Text

Rev 0 to Procedure EDP-ZZ-00005, Assessing Core Damage
	ML20084C203
Person / Time
Site:	Wolf Creek, Callaway, 05000000
Issue date:	04/10/1984
From:	Bryant K; STANDARDIZED NUCLEAR UNIT POWER PLANT SYSTEM
To:	;
Shared Package
ML20084C180	List: ML20084C203; NRC-84-0074, Forwards Rev 0 to Procedure EDP-ZZ-00005, Assessing Core Damage, Re Capability of post-accident Sampling Sys to Meet NUREG-0737,Item II.B.3 Criteria Per Sser 4.Deletion of License Conditions Re post-accident Sampling Requested;
References
	EDP-ZZ-00005, EDP-ZZ-5, NUDOCS 8404270150
	Download: ML20084C203 (35)
	v • d • e

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>* NUN #O I

' EDP-ZZ-00005 March 19, 1984 Revision 0 CALLAWAY PLANT Engineering Departmental Procedure EDP-ZZ-00005 ASSESSING CORE DAMAGE

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RESP. DEPT. U_dwtCA 'h6 PREPARED BY ? N ' J -co- A v APPROVED BY A/. 8 . h% M DATE f -2 0-PV DATE ISSUED //h0 $l INFORMATION O'NLY This procedure contains the following:

UNCONTROLLED 00?, Y Pages 1 through 5 Attachments 1 through 23 Appendices through checklist through SubY t

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EDP-ZZ-0000!

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l . ASSESSING CORE ~ DAMAGE s 1.0 PURPOSE AND SCOPE 1.1 This procedure providesf a methodology for .

determining.the extent of core damage fol-I s lowing an s accident using the Post Accident

, _ Sampling Oystem (PASS). Prehminary esti-mates may also be made based on H2 concen-tracion is the-containment, core exit'thort1 -

mocouple readings, reactor vessel water ,

- lefel, an'd containment radiation; readings.

7

. . 'i .s ,

\

2.0 DEFIN?TIONS ,

s q

s s *

.\- ,

s Clad,danige - Clad damage is char'act'srised

~

l 2.1 l by thes re. ease of! fis'sion product.s which s l' have hecumulated in the gap betweten the clad' '

s' and'the fuel. The fission products which diffusd to'this' gap are the volatile.ones

'suchMsi thafnoble, gasee.; the'iodidad','. and - '

i the 6esiums. '%-

N ' ... , ,. e 2.2 " Fuel. . overheating - Fuel 6verhdating is- V- N

- charai;terized by grain b6undary release and

%diffusi.on from the UO 2 gra' ins . ~ This is es timated to be 20-40% ofi thP noble gas, ,3 ~

,s iodine an,d, cesium invbiltorieE .

s.

C'

. ~ . ,

4 ..

Fuel melt T F uel meltd eads,to rapid release '

2.3 ~

s i 'of many noble gases, halides..andycesiums .j remaining in the fuellafter dverheating.

Significant release of the strontium and - . g )i barium-linthanumyioups"distinysishesthi('ss . .

condition. . g

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i NOTES AND PRECAUTIONS g[- s

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3.1 This procedure may be'cof.iied J so that it'can H be usedi more than once.1 ' Attachment 1 Mill ,

have to be copied for ea6hs isotope-te bag used in the analysis.- ~M\

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a 0 Proced. No. EDP-ZZ-00005

.', Rev. O l

3.2 During accident conditions, it is not known in what order that information will become available. Therefore, this procedure does j not have to be completed in the order that i it is written.

3.3 If hydrogen recombiners or the hydrogen r

purge system are operating, core damage es-timates based on hydrogen in the containment

, may be inaccurate.

3.4 Use as many indications as possible to diff-erentiate between the various core damage states. Because of overlapping values of release and potential simultaneous condi-

tions of clad damage, overtemperature, and core melt, considerable judgement needs to

, be applied.

4.0 PROCEDURE 4.1 Obtain an estimate of core damage using con-l tainment hydrogen concentration,. core exit

! " ' thermocouple readings, reactor vessel water

[" -level, and the containment radiation monitor.

4.1.1 Hydrogen Concentration 4.1.1.1 Record containment hydrogen concentration.

L --

~ 4.1.1.2 From Attachment 9, obtain the %

zirconium-water reaction and record here.

4.g .

[

4.1.2 Core Exit Thermocouple Readings 4.1.2.1 From Attachment 8, estimate the core damage based on core exit thermocouple readings.

{. Core damage:

kB 4.1.3 Reactor Vessel Water Level v

2 3N.s Q

.)

b y'- a 2 3 ; q.

L * ' aq ' (.

I $ k '~

, \

, Proced. No. EDP-ZZ-0000!

Rev. 0 4.1.3.1 Record the duration and extent of core uncovery.

Duration: minutes Extent: %

4.1.3.2 From Attachment 8, estimate the core damage based on core uncovery.

Core damage:

4.1.4 Containment Radiation Monitor 4.1.4.1 Record the Containment Radiation Monitor level R = R/hr.

4.1.4.2 Record the average power during entire period of operation ( from Attachment 2 )

P= %

4.1.4.3 Calculate the normalized dose rate 100%

Normalized Dose Rate = 2.34 x'10 4 xRx P =

R/hr -MWt 4.1.4.4 Record the time since the accident hours.

4.1.4.5 Using Attachment-10' estimate the core damage.

Core damage:

4.2 Estimation of core damage using PASS sample results.

4.2.1 As sample results become available, complete a copy- of Attachment 1 for each isotope. EI f '

an estimation of core damage was made in 4.1, then preference should be.given to--

those' isotopes which are indicative of that type of core damage. Attachment 3-provides a-list for this purpose.

,s e

' ~

Proced. No. EDP-ZZ-00005

.' Rev. 0 4.2.2 Using the percentage of inventory released and the fission product ratio from Attach-ment 1, and using Attachment 8 and 11 to 23, estimate the damage and record below.

I ! ! !

! ! ! Fission ! Estimatec Sample ! Fission ! Percentage ! Product ! Core Time ! Product ! Released ! Ratio ! Damace

5.0 REFERENCES

5.1 WOG-84-111', Draft CDA Methodology

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

i- ' ,

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l

. Proced. No. EDP-ZZ-00005 ;

Rev.- 0 l

5.2 FSAR Table 6.2.2-6 5.3 Table of Isotopes; Lederer, Hollander &

Perlman O

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a ~

h w

n .

7 4.

% (,

i ,4 Proced. No. EDP-ZZ-00005 Rev. O CALCULATION OF PERCENT OF CORE INVENTORY RELEASED 1.0 Isotope 1.1 Decay constant ( from Attachment, 3 ) A =

1.2 Half-life (from Attachment 3) T /2 i

=

2.0 Time and date of shutdown .

3.0 POWER CORRECTION FACTOR 3.1 Determine the power history using Attachment 2.

3.2 For steady-state power (except Cs-134), com-plete the appropriate section of 3.3. For transient power history (except Cs-134), com-plete the appropriate section of 3.4. For Cs-134, complete 3.5 3.3 STEADY STATE EXCEPT Cs-134 3.3.1 Half Life <1 day Power Correction Factor (PCF) =-

Steady state power percentage for prior 4 days 100

=

3.3.2 Half Life >l day Power Correction. Factor (PCF) =

- Steady state power percentage for prior 30 days 100

=

3.4 TRANSIENT EXCEPT Cs-134 3.4.1 Isotope with T 2 /2 >l year Power. Correction' Factor (PCF) = EFPD Total days of operation =

~3.4.2 Total period of operation > 4 x T 2/2' Power Correction Factor (PCF)=IifPi (1-e-Atj),-At*j) 100 =

ATTACHMENT l- Page 1 of.5

. y

, Procod. No. EDP-ZZ-00005 Rev. 0

. I where tj = operating period in days at power Pj where power does not vary more

.than !10 percent power from time average value (Pj )

Pj = percent power during operating period tj t'j = time between end of period j and

time of reactor shutdown in days. '

3.4.3 Remaining trasient cases Power Correction Factor (PCF) =

Ij[Pi(1-e- A t]. ),-At.j )

- 100(1-e-AIjtj) =

3.4 POWER CORRECTION FACTOR FOR CS-134 Power Correction Factor (from Attachment 6) =

(Use average power during entire period of operation from Attachment 2) 4.0 RCS ACTIVITY 4.1 Sample Data 4.1.1 Time and date of RCS sample 4.1.2 Time since shutdown t= (same units as A) 4.1.3 RCS volume (frem Attachment 4) V = ft3 4.1.4 RCS temperature TI = 'F.

4.1.5 RCS water density ratio (from Attachment 7) pl/pstp =

4.1.6 Sample result Cm = pCi/cc ,

4.1.7 Sample temperatur,e T2 =' 'F 4.1.8 Sample water density ratio (from Attachment 7)'

pZ/pstp =

4.2 Decay correction of sample to time of reactor shutdown 4.2.1 -Cc = CmeAt = pCi/cc 4.3 Temperature correction of sample ATTACIDENT 1 Page 2.of 5

~

. Procsd. No. EDP-ZZ-00005 Rev. 0

.4.3 Temperature correction of sample 4.3.1 09/pstjg C = Cc x p'/pstp = pCi/cc 4.4 RCS Activity A(RC) ,

4.4.1 A(RC) = V x pt/pstp x C x .02833 = Ci 5.0 CONTAINMENT SUMP ACTIVITY 5.1 Sample Data 5.1.1 Time and date of containment sump sample 5.1.2 Time since shutdown t= (same units as A) 5.1.3 Containment sump volume (from Attachment 5)

. . V= ft3 5.1.4 Containment sump temperature T1 = *F' 5.1.5 Containment sump water density ratio (from Attachment 7) p 1 /pstp = -

r 5.1.6 Sample result Cm = pCi/cc 5.1.7 Sample temperature T2 = 'T 5.1.8' Sample water density ratio (from Attachment 7) pz/pstp =

5.2 ' Decay . correction of sample to time of reactor shutdown 5.2.1' Ce = Cm x eat , gefjge ,

l

-5.3 Temperature correction of sample 5.3.1 _

pe/osto

'C s'Cc x p:/pstp.= pCi/cc

[

5.4: . Containment Sump Activity A(CS)

Ci.

5.4.1 A(CS) = V x pt/pstp x C x .02833 =

t- 6.O CONTAINENT ' ATMOSPERE ACTIVITY L 6.1 Sample Data I

+

, ATTACEMENT 1-:Page 3 of 5 a

t

. ** l l

l .

Proced. No. EDP-ZZ-00005 Rev. 0 l l

I 6.1.1 Time and date of containment atmosphere sample c 6.1.2 Time since shutdown t =

l (same units as A) l 6.1.3 Containment atmosphere temperature l T1 = *F 6.1.4 Containment atmosphere pressure F1 = psia I 6.1.5 Sample result Cm = pCi/cc 6.1.6 Sample temperature T2 = *F 6.1.7 Sample pressure P2 = psia 6.2 Decay correction of sample to time of reactor shutdown - -

6.2.1 Cc = Cme D= pCi/cc 6.3 Temperature and pressure correction of sample 6.3.1 P, x (T, + 460?

C = Cc x P2 x(Tt + 460) = pCi/cc 6.4 Containment Atmosphere Activity A(CA) 6.4.1 A(CA) = C x 7.075 x 104 = Ci 7.0 TOTAL ACTIVITY A 7.1 A = A(RC) + A(CS) + A(CA) = Ci 8.0 INVENTORY AVAILABLE FOR RELEASE 6.1 Uncorrected inventory (from Attachment 3)

Iu = Ci 8.2 Power Correction Factor (from section 3)

PCF = __

'6.3 Corrected' inventory Ic = PCF,x Iu = Ci-9.0 PERCENTAGE OF INVENTORY RELEASED A

9.1 Percentage of inventory released's x '100% = %

I e

ATTACHMENT 1 Page 4-of 5

Prdcad. No. EDP-ZZ-00005 Rev. 0

! 10.0 ACTIVITY RATIO i 10.1 If th'e isotope is a noble gas, complete 10.2.

1 If the isotope is an isotope of iodine, com-plete 10.3. Otherwise don't complete this section.

10.2 Noble gas ratio = A/A(Xe-133) =

10.2.1 Iodine ratio = A/A(I-131) =

4 ATTACHMENT 1 Page 5'of 5

. ,o

, Procad. No. EDP-ZZ-00005 Rev. 0 POWER HISTORY 1

1. 30-day power history Days Before Shutdown Average Power 1

2 3

4 5

6 -

7 8

9 10 11 12 -

13 14 -

15 16'

. 17 18 19 20 21 22 23 24 25 26 27 28 29

. 30

2. Total number of day's.of operation D=

3. EFPD =

4. Average power during entire period of operation

EFFD x 100%

D ATTACHMENT 2 Page.1 of 1

, .e .

Procad. No. EDP-ZZ-00005 Rev. O COPE DAMAGE T .\ INVENTORY STATE NUCLIDE

-l V.r-87 76m .00912m

~1 4.1(7) clad Failure Rb-88 18m .0385 m 6.1(7)

Xe-131m 12d .0578 d-l -1 6.5(5)

Xe-133 5.4d .128 d 1.9(S)

I-131 8d .0867 d-l 9.8(7) 1-132 2.3h .3014 h"1 1.5(8)

I-133 21h .033 h~3 2.0(8) 1-135 6.8h .102 h"I 1.9(8)

Cs-134 2y .3466y:f 1.3(7) Fuel Overheat Cs-137 30y . .0231y , 1.1(7)

Te-129 68.7= .010mj ,

3.4(7)

Te-132 77.7h .009 h 1.4(7)

Sr-89 53d .0131d-l 8.2(7) Fuel Melt Ba-140' 12.84 -.0541d-f 1.7(8) 1.8(8)

La-140 40h .0173h,~g La-142 90m .0077m 1.6(8)

Pr-144 17.3 .0401m-1 1.2(8) i

~ ATTACHMENT 3. Page 1 of.1.

Proced. No. EDP-ZZ-00005

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l i

l FISSION CORE EXIT !

CORE PRODUCT THERM 0COUFt.E CORE H, Monitor i DAMAGE RATIO READINGS (OT) UNCOVERY (t'01. % H ,)

i 1 j Na Clad Kr-87=0.022 ,,

None Negligible l l

Damane t-133=0.71 0-50: Kr-87=0.022 50%

Clad Dameae 750-1300 5-30 min 0-6.5 1-133=0.71 L 50-100% Kr-87=0.022 1300-1650 100%

6.5-13 Clad Dae. ate 1-133 0.71 5-30 min 0-50% Kr-87=0.22 51650- 50% ..

I i ov-rte perature I-133=2.1 45-75 min 50-100*. Kr-87=0.22 $1650 100% .

evertencerature I-133=2.1 45-75 min i 0-50% Kr-87=0.22 >1650 50%

l Fuel Melt I-133=2.1 $75 min 50-100% Kr-87=0.22 >1650

100% ..

Tuel Melt 1-133=2.1 575 min

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, Enclosure 2 The ranges and manufacturer's accuracies for the SNUPPS Post-Accident Sampling System (PASS) are provided below.

LIQUID SAMPLES ANALYSIS RANGE ACCURACY ~

-3 Gross Radioactivity, 10 Ci/cc to 10Ci/ce; Better than a factor of Gamma Spectrum Isotopic Analysis two Boron Content 0-6500 ppm t2% at 6500 ppm 15% at 500 ppm 140% at 50 ppm Chloride Content 0.1 to 20 ppm 210% to 1 ppm, 10.15 below 1 ppm Conductivity 0.1 to 1000 g os 21% of full scale pH 0-14 !0.1 Dissolved Oxygen 0-20 ppm 11% of full scale Dissolved Hydrogen 0-3000 cc/kg 25% of measured value CASEOUS SAMPLES ANALYSIS RANGE ACCURACY

~

5 Gross Radioactivity 10 p Ci/cc to 10 .C1/cc Better than a factor Gamma Spectrum Isotopic Analysis of two Hydrogen 0-10 Volume Percent 12.5% full scale Oxygen 0-30 wt % 21% full scale The SNUPPS PASS was tested during the factory acceptance tests with two chemical matrixes identified below.

PASS TEST HATRIXES Analysis Matrix #1 Matrix #2 Boron 1200 ppm 600 ppm Chloride 10 ppm 5_ ppm pH 4.9 4.4 Conductivity 33 unhos 17 umhos DO 8 ppm 8 ppm 2

DH 5 ppm _ 5 ppm 2

4'

/

Enclosure 2 Page Two The results of the factory acceptance tests meet the accuracy require-ments of NUREG 0737, Item II.B.3, with the exception of the boronmeter.

At. the time of the factory acceptance testing, the boronmeter was not fully calibrated and results outside the accuracy criterion resulted.

After final calibration, a test matrix was run in the field and the results of the boron analysis were within the accuracy criterion. Note, the boronmeter-employed in the SNUPPS PASS is a fission counter (neutron absorption) which will not be affected by chemical interferences.

In addition to the above matrix tests, the Orion Chloride Analyzer was extensively tested by Orion with the standard NRC test matrix with results meeting the accuracies listed _ in the above table. The remaining inline analyzers are off the shelf analyzers widely used in the chemical and power industries . Their wide acceptance coupled with successful completion of the above tests assures their ability to perform their particular analysis.

The design of the PASS is such that radiation sensitive components are removed from the sample panel where feasible and located in a lower radiation environment. Sample lines and holdup volumes were minimized to reduce the activity in the panel. Detailed dose calculations were performed for the sample panel using the high activity sources postulat-ed post accident. - As a result of ,these calculations worst case inte-grated doses were calculated for the electrical. components in the . sample panel. In all cases, the total dose to, any electrical component was below its_ damageL threshold. All wetted materials in the system were

.also' reviewed in light of the high activity and_ care was taken to select only. those. materials suitable for-this service.

As' a. result of the above mentioned testing and .-analysis the' SNUPPS PASS .

has demonstrated. its ability to ' adequately provide an accurate analysis-

= of _ liquid and -gaseous samples post accident.

' The SNUPP.S PASS was designed for_ use during both normal plant operation and following postulated accidents. The procedures for' sample analysis -

in a post-accident environment are identical to those used to obtainl a PASS sample during normal operation 'with th_e exception iof the need :for Ladditional emphasis- on health physics. requirements resulting from increased post-accident radiatio'n71evels. As discussed above, the PASS

instrument-- accuracies will _ be maintained within required limits in the post-accident-environment. Procedures for sampling;in the post-accident

< : environment, including obtaining grab samples,;are practiced in.conjunc-

~

tion with periodic . emergency planning drills. : The- PASS- sampling pro-

~

.cedures' were evaluated and found acceptable during' the' March 21, 1984:

emergency planning drill at Callaway Plant. 0perabil_ity?of the PASS is

~

assured - b'y' performance :of - system operational: checks ' and; system - func-

-tional checks lonRa periodic basis. The; operational checks.;will verify 1the . ability to analyze . routine samples. LThe functional checks' will'

? verify the. abilityjto analyze known sample concentrations.. ,

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