02/28/1985
From: Enos J
ARKANSAS POWER & LIGHT CO.
To: John Miller, Stolz J
Office of Nuclear Reactor Regulation
References
RTR-NUREG-0737, TASK-2.B.3, TASK-TM 0CAN028513, NUDOCS 8503110233
Download: ML20099C135 (26)
v • d • e


Text

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ARKANSAS POWER & LIGHT COMPANY POST OFFICE BOX 551 LITTLE ROCK. ARKANSAS 72203 (501)371-4000 February 28, 1985 0CAN028513 Director of Nuclear Reactor Regulation ATTN: Mr. J. F. Stolz, Chief Operating Reactors Branch #4 Division of Licensing U. S. Nuclear Regulatory Commission Washington, DC 20555 Director of Nuclear Reactor Regulation ATTN: Mr. James R. Miller, Chief Operating Reactors Branch #3 Division of Licensing U. S. Nuclear Regulatory Commission Washington, DC 20555

SUBJECT:

Arkansas Nuclear One - Units 1 & 2 Docket Nos. 50-313 and 50-368 License Nos. DPR-51 and NPF-6 NUREC-0737 Item II.B.3, Methodology for Estimating Core Damage Gentlemen:

In our letter of January 24, 1984 (0CAN018413) AP&L committed to develop a methodology to estimate core dari' age based upon post accident sampling system (PASS) readings. This estimate vrill be used for informational purposes only and will not be used as input to emergency planning or other procedures. In your letter of August 27, 1984 (0CNA088425) you requested that AP&L submit this methodology by February 28, 1985 for your review. Therefore, our methodology for estimating core damage is attached.

Very truly yours, J. Ted Enos Managar, Licensing JTE:MCS:ds Attachment IL

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MEMBEA MtOOLE SOUTH UTILITIES SYSTEM

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A. PURPOSE This method is to be followed in order to estimate the type and degree of reactor core damage at Arkansas Nuclear One for plant conditions in which fuel failure is suspected.

B. REFERENCES

1) U.S. Nuclear Regulatory Commission " Post-Accident Sampling Guide for Preparation of a Procedure to Estimate Core Damage"
2) U.S. Nuclear Regulatory Commission - NUREG-0737, " Clarification of TMI Action Plan Requirements", Item II.B.3, " Post-Accident Sampling Capability", November 1980
3) U.S. Nuclear Regulatory Commission - NUREG-0772, " Technical Bases for Estimating Fission Product Behavior During LWR Accidents",

June 1981

4) U.S. Nuclear Regulatory Commission - NUREG/CR-1237, "Best-Estimate LOCA Radiation Signature", January 1980
5) AP&L Letter 9CAN058498 to J. Stolz, " Interim Guidelines for Assessing Core Damage at AN0", May 1984
6) Fort Calhoun Station Unit-1 Operating Instructions - OI-PAP-6,

" Procedure to Estimate Core Damage"

7) Babcock and Wilcox Company " Operator training - Degraded Core

, Recognition and Mitigaticn", June 1981 C. DEFINITIONS Core damage or fuel damaae is defined as the progressive failure of the two material boundaries which normally prevent the release of radioactive fission products into the RCS - the fuel pellet matrix and the fuel rod cladding. This method provides for the differentiation between the types and degrees of core damage shown in Table C.1.

Table C.1. Type and Degree of Core Damage Type Degree  % Failure No Damage NA NA Cladding Failure Initial <10%

Intermediate 10% - 50%

Major >50%

t Fuel Pellet Overheat Intermediate <50%

j Major >50%

Fuel Pellet Melt Initial <10%

Intermediate 10% - 50%

l Major >50%

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Source inventory is defined as the total quantity of each fission product isotope present in either source -- the fuel pellet or the fuel rod gas gap. Equilibrium source inventories for both ANO-1 and ANO-2 were calculated using the ORIGEN-2 computer code assuming 100% full power operation and a burnup of 26,000 MWD /MTV. For the case in which steady-state, full power operation has not been adequately maintained, corrections to the equilibrium source inventory will be applied.

Cladding failure is defined as the rupture of the fuel rod zircaloy cladding when the internal rod gas pressure exceeds RCS pressure and 4

the cladding yield strength is reduced due to elevated temperatures.

Cladding failure temperatures depend upon several factors; rate of temperature rise, internal gas pressure and cladding physical and mechanical properties. For the purpose of this method, cladding failure temperatures range from 1100 to 2000 F.

Fuel pellet overheat is defined as the condition in which fission products trapped within the fuel pellet are released at an accelerated rate due to increasing temperature. The release mechanism is driven by the formation, swelling, and coalescence of bubbles of fission gases.

At the temperature involved (>1600 F), the bubbles probably include vaporized cesium and iodine species as well as the noble gases. The expanded bubbles work to mechanically separate the U02 grains allowing the escape of gases and vapor.3 Fuel overheating temperatures typically range from 1600 to 3600 F cladding temperature6 .

Fuel pellet melt is defined as the condition where the fuel has become molten - releasing a large fraction of the remaining core noble gases, iodines and cesiums and detectable amounts of Te-132, Ba-140, and

, Ru-103. For the purpose of this method, the temperature range for fuel pellet melting is 3600 to 5400 F.

D. RESPONSIBILITIES

1) Nuclear Fuel or Nuclear Support Supervisor: The Nuclear Fuel or

< Nuclear Support Supervisor is responsible for initiating and I

directing the performance of this method during accident (or other) conditions which indicate the possibility of core damage.

2) Radiochemistry: Radiochemistry personnel are responsible for operating the Post Accident Sampling System to obtain the results

, necessary for performance of this method.

l l 3) Nuclear Fuel and Nuclear Support: Nuclear Fuel and Nuclear l Support are responsible for maintaining this method and for aiding in collecting data and evaluating the results of this method.

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E. PREREQUISITES

1) Pcst Accident Sampling System is operable - isotopic activities are available from the following locations (if applicable):

a) Reactor Coolant (RC) System b) Containment Atmosphere (CA) c) Containment Sump (CS)

2) Hydrogen concentration in the primary coolant sample and containment atmosphere is available.
3) Core exit thermocouple readings are available from plant computer.
4) Containment pressure and temperature are known.
5) Sample pressure and temperature are known.
6) Level in reactor building sump is known: Unit 1 - The containment building sump contains 26.6 gallons per % level indication. The sump level detection system indicates water levels in %. Unit 2 -

The containment building sump contains 39 gallons per % level indication. The sump level detection system indicates water levels in %.

7) Previous Power history is known.
8) Equilibriuli Fission Product (FP) inventories are known.

F. PRECAUTIONS AND LIMITATIONS General The assessment of core damage obtained using this method is only an estimate. The techniques employed in this method are accurate only to locate the core condition within one or more of the categories of core damage given in Table C.1.

1) The following precautions apply to the analysis of measured F.P.

specific activities:

For the purpose of this method, it is assumed that samples taken from the RCS hot leg are representative of a homogeneous system.

No provisions are made for the sampling of the pressurizer steam space where fission gases may collect disproportionately during normal operation and under accident conditions.

The method relies upon samples taken from multiple locations inside the Containment Building to determine the total quantity of fission products available for release to the environment. The amount of fission products present at each sample location may be changing rapidly due to transient plant conditions. Therefore, it is required that the samples be obtained within a minimum time period and if possible under stabilized plant conditions. Samples obtained during rapidly changing plant conditions should not be weighed heavily into the assessment of core damage.

A number of factors influence the reliability of the chemistry samples, such as the inability to obtain representative samples due to incomplete mixing of the fluids, and equipment limitations.

The accuracy achieved in the radiological analyses is also influenced by a number of factors. For example, samples may not result from a uniform distribution of the sample fluid and cooling or reactions me take place in the long sample lines. Hence, the results obtained may not be representative of plant conditions.

To minimize these effects, multiple samples should be obtained over an extended time period from each location.

2) The following precautions apply to the analysis of hydrogen concentrations:

Hydrogen samples are taken from the containment atmosphere and the Reactor Coolant System Hot Leg. Those samples may contain a mixture of hydrogen generated t>y clad oxidation, radiolytic dissociation of water and oxidation of metals in the containment.

Therefore, a hydrogen measurement is not a direct indicator of the amount of core clad oxidation.

Since measured hydrogen does not directly indicate the amount of core clad oxidation (core damage), the interpretation of hydrogen data employed in this method is based on the presence or absence of excess levels of hydrogen.

3) The following precautions apply to the analysis of Core Exit Thermocouple (CET) temperatures:

Severe core damage cannot be quantified by CET analysis alone.

This method is designed to provide a weighted estimate of core damage based on all available indicators.

The relationship between the CET temperature and the cladding temperature varies with the core uncovery scenario. Due to the lack of more detailed data, the correlation of clad temperature to CET temperature and system pressure developed by Babcock and Wilcox (Attachment A) is used in this estimate.

G. INITIAL CONDITIONS This method is to be employed for analysis of radiochemistry, hydrogen,

-and core exit thermocouple data when it is determined that a plant accident with the potential for core damage has occurred. The following is a list of plant symptoms to assist in this determination.

One or more of these symptoms may exist at or before the time the sample is obtained.

1) High alarm on the Containment Radiation Monitor
2) Reactor coolant sample indicates high activity level

I i

3) Reactor Coolant System sub-cooling low or zero .

I

4) Any other condition exists in which the operator suspects fuel '

failure l H.

SUMMARY

l The core damage estimates obtained using this method will be based on ,

the comparison of measured fission product concentrations from various '

sample locations to the known fission product inventory of the core.

The four types of core damage (no damage, cladding failure, fuel pellet overheat and fuel pellet melt) can be differentiated by the presence, absence or abundance of certain fission products and by other plant indicators (i.e., core exit thermocouples and containment or primary system hydrogen concentration).

METHOD I. Select appropriate isotopes for estimate II. Determine appropriate sample locations III. Obtain necessary data A. Plant data B. Radiochemistry data IV. Correct equilibrium Fission Product (F.P.) inventories to recent power history V. Correct measured sample activities to standard temperature and pressure (STP)

VI. Correct measured sample activities at STP for decay since shutdown VII. Calculate key isotopic ratios to determine source of release VIII. Calculate total activity of isotopes available for release IX. Calculate total activity of isotopes released X. Interpret results The following is a statement of the objectives and brief summary of each step in the method:

STEP I: To select the appropriate isotopes for use in the estimate.

The isotopes, their source of release and selection criteria are given in Table I.A. The fundamental selection criteria for the fourteen isotopes given in the table were; their detectability in early and later points in transients, their radiological significance in regard to off site doses, the sensitivity of detector instrumentation to them

and the experience in analyzing samples for them. Additionally, empirical data concerning the fractional release of these isotopes as well as procedures to estimate core damage utilizing them have been published.

STEP II: To select the appropriate sample locations for the estimate based on the accident scenario. The suggested sample locations for three major accident scenarios are given in Table II.A. These scenarios are adequate to define core damage within the types and degrees defined in Table C.1 and are representative of three fundamental scenarios in which core damage may occur.

STEP III: To record pertinent plant data as closely as possible to the time at which radiological samples are obtained from the PASS. The data include reactor shutdown time, previous power history, RC pressure and temperature, CET temperature, CA pressure and temperature, etc.,

which are necessary in accurately locating core damage as defined in Table C.1. A worksheet is provided in Step III of the method.

STEP IV: To correct the computer generated equilibrium source inventories for re. cent reactor power history. Equations for the calculation of corrected inventories and a worksheet for recording them are provided in Step IV of the method. There are two types of isotopes provided for the estimate of core damage; Group 1 isotopes, which require approximately 30 days of reactor operation at power levels 110%

of some level to reach equilibrium, and group 2 isotopes, which require approximately 4 days of operation as above. It should be noted that severe core uncovery scenarios cannot be quantified by Group 2 isotopes alone. In the event power has not been maintained as indicated above, a general power correction factor is provided.

STEP V: To obtain and correct the measured samples specific activities for standard temperature and pressure. Although the PASS automatically makes this correction, the calculational method is provided in Step V of the 7ethod for the case in which samples have not been corrected for STP. lble V.A is provided as a record of these specific activities.

STEP VI: To correct.the specific ac.tivities obtained from Step V for decay since shutdown occurred. The PASS will automatically make this correction, if Radiochemistry personnel are advised to program the reactor shutdown time. In the event samples have not been corrected for decay since shutdown, the calculational method is provided. An additional calculation is necescoj in order to account for the parent-daughter relationship at Iodine and Xenon.

STEP VII: To calculate isotopic ratios in order to determine whether the source of inventory release is the fuel pellet or the gas gap.

Table VII.A is provided to record the indicated release source based on the ratios provided in Step VII. Additional ratios for normal coolant can be calculated using data obtained from samples taken under normal operating conditions and additional ratios for releases can be deduced from the release fractions of Attachment B.

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STEP VIII: To calculate the total activities of the isotopes available for release from the fuel pellets and gas gaps. The activities available in the fuel rod gas gap are obtained by multiplying the nominal gap release fractions of Attachment B by the power corrected equilibrium source inventories from Step IV. The activities available in the fuel pellets are obtained by multiplying the nominal meltdown release fractions of Attachment B by the power corrected equilibrium source inventories from Step IV. A table is provided in Step VIII for available activities.

i STEP IX: To calculate t:1e activitier, wnich were released based on PASS samples. The released activities, in curies, are the sums of the activities released to each sample location. Equations for this calculation are provided in Step IX. ,

STEP X: To interpret the data and calculate an estimate of the type and degree of core damage. Guidelines for making the estimate of core damage are provided in Step X. Attachment C provides a flow-chart for the interpretation of da+a and the calculation of a core damage estimate.

I. APPROPRIATE ISOTOPES FOR CORE DAMAGE ESTIMATE Table I.A. Isotopes, Soure of Release and Selection Criteria Isotops Source of Release Selection Criteria Kr-85m Gas Gap -Indicators of all degrees of Kr-87 Gas Gap cladding failure.

Kr-88 Gas Gap -When seen in quantities Xe-131m Gas Gap greater than source inventory Xe-133 Gas Gap times the upper limit gap release I-131 Gas Gap fractions of Attachment B I-133 Gas Gap Indicators of overheating and the I-135 Gas Gap source of release is the fuel pellet.

Te-132 Fuel Pellet Indicators of intermediate and Ba-140 Fuel Pellet major fuel pellet overheating and Ru-103 Fuel Pellet all degrees of fuel pellet melting.

Cs-136 N/A Used for calculation of key ratios.

Cs-137 N/A Cs-138 N/A 6

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II. APPROPRIATE SAMPLE LOCATIONS FOR CORE DAMAGE ESTIMATE

/ Table II.A. Sample Locations Accident Scenario Conditions Sample Locations RC CS CA Small Break LOCA Sub-cooling margin not maintained X X* X*

Large Break LOCA N/A X X X Steam Line Break Sub-cooling margin not X in Containment maintained i

  • Required only if alarm on containment building dome monitor.

, _ _ _ _ _ . _ , - . _ - . , - - . - , _ . . _ . _ _ - , _ _ _ _ . . _ , _ . . . _ _ . . ~ _ . - _ _ . . . . . . _ . - . _ . - . _ . _ , . _ _ . . - - - _ . -

l III. RECORD OF PLANT DATA Record the following plant conditions as closely as possible to the time at which radiological samples are obtained from the PASS.

A. Time of reactor shutdown Date Time '.

B. Prior 30-day power history Power  % Duration Days

% Days

% Days

% D;ys

% Days C. Reactor Coolant System (RC):

1) Pressure psia
2) Temperature F
3) CET Temperatute F
1
4) Reactor Vessel Level  %

s,

5) Pressurizer Level  %

ANO-1 2.29x108cc 'ANO-2 2.10x10 scc

6) RC Volume (VRC)
7) Time of Measurement Date Time D. Containment Building:
1) Containment Atmosphere Pressure psia
2) Containment Atmosphere Temperature F
3) Radiation Dose Rate rads /hr s

e r

.g r

i

4) Containment Sump Level ANO-1  % ANO-2  %
5) Containment Sump Volume
  • ANO-1 cc ANO-2 cc
6) Containment Free Volume ANO-1 5.19x1010cc ANO-2 5.15x1010cc
7) Time of Measurement Date Time
  • Containment Sump Volume is calculated as follows:

5.6 gal ANO-1: (Level %) x x 3785.4 EE

% gal 39 a1 ANO-2: (Level %) x x3785.4fc]

3:x s

st t

f I lu I

IV. TABLE OF POWER HISTORY CORRECTED EQUILIBRIUM SOURCE INVENTORY Equilibrium Source Power Corrected Source Decay Power Inventory Inventory Constant History (Curies) (Curies)

Isotope (day 1) Grouping ANO-1 ANO-2 PCF ANO-1 ANO-2 Kr-85M 3.713 2 1.8x107 2.2x107 c Kr-87 1.280x101 2 3.5x107 4.2x107 Kr-88 5.941 2 4.9x107 5.9x107 X:-131M 5.776x10~2 ~

1 7.7x10s 8.3x105 Xe-133 1.315x10 1 -

1 1.4x108 1.5x108 I-131 8.611x10 2 ~

1 6.9x107 7.5X107 I-133 7.998x10 1 2 1.4x108 1.5x108 _

I-135 2.483 ~

2 1.3x108 1.4x108 {

2.160x10 1 9.8x107 Tc-132 -

1 1.1x108 Ba-140 5.415x10 2 ~

1 1.2x108 1.4x108 '

Ru-103 1.750x10 1 1.1x108 1.1x108 Cs-136 5.332x102 ~

1 2.9x108 2.7x108 Cs-137 6.548x10 5 N/A 6.7x108 6.2x108 Cs-138 3.100x101 2 1.3x108 1,4x108 _

Group 1 Power Correction Factor (PCF) = Steady-State Power Level in Prior 30 Days F 100 (__

  • Group 2 PCF = Steady-State Power Level in prior 4 Days 100
    • General PCF =

Where: Pj = steady reactor power in period j, %

t) = duration of period j, days tj=timefromendofperiod j to reactor shutdown, days Ai = decay constant of isotope i, day ~1

  • For use when steady-state power level (110%) for has not been sustained for 30 days, but has been sustained for at least 4 days. Only Group 2 isotopes should be used in estimate.
    • For use when steady-state power level has not been sustained between 110%

for 4 days. Entire 30-day power history should be used in applying the General Power Correction Factor.

V. CALCULATION OF STP CORRECTED SPECIFIC ACTIVITIES

  • Trble V.A. STP Corrected Specific Activities Measured Specific ,, STP Corrected Sp- 'fic Isotope Activities (pCi/cc) Correction Factor Activities (pCi,- ;

RC CS CA RC CS CA RC CS CA Kr-85m Kr-87 Kr-88

, Xe-131m Xe-133 I-131 I-133 I-135 Te-132 Ba-140-Ru-103 Cs-136 Cs-137 Cs-138

  • PASS automatically corrects sample to standard temperature and pressure -

Verify with Raoiochemistry personnel that samples were corrected for STP.

    • If applicable The following calculational method is provided for the case in which samples have not been corrected for STP:

RC samples are corrected for RCS temperature and pressure using the factor for water density in Table V.B. and the following equation:

SRC (corrected) = SRC (measured) X FD CS samples do not require correction for temperature and pressure within the accuracy of this method.

CA samples are corrected using the following equation.

P2 (T,+460)

CA (corrected) = SCA (measured) X P1(T2 +460) where: P1 ,Ti = containment atmosphere sample line pressure and temperature (*F).

P2 ,T2 = containment atmosphere pressure and temperature ( F).

S CA,RC = specific activity (pCi/cc) in CA, RC.

t Table V.B. RC Density Correction Factor, FD FD = RC density correction factor i

RCS Temperature at RCS Density Correction Time of Sample (*F) Factor, F g 100 1 150 0.987 200 0.970 250 0.949 a

300 0.924 350 0.897 i 400 0.865 450 0.830 500 0.790 550 0.741 560 0.731 580 0.709 600 0.683 620 0.653 640 0.620 660 0.582 680 0.531 700 0.441

.The values of F Dabove were calculated assuming a 100 F sample temperature.

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VI. CALCULATION OF DECAY CORRECTED SPECIFIC ACTIVITIES

  • Table VI.A. Decay Corrected Specific Activities Decay Power STP Corrected Corrected Decay History Specific Correction Specific Isotope ' Constant (day 1) Grouping Activity Factor Activities (pCi/cc)

RC CS CA -

RC CS CA Kr-85m 3.713 2 Kr-87 1.280x101 2 Kr-88 5.941 -

2 X -131m 5.776x10 2 1

X:-133 '1.315x10 1 1 1-131 8.611x102-1 1-133 7.998x10 1 2 I-135 2.483 2 Te-132 1 1

Ba-140 2.160x10[2 5.415x10 1 Ru-103 1.750x10 2 1 Cs-136 5.332x10 2 1 Cs-137 6.548x10 5 N/A Cs-138 3.100x101 2

  • PASS will automatically decay correct specific activities if Radiochemistry personnel are advised to program the shutdown time. The equation used is:

= b S

-Ait where: e S = the specific activity of the STP co rected sample corrected back to the timeofreactorshutdown,pCi/cc S = the STP corrected specific activity, pCi/cc Ai = the decay constant for isotope i, day ~1 t = the time period from reactor shutdown to sample analysis, days.

For Xe-131m and Xe-133, an additional calculation must be made in order to account for the increase in Xe inventory due to I decay since shutdown.

Axt Xe-131m(real)=Xe-131m(measured)e -0.008 (A I)I-131(measured) (eAxt.7)

Ax Axt Xe-133(real)=Xe-133(measured)e -0.976 AII-133(measured) (e #-1)

Ax where t = number of days between shutdown and sample analysis.

Ax=decayconstantofXenonisotope, days"It Ai = decay constant of Iodine isotope, days

VII. KEY ISOTOPIC RATIOS IN REACTOR COOLANT AS AN INDICATION OF FUEL DAMAGE Fuel Normal Transient Gap Matrix

  1. Key Ratios Coolant Spiking Release Release 1 I-131:1-133 1 +2 5 0.6 2 I-131:1-135 2 ~4 16 0.6 3 Cs-137:Cs-136 5 ~10 90 6 4 Cs-137:Cs-138 0.6 $4 110 0.04

. Table VII.A. Release Source Determination Ratio # Calculated Ratio Indicated Release Source 1

2 3

4 4

h 4

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-VIII. . CALCULATION OF ACTIVITIES AVAILABLE FOR RELEASE ,

TABLE VIII.A. Total Activities Available for Release *

-Isotope Pellet (Ci) Gap (Ci)

Kr-85m Kr-87 Kr-88 Xe-131m Xe-133 I-131 1-133 I-135 Te-132 Ba-140 Ru-103 Cs-136 Cs-137 Cs-138

  • Total activities available for release are obtained by multiplying the release fractions for the source or release tabulated on Attachment B by the power history corrected source inventories of Step IV.

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IX. CALCULATION OF TOTAL ACTIVITIES RELEASED (A) .

TABLE IX.A. Total Activities Released Sum of Power STP and Decay Corrected Total Corrected Activity History Specific Activities Activity Rel. eased Released Isotope Grouping (pCi/cc) (Ci) (C1)

RC CS CA RC -

CS CA Kr-35m 2 Kr-87 2 Kr-88 2 Xe-131m 1 Xe-133 1 1-131 1 1-133 2 I-135 2 Te-129 2 Te-132 1 Ba-140 1 Ru-103 1-Cs-136 1 Cs-137 N/A Cs-138 2-Total corrected activities released are calculated as follows:

1) RC - If the water level in the reactor vessel recorded in step III indicates that the vessel is full:

A(C1) = STP & Decay Corrected Specific Activity xVRCx10 8 If the water level in the reactor vessel and pressurizer recorded in step III indicate that a steam void is present, then the activity released is calculated as above. However, it must be recognized that the value obtained will-overestimate the actual quantity released. Therefore, this sample should be repeated at such time when the operators have removed the reactor vessel void.

If the water level in the reactor vessel recorded in Step III is below the low end capability of the indicator, it is not possible to determine the activity released from this sample because the volume of water in the reactor coolant system is unknown. In this case, assessment of core damage is obtained using the containment sump sample.

2) CS -

A(Ci) = STP & Decay Corrected Specific Activity X V cs 10 8

3) CA -

A(Ci) = STP & Decay Corrected Specific Activity X V ca 10 8 The total activity released is the sum of the released activity in each sample location.

.J

X.A. INTERPRETATION OF CORE DAMAGE The core damage estimate will start by determining which of the four fuel conditions (no damage, cladding failure, fuel overheat or fuel melt) best describes the core environment. A flow chart has been prepared to show the logic for estimating core damage (Attachment C).

A. No Damage The core conditions which pertain to no damage or normal operation are as follows:

1) The reactor was shutdown with no abnormal conditions and other plant indicators verify that there has been adequate core cooling such as:

- All of the core exit thermocouple (CET) readings indicate that core damage is unlikely per Attachment A.

- No excess amount of hydrogen is found in the primary system and no detectable hydrogen is found in the containment atmosphere.

2) If the above best describes the core environment then divide the total released activity A by the upper limit normal operating activities given on Attachment D.

Isotope Calculated Fraction Kr-85m Kr-87 Kr-88 Xe-131m Xe-133 I-131 I-133 I-135 Te-132 Ba-140 Ru-103 Cs-136 Cs-137 Cs-138

4) If A >l total upper limit activity for any fission products, then it is possible that some cladding failure has occurred. In this case recheck the CET readings to see if any one of them indicated possible cladding failures and determine the percent cladding failure.

B. Cladding Failure The core conditions which pertain to cladding failure are as follows:

1) The reactor was shutdown with adequate core cooling and other plant indicators verify that there has been a possibility for cladding failures, such as:

- one or more of the CET readings indicate the possibility of cladding failures per Attachment A.

- no excess amount of hydrogen is found in the primary system and no detectable hydrogen is found in the containment atmosphere.

A >l total upper limit activity

2) If the above best describes the core environment, then divide the total released activity, A, by the total available gap release activity.*

Isotope Calculated Fraction Kr-85m Kr-87 Kr-88 Xe-131m Xe-133 I-131 1-133 I-135 Te-132 Ba-140 Ru-103 Cs-136 Cs-137 Cs-138

  • Gap release activities are obtained from Step VIII.

Ther. A x 100 = percent cladding failure total gap release activity

3) If A x 100 >100 for any of the above total gap release activity fission products, then there is a possibility that some fuel overheating or fuel melting has occurred. Recheck the following plant indicators to determine if fuel overheating or fuel melting has occurred:

- one or more core exit thermocouple readings indicate the possibility of fuel pellet overheat per Attachment A followed by detection of excess hydrogen in the containment atmosphere or the primary system.

- the presence of any low-volatile fission products (i.e. Te, Ba, or Ru) indicates that some fuel melting may have occurred.

NOTE: Absence of low-volatiles (i.e. , Te, Ba, Ru) is a good indication that no fuel melting has occurred.

C. Fuel Overheating The core conditions which pertain to fuel overheating are as follows:

1) There has been an abnormal shutdown and there is a possibility that the fuel has been partially uncovered for a period of time greater than a few minutes. Other plant indicators verify that there has been fuel overheating such as:

- one or more core exit thermocouple readings indicate the possibility of fuel pellet overheat per Attachment A.

- excers hydrogen is found in the primary system or a detectable amount of hydrogen is found in the containment atmosphere.

A >l

, total available gap release activity I

i.

~

2) If the above best describes the core environment, then divide the total released activity, A, by the total available core inventory.

Isotope Calculated Fraction Kr-85m Kr-87 Kr-88 Xe-131m Xe-133 I-131 I-133 I-135 Te-132 Ba-140 i Ru-103 Cs-136 Cs-137 Cs-138

3) If A <0.2 for all of the noble gases total core inventory then it could be concluded that there has been <50% fuel overheating. This conclusion can be further verified by indication that some of the core exit thermocouple readings indicate the possibility of fuel pellet overheat per Attachment A followed by detection of excess hydrogen in primary system or containment. In case low volatiles (i.e.

Te-132, Ru-103 or Ba-140) were detected, determine the percent fuel melting.

4) If A >0.2 for any of the noble gases total core inventory then it could be concluded that there has been >50% fuel overheating. This conclusion can be furthe. verified by indication that >50% of ccre exit thermocouple readings indicate the possibility of fuel pellet overheat per Attachment A. In case low-volatiles (i.e. Te-132, Ru-103 or Ba-140) were detected, determine the percent fuel melting.

D. Fuel Melting The core conditions which pertain to fuel meltin0 are as follows:

1) There has been a severe accident and the core has been uncovered for a long period of time. The following plant indicators verify that there has been fuel melting.

- more than one core exit thermocouple reading indicates the possibility of fuel pellet overheat per Attachment A.

~

- one or more core exit thermocouple readings indicate the possibility of fuel pellet melt per Attachment A.

- excess amount of hydrogen is found in the primary system or a detectable amount of hydrogen is found in the containment atmosphere.

A <0.2 with presence of low-volatiles Leal core inventory (i.e. Te-132, Ru-103, or Ba-140)

A >0.2 with presence. of low-volatiles total core inventory A >l with presence of total gap release activity low volatiles

2) If these conditions best describe the core environment, then divide the total released activity, A, by the total available meltdown release activity.*

Isotope Calculated Fraction Kr-85m Kr-87 Kr-88 Xe-131m Xe-133 I-131 1-133 I-135 Te-132 Ba-140 Ru-103 Cs-136 Cs-137 Cs-138

  • Nominal meltdown release activities are obtained in Step VIII then A x 100 = percent fuel melted total available meltdown release activity NOTE: The fuel melting estimates will be primarily based on the data from low-volatiles (i.e. Te, Ba, or Ru) since the data from other fission products (i.e. Xe, Kr, I, Cs) will not provide for a good estimate of fuel melting in the lower sub group (i.e. lower than 10% fuel melting). This is due to the fact that Xe, Kr, I and Cs fission products are released in significant quantities during fuel overheating. Therefore, it would be

difficult to determine whether the releases were due to fuel overheating or fuel melting. However, estimates based on Xe, Kr and I can provide verification of the estimates which are based on Te, Ba, or Ru.

E. The following notes should be considered in interpretation of the core damage estimates:

1) If the cladding failure or fuel overheating estimates based on Xe and Kr are significantly different (i.e. Xe and Kr estimates are in different sub groups), recheck the estimates and if necessary obtain new samples to determine new concentrations.
2) If the core has been uncovered for a long time and the estimates based on noble gases and iodines indicate significantly higher percentage of fuel melting than the estimates based on Te, Ru, or Ba (i.e. the estimates are in a different sub group), the low-volatile estimates must be rechecked and if necessary new samples obtained to determine new concentrations. If the results remain the same as before, the fuel melting estimates should be based on the low-volatiles.
3) The fuel melting estimates based on Xe, Kr, or I will not provide for a good estimate in the <10% sub group. This is due to the fact that these fission products are released in significant quantities during fuel overheating and it would be difficult to determine whether the releases were due to fuel overheating or fuel melting.
4) The samples taken during the later stages of the accident may provide for a more accurate core damage estimate, than those taken in the early stages of the accident.
5) If the accident did not involve a breech of the primary l system, CS and CA samples will provide no useful information.

X.B. TABLE OF PERCENT RELEASED DATA Isotope Total Activity Released (A) Percent Release Kr-85m Kr-87 Kr-88 Xe-131m Xe-133 I-131 I-132 I-133 I-135 Te-132 Ba-140 Ru-103 Cs-136 Cs-137 Cs-138

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