Text

Rev 1 to Emergency Implementing Procedure FNP-0-EIP-30, Post-Accident Core Damage Assessment
	ML20101E606
Person / Time
Site:	Farley
Issue date:	12/14/1984
From:	Mcdonald R, Woodard J; ALABAMA POWER CO.
To:	Varga S; Office of Nuclear Reactor Regulation
References
	RTR-NUREG-0737, RTR-NUREG-737, TASK-2.B.3, TASK-TM FNP--EIP-30, FNP-0-EIP-30, NUDOCS 8412260316
	Download: ML20101E606 (46)
	v • d • e

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November 8, 1984 t

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Revision 1 FARLEY NUCLEAR PLANT EMERGENCY IMPLEMENTING PROCEDURE FNP-0-EIP-30 S

'A F

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' POST' ACCIDENT CORE DAMAGE ASSESSMENT 1.0 Objective-

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1.1:

The purpose of this procedure is to provide a method to classify and estimate the extent.of the core damage through measurement of fission products released to the coolant and containment atmosphere together with auxiliary measurements of core exit thermocouple temperature, containment radiation monitors, and containment atmosphere hydrogen monitors.

2.0 References 2.1 Westinghouse Owner Group Post Accident Core Damage Assessment Methodology, Revision 1, March 1984.

2.2 Westinghouse FNP Setpoint Calculations ALA-84-594 dated March 15, 1984.

2.3 Westinghouse FNP Post Accident Core Damage Assessment ALA-84-754 dated August 8, 1984.

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3.0 Responsibilities 3.1 The Technical Manager in the Technical Support Center will be responsible for core damage assessment based on radionuclide analysis -and auxiliary measurements. The On-call Reactor Engineer is responsible for_the detemination of core inventory.

The On-call Chemistry Supervisor is responsible for coordinating the sampling and analysis of the reactor coolant system and the ECCS Sump (via RHR). The On-call Environmental Supervisor has the responsibility of sampling and analyzing containment atmosphere and providing support for radiological analysis of reactor coolant and ECCS Sump samples.

4.0 Limitations 7

t 4.1 Due to the accuracy and errors associated with this assessment, estimates are limited to the following categories:

4.1.1 Less than 50% clad failure 4.1.2 Greater than 50% clad failure 4.1.3 Less than 50% fuel overtemperature L

4.1.4 Greater than 50% fuel overtemperature 4.1.5 Less than 50% fuel melt r

4.1.6 Greater than 50% fuel melt O

i Rev. 1 l

1

FNP-0-EIP-30 5.0'. Applicability 7_i. J 5.1 Any plant condition in.which inadequate core cooling is suspected..

5.2 Any plant condition in which failed fuel is suspected and an estimate of the amount of failed fuel is-required.

16.0 Instructions 6.1 1Nuclide Sampling 6.1.1-Obtain post accident samples from the following systems per FNP-0-RCP-25 as appropriate:

6.1.1.1 Reactor Coolant System.

6.1.1.2 Containment ECCS Sump (via RHR system during SI recire phase).

6.1.1.3 -

Containment Atmosphere.

6.1.2 Analyze the sc_lected samples for isotopic specific activity with no decay correccion applied to sample activities. Table 2 lists the selected nuclides for core damage assessment. Not all of the nuclides listed in Table 2 need to be analyzed but as many k

as possible should be analyzed to determine a reasonable approximation of core damage. A suggested minimum list of nuclides for this procedure are Xe-133, I-131, Cs-137, and Ba-140. To use the nuclide activity ratio characteristic of Section 6.7, the nuclide analysis should include another noble gas besides Xe-133 and another iodine besides I-131. However, if the above nuclides are unable to be analyzed, continue with the data that is obtainable. The nuclides listed in Table 2 characterize a mechanism of release associated with the core damage states. The lack or presence of particular nuclides in the sample indicates the upper bounds of fuel damage.

L 6.1.3 If sample is available, complete Table 3A, RCS Activity Worksheet as follows:

6.1.3.1 Record elapsed time from reactor shutdown L

to sample count.

6.1.3.2 Record specific activities of nuclides.

[

6.1.3.3 Determine and record decay correction i

factor using Table 4, Decay Correction A

Factor With Parent-Daughter Effect.

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l.

Rev. O i

2 i

r FNP-0-EIP-30

'6.1.3.4 Determine and record the corrected specific JN activity by multiplying the measured

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specific activity by the decay correction factor.

6.1.4 If sample is available, complete Table 3B, Containment ECCS Sump Activity Worksheet as follows:

6.1.4.1 Record elapse _ time from reactor shutdown to sample count.

6.1.4.2 Record specific activities of nuclides.

6.1.4.3 Determine and record decay correction factor using Table 4, Decay Correction Factor With Parent-Daughter Effect.

6.1.4.4 Determine and record the corrected specific activity by multiplying the measured specific activity by the decay correction factor.

6.1.5 If sample was available, complete Table 3C, Containment Atmosphere Activity Worksheet as follows:

6.1.5.1., Record elapse time from reactor shutdown (v")

to sample count, 6.1.5.2 Record specific activities of nuclides.

6.1.5.3 Determine and record decay correction factor using Table 4, Decay Correction Factor with Parent-Daughter Effect.

6.1.5.4 Determine and record the corrected specific activity by multiplying the measured specific activity by the decay correction 4

factor.

6.2 Liquid Mass 6.2.1 Estimate the total liquid mass by completing Table 5, Estimate of Total Liquid Mass Worksheet.

6.2.2 If both an RCS sample and a CTMT ECCS Sump sample were obtained, an estimate of the RCS water mass and containment water mass is needed.

Use Table 6, Estimate of RCS Water Mass and Containment Water Mass Worksheet to estimate the distribution of the water. Record the RCS mass in Table 3A and the containment mass in Table 3B.

6.2.3 If only one of the liquid samples (RCS or containment ECCS sump) was obtained, use the total liquid mass Rev. 0 3

i

FNP-0-EIP-30 calculated in 6.2.1 as the water mass associated

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with that sample. Record water in either Table 3A

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-(RCS) or Table 3B (containment sump).

6;3 Containment Volume 6.3.1 Since'the containment' atmosphere sample is collected

-at the containment building pressure and the sample t

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volume'is not corrected to standard conditions, no adjustment factor is needed to the containment free air volume.

(5.66 E10 cc per reference 2.2).

6.4 Total Activity Released 6.4.1 RCS~

6.4.1.1 Calculate total activity of each nuclide released to the RCS by multiplying the decay corrected specific activity by the RCS mass. Record in Table 3A.

6.4.2 Containment Sump 6.4.2.1 Calculate total activity of each nuclide released to the containment water by multiplying the decay corrected specific

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activity by the containment water mass.

Record in Table 3B.

6.4.3 Containment Atmosphere 6.4.3.1 Calculate total activity of each nuclide released to the containment atmosphere by multiplying the decay corrected specific activity by the containment volume.

Record in Table 3C.

6.4.4 Total Activity Released of Each Nuclide 6.4.4.1 Record in Table 7, Total Release Activity / Percent Released, the activity of each nuclide of each sample location.

6.4.4.2 Sua the activities of each nuclide of each sample to determine total activity released of each nuclide. Record in Table 7.

6.5 Total Core Inventory 6.5.1 Power History 6.5.1.1 Use Table 8, Power Correction Factor Worksheet, Step 1, to record the plant power history during a minimum of 30 days prior to shutdown.

Rev. 0 4

FNP-0-EIP-30 1

6.5.2 Power Correction Factor jy V

6.5.2.1 Follow Table 8, Step 2, to determine the power correction factors for the appropriate nuclides. Record in. Table 7.

6.'5.3 Corrected-Core Inventory 6.5.3.1 Determine and record in Table 7 the corrected core inventory for each nuclide by multiplying the equilibrium full power inventory (listed in Table 7) by the-power correction factor.

6.6 Estimation,of Percent Fuel Damage 6.6.1 Determine the percentage of the corrected core inventory released of each nuclide by dividing the total activity released by the corrected core 4

inventory. Record in Table 7.

4 '

6.6.2.

Using the appropriate core damage graphs, Figure 4

.through 16, determine the percent clad failure, fuel

' pellet overtemperature, and fuel melt as a function of the nuclide release percentage.

Use the average curve of the clad damage figures to determine

O cc* cia 4

- r* *i a a 1 6

P curves provide the bounds of the amount of damage.

Use the nominal curve of the fuel overtemperature/

melt figures to deterraine percent overtemperature/ melt.

The maximum and minimum curves provide the s

bounds of the amount of damage. Even though the bounds of the possible damages are presented, use the average / nominal estimations of core damage in this procedure. Record the percentages of clad damage, fuel pellet overtemperature, and fuel melt in Table 10, Core Damage Assessment Evaluation Sheet.

NOTE:

Iodine spiking should be considered for cases where the assessment is between no fuel damage and minor clad failure. If percent clad failure is not in agreement with values obtained from other nuclides, spiking may have occurred.

Refer to Figure 8 if this is the case.

6.7 Nuclide Activity Ratios 6.7.1 Determine the activity ratios for noble gases and iodines by completing Table 11, Nuclide Activity O

a ti Rei. 1 5

m FNP-0-EIP-30 6.7.2 Compare the calculated activity ratios'with a gap 7

?"F activity. ratios and fuel pellet ratios listed in Table

'11.

Calculated. activity ratios less than gap activity ratios are indicative of clad failures. 1 Calculated activity ratios, greater than. gap' activity ratios are -

indicative'of more severe failures (fuel overheat and fuel melt).-

6.7.3 Record in Table 10 the calculated' core. damage state.-

6.8 Auxiliary Indicators 6.8.1 Record any available evidence of core uncovery in Table 10.

6.8.2 Obtain core exit thermocouple readings and compare these. values with those listed in Table 12.'

Based on Table 12, Characteristics of Categories of Fuel Damage, record temperature in Table 10 under appropriate core damage state.

6.8.3 Obtain containment hydrogen concentration. Compare hydrogen concentration under appropriate core damage state.

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"'d'***"

c c = '**" "* '**"' 27 '-

determine extent of zirconium-water reaction. Record l

percentage.of zirconium-water reaction in Table 10.

6.8.5 Obtain the containment high range area radiation monitor readings and the time after-shutdown the readings were obtained.

Compare the readings.with Figure 18 to estimate the corresponding extent of core damage.

6.8.5.1 On Figure 18, find the point of intersection of lines created by the comparison of high range monitor reading with time after shutdown. Using the legend on Figure 18, estimate the extent of core damage by determining to which line the point of intersection is closest.

e.g. If intersection point occurred between line A and the 0.3% Noble Gas Release line, then the estimate l

of fuel damage would be in the range of 50% to 100% clad damage.

6.8.5.2 Record the monitor reading in Table 10 o

under the appropriate core damage state.

l 6.8.5.3 If containment high range area radiation monitor readings are taken at 10 hours1.157407e-4 days 0.00278 hours 1.653439e-5 weeks 3.805e-6 months Rev. 1 6

L I

FNP-0-EIP.

s after shutdown compare these values with 3j7-).

those on Table 12 to provide more insight to the extent of fuel damage.

6.8.5.4

If containment high range area radiation

-monitor readings are taken a time OTHER THAN 10 HOURS, the following methodology must be used to obtain new containment radiogas monitor ranges for Table 12.

6.8.5.4.1 on the X axis, locate time after shutdown corresponding to time of monitor reading.

6.8.5.4.2-Follow a verticle line from the-X axis to line A.

The point of intersection of this verticle line with line A is the upper boundary (R/hr) for 0-50%

clad damage. Indicate range on Table 12.

6.8.5.4.3 Continue tracing verticle line until intersection with "0.3%

Noble Gas Release" line occurs.

The point of this intersection

( ]).

is the upper boundary (R/hr) for 50-100% clad damage.

Indicate range on Table 12.

6.8.5.4.4 Continue tracing verticle line upwards until all upper boundaries (R/hr) have been determined.

Indicate each range on Table 12.

The significance of each line is noted below.

Line A: Upper boundary for 0-50% clad damage.

0.3% Noble Gas Release: Upper boundary for 50-100% clad damage.

Line B: Upper boundary for 50% fuel overtemperature.

52% Noble Gas Release Line:

Upper boundary for 100% fuel overtemperature.

Line C:

Upper boundary for 50% fuel melt.

Beyond Line C:

100% fuel melt.

Rev. 1 7

FNP-0-EIP-30 6.9 Core Damage Assessment

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6.9.1 Perform the final core' damage assessment by evaluating the data in Table 10.

It is unlikely that complete agreement between the indicators will result in the same estimate of core damage. The evaluation should be the best estimate based on all parameters, their

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The user should use as many indicators as possible.

to differentiate between the various core damage states. Because of overlapping values of release and potential simultaneous conditions of clad damage, overtemperature, and/or core melt,. considerable judgement needs to be applied.

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Gen. Rev. 1 l

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FNP-0-EIP-30 f s; TABLE 1 IU Sugaested Sampling Locations Principal Other

-Scenario Sampling Locations Sampling Locations Small Break LOCA Reactor Power > 1%*

RCS Hot Leg, Containment RCS Pressurizer Atmosphere Reactor Power < 1%*

RCS Hot Leg RCS Pressurizer Large Break LOCA Reactor Power > 1%*

Containment ECCS Sump (via RHR System), Containment.

I

--Atmosphere, RCS Hot Leg Reactor Power < 1%*

Containment ECCS Sump (via RHR System, Conteinment Atmosphere Stram Line Break RCS Hot Leg, Secondary System RCS Pressurizer Containment 6

Atmosphere b Steam Generator Tube RCS Hot Leg, Secondary Containment Rupture System Atmosphere Indication of Signifi-Containment ECCS Sump (via RHR

= c:nt Containment Sump System), Containment Inventory

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Atmosphere C:ntainment Building Containment Atmosphere R:diation Monitor Alarm Safety Injection RCS Hot Leg RCS Pressurizer j

Actuated l-Indication of High RCS Hot Leg RCS Pressurizer Radiation Level in RCS

Assume operating at that level for some appreciable time.

o Rev. 0

FNP-0-EIP-30 TABLE 2

)

SELECTED NUCLIDES FOR CORE DAMAGE ASSESSMENT

-Ctre Damage State Nuclide Half-Life

Predominant Gammas (Kev) Yield (%)*

Clad Failure Kr-85m**

4.4 h 150(74), 305(13)

Kr-87 76 m 403(84), 2570(35)

Kr-88**

2.8 hr 191(33), 850(23), 2400(35)

Xe-131m 11.8 d 164(2)

Xe-133 5.27 d 81(37)

Xe-133m**

2.26 d 233 (14)

Xe-135**

9.14 h 250(91)

I-131 8.05 d 364(82)

I-132 2.26 h 773(89), 955(22), 1400(14)

I-133 20.3 h 530(90)

I-135 6.68 h 1140(37), 1280(34), 1460(12), 1720(19)

Rb-88 17.8 m 898(13), 1863(21)

Fuel Overheat Cs-134 2 yr 605(98), 796(99)

Cs-137 30 yr 662(85)

Te-129 68.7 m 455(15)

Te-132 77.7 h 230(90) ll)FuelMelt Sr-89 52.7 d (beta emitter)

Sr-90**

28 yr (beta emitter)

Ba-140 12.8 d 537(34)

La-140 40.22 h 487(40), 815(19), 1596(96)

La-142 92.5 m 650(48), 1910(9), 2410(15), 2550(11)

Pr-144 17.27 m 695(1.5)

Values obtained from Table of Isotopes, Lederer, Hollander, and Perlman, Sixth Edition.

- These nuclides are marginal with respect to selection criteria for candidate nuclides; they have been included on the possibility that they may be detected and thus utilized in a manner analogous to the candidate nuclides, p,

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. TABLE 4 gq]c'

. DECAY! CORRECTION F CTOR*-

WITH PARENT-DAUGHTER EFFECT N'uclide' Correction Factor

' 'Kr 85m.

. [01158tl 0.5.47 t1 Kr 87"

' e 0.248t.

Kr 88

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-Xe 131m--

1/ [-2'.66;(-3.'59E-3)t,- 3.66e(-2.45E-3)t -.

y Xe'133 1/ [-0.'187e(-3.41E-2)t ~ - 0.10e(-5.48E-3)t,.1.287e(-1.28E-2)t..

j Xe 133m-1/ [-0.10e(-3.41E-2)t.+ 1.11e(-1.28E-2)tj:

Xe-135 1/[-9.14e(-1.04E-1)t - 0'.033e(-2.67)t +-10.17e(-7.58E-2)tj fI-131 e(3.59E-3)t-I: 132 >

1/[1.03e(-8.92E-3)t - 0.03e(-3.07E-1)tj

D-I 133-e(3.41E-2)t-o 0.104t

ya135 e

1/ [1.10e(-0.248)t - 0.10e(-2.34)tj Rb"88 -'

'Cs 134 1.0 i

l Cs 137

'1.0 (0.605)t f.

Te 129 1/[1.09e(-0.161)t + 0.167e(-8.47E-4)t-0.257e j

Te 132-

'e(8.92E-3)t Ba 140 e(2.26E-3)t La 140 1/[1.08e(-2.26E-3)t - 0.08e(-1.72E-2)t -

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La 142 1/[-0.145e(-3.78)t + 1.145e(-0.450)tj Pr 144 1/[0.909e(-1.02E-4)t + 0.091e(-241)tj i* Time, t, is the number of hours between shutdown and time of sample count.

TCN 1A

FNP-0-EIP-30

]rj A./.

TABLE 5 ESTIMATE OF TOTAL LIQUID MASS 1.

Estimate the volume added for the following:

Estimated Volume Maximum Volume Tank Added Added (Rallons)

a.

Refueling Water Storage Tank 450,000(with CS) 350,000(without CS) b'.

' Accumulator A 6920 c.

Accumulator B 6920

'd.

Accumulator C 6920 e.

-Boron Injection Tank 900 f.

SecobdaryWater 2500 l

g.

Other source Total 2.

Convert estimated volume added from gallons to grams.

Added volume:

gallons x 3785 gas / gal =

gas l

3.

The average Reactor Coolant System Mass is 1.87 x 10s g,,,

4.

Determine the Total Liquid Mass as follows:

Mass added gas + RCS mass 1.87 x 10s gms =

gas l

O i

i Rev. 1

m -y _

FNP-0-EIP-30 TABLE 6-l.)

ESTIMATE.'0F RCS WATER MASS AND CONTAINMENT WATER MASS

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AVERAGE' OPERATING RCS VOLUME = 9146'ft3

-1.

Record the containment ECCS sump level.

ECCS Sump level =

ft 2.

Determine ECCS sump water volume from Figure 2 using the level from Step 1.

ECCS Sump Water Volume =

ft3 3.

Determine ECCS Sump specific gravity.from Figure 1 using RHR inlet temperature.

"ECCS Sump specific gravity =

4.

Determine containment water mass as follows:

8 ECCS Sump volume x specific gravity x 1.0 ga x 28.3 x 10 cc 3

cc ft,

3 ft3 x 1.0 ga x 28.3 x 10 cc =

gas x.

cc ft8 5.

Determine RCS maas by subtracting ECCS Sump Mass (Step 4) from Total Water Mass (Table 5).

Total Water Mass =

gas 3

ECCS Sump Mass =

gas RCS Mass =

gms O

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a.

a r

B NK K

E X

X X

I 1

I R

C C

T T

B 1

L P

FNP-0-EIP-30 TABLE 8 POWER CORRECTION FACTOR WORKSHEET I

T ' 1..

Record Power history by interval _for the 30 day period prior to shutdown.

_ Choose intervals such that power does not vary more than +10% from the average power of the interval.

D j

j j

t*j Interval Average power of Length.of interval Time between end of number interval in in hours interval and reactor megawatts thermal shutdown time in hours:

(100% power =

2660 MWT).

1 2

3 4

5 6

7 8

9 10 11 12 13 14

-15 16 17 18 19 20 21 22 l

23 24 25

(

2.

Table 9 presents the half-lives and the decay constants (A ) of each nuclide.

d i

Based on the letter shown in column 2 of Table 9, refer to that letter of this worksheet to determine the power correction factor for that particular nuclide.

A.

Half-life of Nuclide 1 < 1 Day For these nuclides, only the power history of the prior 96 hours0.00111 days 0.0267 hours 1.587302e-4 weeks 3.6528e-5 months (4 days) prior to shutdown is required to be considered.

A1.

Has the power remained constant over these 96 hours0.00111 days 0.0267 hours 1.587302e-4 weeks 3.6528e-5 months ?

If so, the power correction factor (PCF) is calculated as shown below.

If power has not 7_

.been constant, go to A2.

l g

l l

PCF = Average Power (MWt) 2660 MWt Page 1 of 4 Rev. 0

FNP-0-EIP-30 TABLE 8 (Cc2tinued)

POWER CORRECTIGN FACTOR WORESHEET

,D As

~~

A2. JSince the power has not remained constant over the 96 hours0.00111 days 0.0267 hours 1.587302e-4 weeks 3.6528e-5 months , the power.

correction factor-is calculated as follows:.

-A t

-A t'.

PCF = I P.(1-e d) e I d 3

2660 W t where P, 't),: and t* are found' from step 1, and Ag (decay constant in 3

j hours" of the particular nuclide) is found from Table 9.

Table 8A-as'a guide in calculating the-PCF for each nuclide from the above equation.

B.

Half-life of Nuclide > 1 Day For these nuclides, the entire power history shown in step 1 (minimum time of 720 hours0.00833 days 0.2 hours 0.00119 weeks 2.7396e-4 months ) is required to be considered.

Bl.

Has the power remained constant over these 720 hours0.00833 days 0.2 hours 0.00119 weeks 2.7396e-4 months ? If so, the PCF is calculated as shown below.- If power has not remained constant, go to B2 PCF = Average Power (Wt) 2660 Wt E2. 'If the power has not remained constant over the 720 hours0.00833 days 0.2 hours 0.00119 weeks 2.7396e-4 months , the PCF is

. calculated as follows:

-A.t

-A t

PCF g I P3 (1-e

d) e i d 2660.Wt where P, t, and t are found from step 1, and Ag (decay constant in 3

j 3

-hours of the particalar nuclide) is found from Table 9.

~I Table 8A as a guide in calculating the PCF for each nuclide from the above equa tion.

C.

Power Correction Factor for Cs-134 and Cs-137.

C1.

Cs-134 1

Determine the average power percentage during the entire operating period during the cycle prior to shutdown.

Page 2 of 4 Rev. 0

.L FNP-0-EIP-30

' TABLE'8 (Ccatinu:d).

- ~.

POWER CORRECTION' FACTOR WORKSHEET Power Percentage '= Averate Power for Cycle Wt 6

2660 We Use the power percentage and Figure 3 to determine PCF for Cs-134.

Power Percentage =

~%

PCF =

C2.

Cs-137-Determine the PCF for Cs-137 from the following equation.

-PCF = EFPD(total for previous two cycles)+EFPD(actual for present cycle)

EFPD(total for. previous two cycles)+EFPD (design for present cycle)

-PCF =

EFPD =

EFPD Example:

The core is in its third cycle of an expected equilibrum cycle operation of 1075 EFPD. For the first two cycles the core has operated for a total of 600 EFPD. Prior to shutdown in the third cycle, the core has operated for 75EFPD. The power r.

correction factor (PCF) would be calculated as follows:

PCF = 600 + 75 EFPD = 0.63 1075 EFPD l

t l'

1

{

{-

l 1

l l

I i

i lO i

l L

Page 3 of 4 Rev. 1

O l'.)

. CL;.

r=e-0-=1P-30 TABLE 8A PCF WORKSHEET FOR FLUCTUATING POWER.

t?

PCF =.I P3 (1-e i 'j) e i J

~

2660 W t 4

Nuclide, i =

A E hours' g

0

.A t

d d

.A' 3

l P., HWt t

?

A = 1-e B

P x B, NWt J

j, hours tj, hours j

j=e jxAJ j

i i

I

=

4

Susg;this column to get.I P3 x A) x B) xA xB.

PCF =

j 3

3_

ggt _

l 2660 MWt 2660 MWt -

o Note: P.t, and t are found from step 1 of Table 8.

JlsfoundfromTable9.

3 l

Ag Page 4 of 4 Rev. O

p i

FNP-0-EIP-30

- g,.

r~',.

TABLE 9 J-DECAY CONSTANT (A ) 0F EACH NUCLIDE g

i Nuclide Type Half-Life i, (hours ~1) 1 Kr 85m

-A 4.4 h 0.158 2

Kr 87 A

76 m 0.547 3

Kr 88 A

-2.8 h 0.248 4-Xe 131m B

11.8 d 2.45 -3 5

Xe.133 B-5.27 d 5.48 -3 6

Xe 133m B

2.26 d 1.28 -2 7

Xe 135 A

9.14 h 7.58 -2 8

I 131 B

8.05 d 3.59 -3 9

I 132 A

2.26h 0.307 10 I 133 A

20.3 h 3.41 -2 11 I 135 A

6.68 h 0.104 12 Rb 88 A

17.8 a 2.34 13-Cs 134 C

2 yr 3.96 -5 14 Cs 137 C

30 yr 2.64 -6 15 Te 129 A

68.6 m 0.605 16 Te 132 B

77.7 h 8.92 -3 17 Ba 140 B

12.8 d 2.26 -3 18 La 140 B_

40.22 h 1.72 -2 19 La 142 A

92.5 m 0.450 20 Pr 144 A

17.27 m 2.41 Rev. O

1.

FNP-0-EIP-30 4

i TABLE 10 V

CORE DAMAGE ASSESSMENT EVALUATION' SHEET-Percent Clad Percent Percent Indicator.

Damage Overtemperature Fuel Melt

< 50%

> 50%

< 50

-> 50%

< 50%

> 50%

Radionuclide Analysis

'Kr 85m Kr 87 Kr 88 Xe 131m Xe 133 Xe 133m Xe 135 I 131 I 132

(}

I-133 I 135 Rb 88 Cs 134 Cs 137 Te 129-Te 132 Ba 140 La 140 La-142 Pr 144 Ratios Kr 85m/Xe 133 Kr 87/Xe 133 Kr 88/Xe 133 Xe 131m/Xe 133 l

Page 1 of 2 Rev. O Rev. O

7-d}-

FNP-0-EIP-30

.v W

,-(

TABLE 10 (continued).

'Sl p

CORE DAMAGE ASSESSMENT EVALUATION SHEET Percent Clad Percent Percent Indicator Damage Overtemperature Fuel Melt

< '50%

> 50%

< 50

> 50%'

< 50%

> 50%

~

Ratios _(Con't)

-Xe 131m/Xe 133 Xe 135/Xe 133' I 132/I 131 I 133/I 131 I 135/I 131

-Auxiliary Indicators

('])

R.

Core Uncovered Core-Exit Temp 0F Containment H,%

Zire - Water Reaction %

Has a hydrogen burn occurred as indicated by a containment f

pressure spike?

High Range Containment Monito-Reading R/hr Page 2 of 2 Rev. O

p:

FNP-0-EIP-30 j

,y TABLE 11 NUCLIDE ACTIVITY RATIOS

' Total Isotopic Calculated Ratio Gap Fuel Pellet Activity of Noble Gas-Nuclide' Activity Ratio Activity Ratio (Table 7 Column 4) to Xe-133 Xe 133 1.0 1.0 1.0 Kr 85m 0.026 0.13 Kr 87-0.025 0.25 Kr 88 0.054 0.35 Xe 131m 0.004 0.004 TN Xe 133 1.0 1.0 V

Xe 133m 0.103 0.15 Xe~135 0.061 0.23 Calculated Ratio of Iodine Isotope to I-131 I 131 1.0 1.0 1.0 I 132 0.17 1.5 I 133 0.71 2.1 0

1 135 0.39 1.9 Rev. 1

FNP-0-EIP-30 TABLE 12 CHARACTERISTICS OF CATECORIES OF FUEL DAMACE Containment Radiogas Monitor (R/hr)

Containment Atmosphere Percent 10 hrs after shutdown

Core Exit Hydrogen Monitor and Type Fission Product RE-27 A or B Thermocouples Core Monitor Core of Fission Ratio (Highest reliable)

Readings (Des F)

Uncovery (Vol % II,)

Damage Products Kr-87 or I-133 QID2 IRE 0027(A or B)

NIC56C00(IB-518) Indication QlE23AIT2703(A or B)

Category Released Xe-133 1-131 Q2D21RE0027(A or B)

N2C56000(IB-51R) (RVLIS)**

Q2E23AIT2703(A or B)

No clad damage Kr-87 < Ix10~8 Not applicable

< 750 No uncovery Negligit-Ie 8

Xe-133 < lut0 I-131 < 1x10~8 l-133 < 1x10~8 0-50% clad damage Kr-87 10[8-0.01 Kr/Xe = 0.022 0 - 150 750 - 1300 Core uncovery 0-7 Xe-133 10 s - 0.1 I-131 10~8 - 0.3 133/131 = 0.71 1-133 10~8 - 0.1 50-100% clad damage Kr-87 0.01 - 0.02 Kr/Xe = 0.022 150 to 300 1300 - 1650 Core uncovery 7 - 14 Xe-133 0.1

- 0.2 I-131 0.3 - 0.5 133/131 = 0.71 I-133 0.1

- 0.2 0-50% fuel pellet Xe-Kr, Cs,1 Kr/Xe = 0.22 300 to 2.5 E4

> 1650 Core uncovery 7 - 14 overtemperature 1 - 20 Sr-Ba 0 - 0.1 133/131 = 2.1 50-100% fuel pellet Xe-Kr,Cs,I Kr/Xe = 0.22 2.5 E4 to 5.0 E4

> 1650 Core uncovery 7 - 14 overtemperature 20 - 40 Sr-Ba 0.1 - 0.2 133/131 = 2.1 0-50% fuel melt Xe,Kr,Cs I 40 - 70 Kr/Ke = 0.22 5.0 E4 to 7.0 E4

> 1650 Core uncovery 7 - 14 Sr-Ba 0.2 - 0.8 Pr 0.1 - 0.8 50-100% fuel melt Xc,Kr,Cs.I,Te Kr/Xe = 0.22 7.0 E4

> 1650 Core uncovery 7 - 14

> 70 Sr,Ba > 24 133/131 = 2.1 Pr > 0.8

Values should be revised for times other than 10 hours1.157407e-4 days 0.00278 hours 1.653439e-5 weeks 3.805e-6 months . Reference section 6.8.5.4 for instructions.

- Reactor Vessel Level Indicator System - Committed for installation by 8th refueling outage for Unit I and 5th refueling outage on Unit 2.

Rev. 1 f

1 1

e

1 a

s FNP-0-EIP-30 J '

,g

\\J-O 400.

\\

s s

s 350.

h s

c.

300.

N 250.

N O

200.

g

\\

t

\\

150.

\\

100.

50.

0.

k 9

k k

o o

i l

p/o

[

STP l

FIGURE 1: WATERDENSITYRATIO(TEMPERATUREVS.STP) i Rev. 1

FNP-0-EIP

T 1

r 80000.

s 70000

/

60000.

^

m 4 tt

50000, m

/

~

40000,

/

a:

o n

t

30000, m

/

3

- g e

20000, f

~

10000, 0

d 4

ci M

4 c'

ai d

d w

ECCSSUMPWATERLEVEL(FT) 1 S

FIGURE 2 ECCS SUMP WATER VOLUME VERSUS ECCS SUMP WATER LEVEL O

t Rev. 1

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

m FNP-0-EIP-30 f3-a 1.0 l......._...-

0.9 30%POWElt f

~

.. L A

0.8 n,.7 y ygg.pnyg g 0.6 E

g' 0.5 g

0.4 50% POWE R

/

0.3 s

o

~

j

0. 2 1

0.1 0.0 1

0 200 400 600 800 1000 CYCLEOPERATION(CALENDARDAYS)

FIGURE 3: POWER CORRECTION FACTOR FOR CS-134 BASED ON AVERAGE POWER DURING OPERATION Rev. I

FNP-0-EIP-30 C

a t

0.-l.

.05 j

.. i

/

.01

'005

/

-r 1

e

/

r e

T

.001

/

e e

5

{O j

5.0E-4 i.

4f j

i b

/

l s*

a

~

1.0E-4 l

g 0

_4,

i o

L v

5.0E-5

\\

/

7

-~

l l

1.0E-5 i

l-t m

M o

o W

O

.c 8

i.

CladDamage(%)

FIGURE 4:

RELATIONSHIP OF % CLAD DAMAGE WITH % CORE INVENTORY i

RELEASED OF KR-87 s

I.

l l

Rev. I l

..--... ~

r f

FNP-0-EIP-30

/~h

lj i

1.

- ~~

1 --

---:==-T 1

1 l

0. 5

~

e d

Y

/

s

.- i

/

0.1

[

/

M e

wo

/

i

/

I

.05

.y

/

.e E

Q.

b

$s 2l

$,4,

/

w,

.UL a

v

@gSQ*,

E i

o

/

e

.005

-r p,

t.o*

^

.001 d

N

~

o o

o m

~

CladDamage(%)

l, FIGURE 5:

RELATIONSHIP OF % CLAD DAMAGE WITH % CORE INVENTORY O

RELEASED OF XE'-131M Rev. I

E

.t-FNP-0-EIP-30 O

v 1.

0. 5

~,

P

\\

/

7 0.1 i

r r.

2 z.

i

/

y

.05

/

O s

/

2

/

s

/

O p$ ' '

/

.01 s

A

/

}

/

a af p

/

ds

.005

/p/

g

/

g#f,,

p

.001 d

A A

Y CladDamage(%)

FIGURE 6: RELATIONSHIP OF E CLAD DAMAGE WITH % CORE RELEASED OF XE-133 Rev. 1

7.

j FNP-0-EIP-30 O

i I

. 1.

0.5

^ '

2

., p r

s

/

0.1 l

.05

/

f a' s y

s 9

l r

~

p$'

/

' s' g

Y,4'&,-

l

.02 r

.}

'.' 005 f-

}

/

kfe '

s#p

.001 _'

/

i 4

5.0E-4 S

a s'

e em r

e a

E s

8 o

1.0E-4

s l

5.0E-5 1.0Ec5-i in q

ci. '

'A d

f g

g o

CladDamage(%)

FIGURE 7: RELATIONSHIP OF 5 CLAD DAMAGE WITH % CORi i

RELEASED OF I-131' Rev. 1 i

l l

.--------2

s r

\\

c FNP-0-EIP-30 s.

1.

i 3

0. 5 *

/

~

s f

f

/

0.1

~

.05

^

/

4,/, '

p e

.g a

4 3

~

/

s 01

/ p_ '

Y,e.

a.

g

.00s

/

Lo I

'+W y

d s'

l 5

.001 j.

j-5.0E-4 o

1.0E-4 i-5.0E-5 1.0E-5

~

m d

6 J

g S

8

~

CladDamage(%)

FIGURE 8: RELATIONSHIP OF %* CLAD DAMAGE WITH % CORE INVENTORY

,O RELEASED OF I-131 WITH SPIKING Rev. I

r-FNP-0-EIP-30 l

.y I

T

,,,.---.=r..

m.g _.

0.1

.05 e

7

/

- -j s.

(<..

,/

/

~,

=

we r

1

.005 s'

)

s Q

/

f.

r.

e i,

/

g 2

/

of,.

p s

/

j j

=

.001 e

E sh,,,

c 5.0E-4

.r s

g

/

o.

e

/

u

/

,/

/

c j

/

s o

/

,/

d /

l 1,0E-4 y'

1 1

5.0E-5 s

/

1.0E-5 m.

o o

m o

o 8

m

)

.=.

Q Clad Damage (%)

~

l FIGURE 9: RELATIONSHIP OF % CLAD DAMAGE WITH % CORE INVENTORY l

RELEASED OF I-132 Rev. 1 m

..-..m.

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

.. =..

. _ ~.

4..

FNP-0-EIP-30 "O.

I 1.

0.5 4

.1 s

~

^

.05 5

~

/

s 3!

.01 e

[

sd g~

^

4D j 4/

p s

a

.001

~

E

~

o 5.0E-4 s&<,-

s j

r

/

1.0E-4 l

5.0E-5

~

=

1.0E-5 o

o m

o

?

S 8

~

.n l

CladDamage(%)

O FIGURE 10: RELATIONSHIP OF % CLAD DAMAGE WITH % CORE INVENTORY i

RELEASED OF I-133 Rev. 1 1

.----.-. - -.~-.~.-

n.-

FNP-0-EIP-30' O'

1.

0.5 0.1

.05 4

/

f

_ 21

- t'

.01 3

+,* 4ofS

.005 O

T, g

.001

'g1Sp Rs c

5.0E-4 E

4,p _

s e

a

/

y&, '

1 a

/

1.0E-4 5.0E-5 i

1 1.0E-5 i

m d

d 5!

f I

c O cd

'**S' (5)

' FIGURE 11: RELATIONSHIP '0F % CLAD DAMAGE WITH % CORE INVE

, RELEASED OF I-135 i

I

,. _--e we e m,w

FNP-0-EIP-30 4

1. J

+

- 100.

- 50.

~

/

r

/

f.

?

/

/ l-. *,

g.

~ ~

/

i 10.

/

s

'sh?'

5.

/*b y

/

r p

,/

3 f

s a

/

v

/,f

/

c

/

s E

,' /, '

o

/

1.

/

O

/

r 0.5'

+

/

0.1 l

1 o

d d

m o

m Fuel Overtemperature (%)

' O FIGURE 12:

RELATIONSHIP OF % FUEL OVERTEMPERATURE WITH %

CORE INVENTORY RELEASED OF XE, KR, I, OR CS r

l Rev. 1

!1.--,.-,n.,--

-. ~. - - - ~ ~ ~ ~ ~ ~ ~ ~ - - - - - - + - - - - - - - - - - -

~~---- '

n.-

-,,-- -.v. - _

FNP-0-EIP-30 J

-1.

~

0.5 s

/

1 0.1 r

m, e

r s

~

. 05 --

g g

/

a:

c

\\g\\f,'

, /

f~

s

.01

/

,o O"~

.E s'

.5

.005 f

E

_- '/

l 0

s '/

/

c i

.001 o

5.0E-4 l

l 1.0E-4 b

g t

Fuel Overtemperature (%)

FIGURE 13:

RELATIONSHIP OF % FUEL OVERTEMPERATURE WITH %

CCRE INVENTORY RELEASED OF BA OR SR Rev. 1 L

p

.s.

T FNP-0-EIP-30

..Q 100.

s g

.50.

,;r a

,'/.e' f

/

e, /

[

p+\\.,#e jp* '

o s.e

/

i 10.

/

,- '_ 9'Nd J

i 4

.Q 5.

/

/ / <.,'

F '

3 i

b B

/

s c.

s y

,/

3 c

y

.o.

v-0.5 1

i t

0.1 I

'5!

S f

FuelMelt(%)

O FIGURE 14: RELATIONSHIP OF % FUEL MELT WITH % CORE INVENTORY RELEASED OF XE, KR, I, CS, OR TE Rev. I f

FNP-0-EIP-30

-100.0 50.0

...t.....

y y

s

/

10.0

/

-T n

P 5

5.0 t

s

'h

,e

/

s s

/

t d

/

(

}

/

$fi

^

i

/

1,0 O

\\

1 0.5~

t

/

i y

/

8

_s'

/

t

'/

o-0.1 l

0.05 l

0.01 1.0 5.0 10.0 50.0 100.-0 l

t FuelMelt(%)

l FIGURE 15: RELATIONSHIP OF % FUEL MELT WITH % CORE RELEASED OF BA OR S,R l

O Rev. 1 i

e N

i

r s

FNP-0-EIP-30 t

1 100.0

' " ~ '

" " ~ " ~

5.0 10.0

.5.0 9

7, 3

1.0 y

n*5 D

~\\

s

=3

.s sy s

,,1 4?!

.s i

S 0.05

' s' s

/

o f

0.01 l

l 0.005 o

0.001 1.0 5.0 i0.0 50.0 100.0 t

l FuelMelt(%)

l-FIGURE 16:

RELATIONSHIP OF % FUEL MELT WITH 1 CORE INVENTORY RELEASED OF PR Rev. 1 L

g 1.

(>-

/,;

FNP-0-EIP-30 rT

')

F;

.a,.

s 15.

p s.

k i

12 5 g

/

.s

/

lo.

E

/

c 8

O i

/

7.,

5

/

l 3

o

/

5.

/

t 25

/

l

(

I I

l i

o.

(

l 2

2 2

?

A 3

's i

s

~

O ziac area acactica acace race FIGURE 17 CONTAINMENT HYDROGEN CONCENTRATION BASED ON ZIRCONIUM WATER REACTION I

Rev. 0 1

J

y FNP-0-EIP-30 1.0E+6 gain um nu a a a u i mimus uma su a a u um muu in a a u I E I w

- a i

1.0E+5

.o:

n, obi,_g _

-m s i h I.00% Noble Gas Release i ma ni a m u s i !

o m 1 OE+4 h

maim umu mm e n e n imm um m e i h i umu mm mm u un i

g N

L

~

v x

w h

5 E E R I ! ! !-

M E. E l l ! ! M M gCi m i r g

E N

t g

1.0EM!

,,0.3% Noble Gas Re lease.

B.

{

. M imu D E u II muu mu u I I I I:

mumm uma um a nu n l I

-g

'~

i I

1.0Et1 3

,,,,,,,, i i m,, e m,,,,

i gang, agg b, g, i n i '*

h w

[

u

.0

& g g g g i,

M No e Gas e se e ' ' ' ' " '

g

=

x 0

1. J 2 s n=

........, i

=

s 3

1.0E-2

\\

1 10

'100 1000 t

TIMEAFTERSHUTDOWN(HOURS) e FIGullE 18: PERCENT NOBLE GASES IN CONTAINMENT Line At 50% clad damage.

l 0.3% Itoble Cas Release line:

100% clad damage.

I Line 1:

50% fuel overtemperature.

52% Noble Gas Release line:

100% fuel overtemperature.

Line C:

50% fuel melt, i

l 100% W ble Cas Release line:

100% fuel melt.

n l

I l

l Rev. 1 O

l 1

'9-l

}

~

m

~

_v

, ~ <

^

~

" W -:

Alabama Power Company -

- 600 North 18th Street.

O

. Post Office Box 2641 g

. Birmingham. Alabama 35291 Telephone 205 783-6090 R. P. RecDoneW

. Senior Vice President ^

Flintridge Building AlabamaPbwer

^

. theso:,t%rrte&217t. sit"n

' December 14,~1984 i

+

Docket Nos. 50-348 -

s 50-364 ai Director, Nuclear Reactor Regulation i

U.- S. Nuclear. Regulatory Commiission Washington,-D.C. 20555-Attention: Mr. S. A. Varga e

' Joseph M. Farley Nuclear Plant --Units.1 and 2 s

NUREG-0737, Item II.B.3, Post Accident Sampling System Modifications-Gentlemen:

i /

JBy letter of October 11,1984, the NRC requested a copy of the Farley Nuclear Plant final procedure for estimating the extent of core

' damage following a severe reactor accident. A copy of this procedure is attached as requested.

If you have any questions, please advise.

Yours truly, J

i R. P. Mcdonald RPM / JAR:bdh-D6 Attachment cc: Mr. L. B. Long i

Mr. J. P. O'Reilly Mr. E. A. Reeves Mr. W. H. Bradford 8

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ML20101E606

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