ATTACHMENT·A DRESDEN STATION UNITS 2 and 3 PROCEDURE … · e lit where: l· 1 = t 112 = half life of fission product i t = line from reactor shutdown to sample time t 112 and t

. ..

~ .....

ATTACHMENT·A

DRESDEN STATION UNITS 2 and 3

PROCEDURE FOR ESTIMATING CORE DAMAGE

0023K

850 5290728 850506 . I ... · <.PDR : ADOCK 05000237.

'I F PDR .. ·. 0

\

.-,,

l ' 19 i l. i-

.. : i ;

,.

--

A .

8.

PURPOSE

~..::.·---~~--c·--·--.. ··-·---•·····-_:_, __________ : - '---·- ··--·· ·--·-"'·-·-----····-···--•• ····-·· '··\

. ·. e ..

DETERMINATION OF THE EXTENT OF CORE DAMAGE UNDER ACCIDENT CONDITION

EPIP 300-18 Revision 1 APRIL 1985

The purpose of this procedure is to determine the extent of core damage under accident conditions.

REFERENCES

Procedures for the Determination of the Extent of Core Damage Under Accident Conditions, C.C. Lin, General Electric NED0-22215.

C. PREREQUISITES

1. Post-accident High Radiation Sampling System (HRSS) or other suitable means of obtaining the necessary sampler.

2. Gamma ray spectrometry system.

D. PRECAUTIONS

E.

0059E 0125b

Samples under accident conditions will be highly radioactive, proper radiation protection practices are to be followed when obtaining, handling, transporting or diluting highly radioactive samples.

LIMITATIONS AND ACTIONS

1. Measurements of Cesium (Cs)-137 and Krypton (Kr)-85 activities ·may be difficult to measure until most of the shorter-lived isotopes have decayed.

2. It is recommended that both the water and gas phase samples be measured in order to reduce the uncertainty in core damage estimations.

3. If water sample results show unusually high concentrations of some less volatile isotopes such as Strontium (Sr)-92, Barium (Ba)-140, Lanthanum (La)-140, and Ruthenium (Ru)-103 some

· degree of fuel melting may be inferred.

4. The ratio of isotopes released from either the fuel gap or the molten fuel are significantly different as shown in Table 1, thus the source (fuel or gap) of release may be identified with the use of Table 1.

5. The fission product inventories in the core are calculated based on three years (1095 days) of continuous operation at 3651 MW, or 1023 of rated power for the reference plant. These parameters were used to formulate Figures 1 through 4.

APPROVED

1 of 17 APR -,,n '85

D.O.S.R. I

----·--·- -.. -----···-~---·-. --~---~-··----·----,~---~--.'-:-_--,.......-:--··---···-·----:-----·-----"'."~---·----..-. ~------·--·-- -···-· -- .. --·-It

---··---· ..... ___ .--·· ....... ·---· ------·- -~ ---··~ :----··--- _ __,_ ___ -----·· -···---~--.:.·-"------.:~_.:__. ________ ~..:...--.:...,,__ -·'-···-·--- ...... '-··--·~· - - .. ·-- ·- '" .... _;.,_ -i ·r ..•

,.·.1 ! . :\ l j ··:-..:

6.

EPIP 300-18 Revision 1

If the concentration of a fission product in reactor water or drywell (corrected by decay to the time of reactor shutdown),· is measured to be higher than the baseline concentration shown in the lower right hand corner of Figures 1 through 4, then the extent of fuel of cladding damage can be determined from the curves in Figures 1 through 4 based on Iodine (I)-131, Cs-137, Xenon (Xe)-133, and Kr-85.

7 . Sampling.

a. For gas sampling, the recommended sampling locations are as follows:

b.

c.

Event Type

(1) Non-Breaks (e.g., MSIV Closure)

(2) Small Breaks

(3) . Large Breaks (liquid or steam) in Containment

(4) .·Large Breaks outside containment

Sample Location

Suppression Pool Atmosphere

Drywell (before depress.) Suppression Pool Atmosphere (after depress.)

Drywell

Suppression Pool Atmosphere

For liquid sampling, the optimum sample point for all events is the jet pumps as long as there is sufficient reactor pressure to provide a sample from that location. If there is not sufficient reactor pressure to allow a -sample to be taken from the jet pumps, then the sample should be taken from the sample point in the LPCI system.

In order to ensure a representative liquid sample from the jet pumps at low (< l~) power conditions for small break or non-break events, the reactor water level should be raised to the level of the moisture separators. This will fully flood the .moisture separators and will provide a thermally induced recirculation flow path for mixing.

F. PROCEDURE

0059E 0125b

1. . If not already performed, collect the required samples using the appropriate Dresden Sample Building Procedures (DSBPs).

2; Analyze the samples on a gamma ray spectrometry system using station procedures. Be sure to decay correct sample results to

. the proper sample time.

2 of 17

APPROVED

APR :SO '85

D. O.S. R.

-- ·---- .. ·-- --· --··-·-··--··· - -·--·- ·····-----··--:--·-··-·····-~-:--------~ .. ·--··-:--··· -···--········· --· .. -. ·-·-·-y-·--·- ··-· '

. . -:--;. ·>.·1-;."··ot_~, +.- .•.• .,_· (,;;.~;.,L ·,: ·- .... ,:_)~.: ·· -.~:. --~ ~-~-~~:%.:,-.,_,:;t._./-.,~;;.:.·,.; .:: .::;p,: ,;,~._:;,;.lb ;-,.:,.:; 1'.fl.!· ~:..:. --~~·;:.::~~,:~tf;~;,:~.;, ~:.-~::~ ·(>: .. -.c_._.. _ _:.-'--~,~-~_,,_:-~~;~~-'!...'-:_,,;.-'-'..:::_,-..:_.,,_,_d~·c&•:;.;,;.";c, .. , .. co.''. ... ~;.:,.:.,_ ... ·.

,.,!. ' ~ .

3.

EPIP 300-18 Revision 1

From the data obtained in Step F.2., pick out the results for the desired fission product. Identify the fission product as:

a. CWi for a fission product from a liquid sample.

b. Cgi - for a fission product from a gas sample.

In case the fission product .; concentrations are measured

separately for the reactor water and suppression pool water or the drywell gas and the torus gas, the measured concentrations (cone.) CWi or Cgi would be averaged from the separate measurements:

CWi = (cone. in Rx water) (Rx water mass) + (cone. in pool) (pool water mass) Reactqr water mass + pool water mass

Cgi = (cone. in drywell) (drywall gas vol) + (cone. in torus) (torus gas vol) drywell gas volume + torus gas volume

0059E 0125b

4. Decay correct sample results to the time of reactor shutdo'wn.

e lit where: l· = 1

t 112 = half life of fission product i

t = line from reactor shutdown to sample time

t 112 and t must be in the same units of time

5. Correct the measured gaseous activity concentrations for temperature and pressure differences in the sample vial and the containment (drywall/torus) gas phase.

The following correction. for the measured concentration is needed if the temperature and pressure in the

.sample vial (T1 , P1) are different

.from that in the containment (T2 , P2).

Cgi = Cgi(vial) x P2T1

P1T2

3 of 17

APPROVED

APR )() '85

D.O.S.R.

••••·•-·• •••·-·• ··- • •••.,•-••-••" ._,, ··-·--·-"••••---·••·------··•-•••-----~·----·-A--·-••-•·--•

·; ,,

I =j

. 0059E 0125b

•• EPIP 300-18

Revision 1

6. Calculate the inventory correction factor (Fri>· See the attached Appendix A for example of the calculation.

7.

Fri = Inventory in reference plant Inventory in operating plant

where: li =decay constant of isotope.i

= steady reactor power operated in period j CMWt> it

Tj 0 = duration of operating period j (day)

Tj ·. = time between the end of operating period j and time of the last reactor shutdown (day)

In.each period, the variation of steady power should be limited to ± 2M..

Calculate the plant parameter correction factors (Fw, Fg). · Refer to Table 2.

Operating plant coolant [grams (g)J

Operating plant containment gas volume (cc) Fg = Reference plant containment gas volume (4 x lo10cc)

8~ Calculate the reference plant equivalent concentration.

l t Cwi = (Cwi) (e i ) (Fii) (Fw)

l t Cg i = ( Cg i ) ( e i ) ( F Ii ) ( Fw)

Cwi from Step F.3.

l t . e i from Step F.4.

Fri from Step F.6.

Fw from Step F.7.

Cgi from Step F.5.

4 of 17

APPROVED

APR 10 '85

D.O.S. R •

·----- -- -·· - .- .... --.. ----~- -· -··· ·- ··---·--··-· --~ ... -----·------·----·---------~-----------------------

;. ,, l .I

.1 .. ] .

'.! ., ! ·.! i le··.· :j

·1 ·.J ' j

·1 . 'J.

I

i.

• •••• EPIP 300-18

Revision 1

9. · Use Figures 1 through 4 to estimate the extent of fuel or cladding damage.

10.. Determination of Clad Damage from Hydrogen Monitor Reading .

11.

0059E 0125b

a. Obtain containment hydrogen monitor reading, [HJ, in~- (or use results from HRSS Samples if available).

· b. Using the appropriate curve in Figure 5, determine the metal-water reactor for the reference plant, KWref at [HJ.

c. . The metal-water reaction for the actual in-plant conditions (KW) is determined from the following equation:

'I.MW = .<KWref> c500) N.

___ v'---> 350,000

MK I·

where N = number of bundles = 724 for Dresden 2 or 3

V = total containment free volume = 2. 75 x 105 f.t. 3

Determination of Clad Damage from Containment Radiation Monitor Reading.

The procedure for determination of fraction of fuel inventory released to the containment is as follows:

. a.

b.

c ..

d.

Obtain containment radiation monitor read"ing, (RJ in Rem/hour.-

Determine elapsed time from plant shutdown to the containment radiation monitor reading [tJ in hours.

Using Figure 6, determine the fuel inventory release for the reference plant [IJref in ~.

Determine the inventory release to the containment [IJ using the following formula:

. - 1670 v [IJ = (I1ref (----p--) (237,450> (6/D)

where: . P = reactor power ·level, MWth = 2527 HWth

V = total containment free volume = 2.75 x 105 ft.3

D = distance of detector from reactor biological ·. shield wall = 17 feet.

5 of 17

APPROVED

~PR 30 '85

O.O.S. R . . --·---·--- -··-·- ... -·----· -- ---·-·---------· ---·----·-.

. -l

·. 12.

0059! 0125b .

EPIP 300-18 Revision 1

Application of other significant parameters to core damage estimate.

a. Steps F.1. through 9. show how to determine an estimate of· core damage based on radionuclide measurements .. Based on these steps, an initial assessment of core damage is made. Based on a clarification provided by the NRC, that

·assessment would appear in a matrix as follows:

Degree of Minor Intermediate Major Degradation ( <10'1.) (10'1. - 50'1.) (>50'1.)

No·fuel damage 1 1 1 Cladding Failure 2 3 4 Fuel Overheat 5 6 7 Fuel Melt 8 9 10

b. As reconunended by the NRC, there are four general classes ·of damage and three degrees of damage within each of the classes except for the "no fuel damage" class. Consequently, there are a total of ten (10) possible damage assessment categories. For example, Category 3 would be descriptive of the condition where between 10 and 50 percent of the fuel cladding has failed. Note that the conditions of more than one category could exist simultaneously. The objective of the final core damage assessment procedure is to narrow down, to the maximum extent possible, those categories which apply to the actual in-plant situation.

c. . The initial core damage assessment based on radionuclide

d.

. measurement will provide one or several candidate categories which most likely represent the actual in-plant condition. The other parameters should then be evaluated to corroborate and further refine the initial estimate.

For example, fission product measurement using PASS may indicate Category 4 core damage and, additionally, the potential for fuel overheat and fuel melt (i.e., Categories 5 through 10). Measurement of hydrogen in

.. containment and use of the hydrogen correlation could be used to verify that extensive clad damage had occurred.

·use of the containment radiation monitor reading would verify that a significant fission product release to the containment had occurred, further verifying the initial assessment.

Further analysis of the PASS samples for concentrations of Ba, Sr, La and Ru and consideration of the relative amounts of fission products released would indicate if any fuel melt had occurred. APPROVE[)

6 of 17 nPR .-.,n ·s~

0.0.S.R. - - -· -- --- ·--·- .. --- -- _.,.. ·- - -- .-- ·- ---·- -· ---· ~---. ~----· - --

., ·, ·' ·. :·.

i

i ~ j

1 I !

l l·

'· !

l j\

I!

,I

·l -~ ~-r ··.I _,

A~--~. j :i .~i J. -,

-.i

' '-

G.

EPIP 300-18 Revision 1

e. The flow diagram in Figure 7 indicates how the analysis of the other sign_ificant parameters relates to the estimation of core damage based on radionuclide measurements.

CHECKLISTS

None.

H. TECHNICAL SPECIFICATIONS REFERENCES

0059E 0125b

None.

7 of 17

APPROVED

~PR 1n '85

D.O.S. R.

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

J' :j ·:~

s:.;

: ~ I

·.~.TABLE l

EPIP 300-18 Revision l

. RATIOS OF ISOTOPES IN CORE INVENTORY AND FUEL GAP

Activity Ratio* in Activity Ratio* in Isotope Half-Life Core Inventorx Fuel Gap

Kr-87 76.3 minutes 0.233 0.0234 Kr-88 2.84 hours 0.33 . 0.0495

· · . Kr-85 4.48 hours 0.122 .. 0.023 Xe-133 5.25 days 1.0* .1.0*

I-134 . 52.6 minutes 2.3 0.155 I~l32 2.3 hours 1.46 0.127 I-135 6.61 hours 1.97 0.364 I-133 20.8 hours 2.09 0.685

.. I-131 8.04 days 1.0* 1.0*

*Ratio = noble gas isotope concentration Xe-133 concentration

for noble gases

Plant

= Iodine isotope concentration I-131 concentration

for iodines

·.TABLE 2.

PLANt·PARAMETERS

Primar1 Coolant* ·.Reactor Reactor

Type/ Rated Water Suppression Containment Power Mass Pool Water

Design . (MWt) ~108 g2 ~109 g2 Dresden 2 & 3 BWR 3/I 2527 2.61 3.18

Containment Gas* Drywell Torus/ Gas Containment

Volume Gas Volume 9

~10 cc2 9

~10 cc) 4.48 3.30

*Total Primary Coolant Mass = Reactor Water + Suppression Pool Water

.Total Containment Gas Volume = Drywell Gas + Torus (or Primary Containment in Mark III gas

0059E 0125b

8 of 17

APPROVED

APR ?0'85

0.0.S.R.

··--... -.- .. -- ---·----·-.. ·- . ... ~.------- _.,._ ---··--.... -------~---~------:---'·~__,.. .. -- ... ------..-- ,....~ - _,.. __ ---·---· ., _______ -- -- - --- --·--

·' . .! I I

- ~

.I I

:! l ·Jj

. . . . ~: ... · ·.; .... . - ~ i .. t:·· . , . . . ·. ' , .. · .. . --·:.

.. ~:

·a:·· .. -.,, ___ ... ;, -... -: .

·-r:··,··

· EPIP 300-18 ' Revision 1 ....... ;.·:· -·

'( . \ .. ~,~--- "'-~---·~.~~~'.)105--------------------------.......,---------~ ... ·._ i.·e-'\ .. r··

I f.

10~

.... -u. ..· . .J . _,3 ta:: .,_ z < ~ 0 Q u. ~· < ! er a-Z''

. -. z: 0 ~ < I: . l>­z "' u· z Q u ··-"' -· ~

·. 10

1.a

.•:.·

/ /"

/· /

, .. _ · ..

/ '/

/ _, /

/ /

FUEL MEL rnowtt-.~· .

CUCCING ~AILURE

. ~UPPER REL.EASE l.IMIT .. /. . ..

/ . . BEST ESTIMATE . /" · . . . LOWER RELEASE l.IMIT ,

NORMAL SHUTCOWN ·. CONCENTRATION

IN REACTOR WATER

UPPER UMIT: 29.0 loLCi/g · NOMINAL;. . 0.7 ~/g

. 0.1._ __ .... _.._..~ ........... --..... ____ .................. .._ ___ ...__.__._ ............. __________ .._. ......... __ ..._ ____ ....., ....... ,,.

0.1 1.a. . 1CJ; 100..

,.,.. ..... --------~Ct.AOCING FAILURE-------~

1.0 . 10. 100

'"'" ••~--1' FUEL;.MELTCOWN ___ _,.._

, . ~ Rel.ationship Bec:-.;een I-131 Conce.'lcracion in the P-:i:iary Coolax:ic. i (Reactor Yacer + Pool Yater) and the Extent" of Core D~ge in : Reference Plan~ APPROVED

,.--·-------,.-- ~--------:---------------·-- ---------- --· .. : •....... - . . i .. . . . '.

'··· ..... .

Figure 1 ·

· 9 of. 17

APR 30 '85 .

o.o.s4 R. . -···-······-:·----:---·-.. -·------..-·.:---:----. ------·~-·

.·''

· .. \

·. _.:i

. --~·

·1 l i -

r ' ' f ·-L

i l

' ..... ·-----·---~_:·.

e.

u .a-·· .... ,~ z < ---~

0 ·o ---·-., u. -i~ '< l~ if: \~.- 10 iz. ·_ '0

;~ :a:.

_, .... _

z. ' '"' u

z. 8 :::;. -ict

. !· \

·1 I I.

'i i.

·:}. ___ __.,

1.0

0.1 - - ·- -· --- --..--=-\ .

r ·\·

/ /'

"-, r

r

, , r

r , /

"I' " . , /" . , . . .

-~~ CLAOOING FAl~UAE , "'-~UPPER REL.EASE' LIMIT

,'· - .· ···-· ~BESTESTIMATE ,~ • LOWER REL.EASE LIMIT•

Revi9ion. 1

. ...-

NORMAL SHUTCOWN­CONCfNTRA TlON IN REACTOR WATER

UPPER' LIMIT: NOMINAL..

Q.l. IA0/9-0.a::t~/9

t ..... ---------- 4J(, CLAOOING FAILURE------.••"'41

'~ 1a· ·1~

-1 ..... __ ---"FUEL MEL.TOOWN ---··I . - -- . ·--:··-·.-----·---:--: -,,_;

AppRQVED Relat:ionship 3et:lleen· Cs-137 Concentration in the Primary Coolant: • ·\ ' · . .c (Reactor Water + Pool. iJa~~;-)__ __ ~i; __ ;l:ie _ !=xtenc of Core Damage- in.

· · · -· l : Reference· Plane -·------------ --- - ·

APR 30 '8_5

o.o.s.R. ' ['·.·. Figure 2

10 of 17

--------··-·-··------.:..-~---------·---:--r----'I""-·-.:~-----

J l

' -._]

,,j' :.

0.1

·UPPER RELEASE L.IMll ~ ~

· .. · BESTESTIMA-ra- ~ ~ ~ /: . - /

. .. LOWER RELEASE L.lMlT'. ~ / . . :_,.' ... . /

/ ~

NORMAL. OPERATING CONCENTAA TION IN ORYWEl.L

UP!JER LIMIT: NOMINAL.:

10 ·4 ,.Ci/ci:

10 ~,.cue:

10-~ .... __ ....... _,_.u..i..i.;u..._....i.._. ....................... __ .................................... ~ ..... --........... --.i..1.1. __ __..._.._..,...i..i,......., a.1 1".0 10 100

-! ' "-CL.ADD.ING. FAIL.URE

10 100 . . •··I""' • ..,_ __ _,__

-·----·-,-· -···--··· -------.._

APR 30 '85

o.o.s.R.

: i ·1

. 1Jl..

I- '!I. FUEi. MELTOOWN -1

Rela~ionship !eeo.:een Xa-133 Concencracion in the Concainmenc Gas (Deywell + To_~-~~~L--~~--~-h~--~~-;~;;_t:. ~(_CQre _Damage in. Reference Plane .~ · · · ·

. .. ~ ,. - ,. '. -· .

Figure 3

· · .11 of 17 ·

. - ~-

.i

'. . ~:~

I I :

-, : . ·.

..

j· ! .

. 1.

'I . i ·.,· 1

. i i I I

!

10-3'

·. ·;

. -~:. . : · .. -....... "'~ ..

. -:: .:.:_..· .. ·

FUEL MEL TOOWN -·

UPPER" RELEASE- L.IMIT

CLAOOING FAILURE

UPPER' RELEASE LIMIT

8ESTESTIMATE

NORMAL OPERATION CCNCENTRA TION IN ORYWELL

. ~·-

UPf'ER LIMIT: . 4 11. 10-S ,.c;Jcr: ·.· ' NOMINAL. 4lL 10-0 ,.a/=.

...... .·

I""··-------~ CLACOING. FAILURE"-------*

APR "30 '85 .

D.O.S.R.

. 1.0. 10 100

. . .1.,. •----'{,FUEL. MEL.TCOWN-----,.•~1

. :.~. ··.Figure 4· ' . ,,

12. of 17 . . . '("' . ..;;.·~

· .. :·

... ·-

. ' .. . •, ,.

. ,,:·

% H2 Mark III:. L36E6 Ft3/748 Fuel Bundles

•

I

'

.. ...

% H2·Mark .I/II: 3~6E5.Ft3J500 Fuel Bundles

8 ... • ' :ii-· - _s::_. __ J!~'--,-~ ·---~-··.:-~-:-·--::----···-·-:::··----. -· --....-···---~ ........ ,,.-·

-~:-·;;t·. -~ !.- ···: .. ~ '." -·;.,·

:!. Figure 5

13 of 17

... ·· .. ~ .

.,·.·

· .. !·

-------,,-;-.... --...,.·-·-------~-.,_.--,.,-~

. ... . - -- 911111! _____ _

·-·.,

·,.::

•

i

8

0 ra

0 ...

R

0 -0

0

.. .. .. 4' (9 ... c :1

.Z

z 0

~ c ... cc cc ... ... c ~ I ~ c ... ... ~ -

= ""4 0 ""4 .... -., ""4. (J

Cll .x 41 w g:: = ::c w

41 ... ., 0 .!!

IW - I = .... 0 Cll .... ., ... 41· as ::c ... ., .... c 0

~ C: c 0 0 .... c.J: ., .

(J

= = 41 = coc. 0 "" at ~

~=

.APPROVED

·. APR.3n '85

D. 0. S.R .

• .

. ,'. ,'.;

.. . L

,/ cj l

·.'

: ·-. -

: ..... --. •,

' . ··: .··

. •· .. ·

e .'.

FIGURE 6

.EPIP 300-18 Revision 1

PERCENT OF FUEL INVENTORY AIRBORNE IN THE CONTAINMENT .... -· . . . . : .. ·- :-;- ~- .

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

.. <.

;_. \·: .

. -~ . : ·' . ' ,·. \- ·· ..

t. Fuel Inventory

··Released

100.0 . 50.0

10.0

.3.0

APPROVED 1.0 o·.1

APR 1n '85

o.o~s. R. e

0.01

lo-3 lo-4

5 x lo-6 10-6

0059E · 0125b

lDO\ ~l lnYWlS~cry • 100• ~~l• ~•es ·.· .· -. · • 2!' :=!:le

• l\ p&r.i:ul&U S

..... <· . . - .: :·. ~ : . ·, -.. ._ .. ~ .;:.·;~-:·~ .: .. ~ .)·· ... :_. :u· ".' . . •\ .-~ . .

Approximate Source and Damage Estimate

1003 TID-148~4, 1003 fuel damage, potential core melt. 503 TIO noble gases, THI source . 10!. TID; 1003 NRC gap activity, total clad failure, partial core uncovered. 3t. TIO, 1003 NRC WASH-1400 gap activity, major clad failure. · lt. TIO, 103 NRC gap, Max. 103 clad failure. O.lt. TIO, lt. NRC gap, 13 clad failure, local heating of_

. ~ 5-10 fuel assemblies. O.Olt. TIO, 0.13 NRC gap, clad failure of 3/4 fuel element (36 rods). 0.013 NRC gap, clad_ failure of a few rods. 1003 coolant release with spiking. 1003 coolant inventory release. Upper range of normal airborne noble gas activity in con tai nmen t .

14 of 17

·. ·,

-~,. :

-~ -..

·J

'. -~. ,. ~ ·.

, .. ·.

···::.·

.. ·· ~ •'/." .· .. ,,

.·. ' . ~-

·".

. . . . ·.

· .. ~ ; .

Determine Optimum Sample . Point : ,'

·:. ~ ... '

. ',, I l .• ~·

. -~· . . . :: -: ~-

~: : .

; : '•'

....

Hydrogen. t-----~ Analysis

· (Confirin)

Low No

Core Damagee---~-· Estimate .From PASS

·..,:_·,:

·'. ..... ~ '.

High' No

Hydrogen 1----"---:~ Analysis

(Confirm)

; ·: .. :~- .... · .. ' •.'

'·:: .. ...

·,',."

·.·:_· ....

. · .. --._.'.

•, ·· .. .. ·> .t

·. ~-- ".

'.:·

. •. ,·· ' •. '. -~. ·' . ,,r

. ; ·_,·. 1 •• ,, . . '

. . ·,

; .· .. :_·

. •'.'

. . . . ... .. .. .• . .•. .• : .... :.:~--. • ....,,..·.v: .. :-~ .. = .. ~':r' .... :1: ••• '('"..;,.~.~·~ .... !t._, ··;:~~-~~~~~7·ric:-·~?C .. ,. . .rsrtC

FIGURE 7 SEQUENCE OF ANALYSIS FOR . ESTIMATION OF CORE DAMAGE

Containment: i--· ..-Y.-.e=s__,_.. Rad ia ti on :

(Confirm)·

Water 1---"Y_e_s.__~ Level

(Confirm)

_,;

.. No · .. No

· .. _

No No

Containment ......,_Y=e.-s_·~ Radiation· Yes

Water Level

. \•'

. . :1 i

.....

· (Confirm)

· .. : ~ . .·:,·

. ·'

. (Confirm)

MAJOR CLAD DAMAGE JWEL OVERHEAT · FUEL _MELT.-

. . ··:·· .... · ....... :::-· . . . . · .... ·_, .· ......._ _________ _

• ' '.• 1 • • ~' ' ·.' • .·" I •• f 1' • •

· ...

eof 17

EPIP 300..:.18 Revision 1

APPROVED .

Yes APR 30 '85

D.O.S.R.

NORMAL OPERATION MINOR Cl.AD DAMAGE·

Yes

· .. ,

Analysis For Ba,· Sr, La, Ru

·._ · .... •• '· > •••

. Determination Yes .r Of Fission

i ·· 1 ...

,.··. ......

Product Ratios

No

CLAD DAMAGE POSSIBLE FUEL OVERHEAT NO CORE MELT ..

.·· ... ·

.. ; .

......

J

I ~ I

J ~ i

I I I

I I

. I

I r

I ! .

.·- . j· ·1-

<I . I

['

I i j.

t: 1 .... . -· ,. :..... ... ·-.

J }'

y · ..

'·'" -, ·-

:: ' .. . .. -~

APPENDIX A

SAMPLE CALCULATION OF FISSION PRODUCT ·INVENTORY CORRECTION FACTOR

_.\. .. . . -, ..

. ·.- '.

', ... ·.· . EPIP 300-18

Revision 1

Inventory of nuclide i in reference plant Fri = Inventory of nuclide i in operating plant

where:

Pj = ·steady reactor power. operated in period j (MWt>

Ai = decay constant of nuclide i (day-1)

Tj = duration of operating period j (day)

.'.','

= time between the end of operating period j and time of last reactor shutdown (day)

· .3651 =ave. operati_on power (in MWt) for the reference plant.

1095 = continuous operation_ time (in day) for the reference plant.

Assuming a Reactor has the following power operation history:

Operation Period

1A . 18

2A 28

3

·. 0059E . 0125b

h •'

Days Since Startup

1 60. 61 - 70

71 - 270 271 300

301 - 314

Operation Tj (day)

60

270

14

16 of 17

Time T o

l

254

44

0

APPROVED

. APR ~n ·as 0.0.S.&

Average.Power pj (MWt) ·

1000 0

2000 0

3000

---·· ________ ___, .. --~-·-- ..... ---,,,.--=--·--r·----~~-------------------~-...... ------·- ..

... ·.•::

. i

;I : ~ 'i

'., -~i ~:

;~ . .,.;

. n .~!

:'i

! I.

j

~- - ,

... _­--·--'

;;

" t. -

: . -

-._ - , . . . :: ·: :. . ' . =~ : .

':." .. -··-'·e··-.. ,_: . :,•

. - . ... . ..- . ·.

.. ., '

.' .- .·; . . . -.,_··.o·

. ·, · ... ·,.

. · .·

. . . . -- -1. For I -131 (l = 0.0862 day )

. ~' ·., .·

APPENDIX A (Cont'd)

- .. -. . ,~ .

·.·,,"·'

... ··

EPIP 300-:-18 Revision 1

:-·:· .. ·' ..

- FI(I-131) = ____ ··_·-·-_-· ____ ·-_-•--_. ___ 3~6-.5~1~Cl~-~e_-_0 _·0_8_6_2_<1_0_9_5_>~1 ____________ ___

1oooc 1-e-0.0862(60>ie-0.0862(254~+ 2000 c1_9-0.0862(200> 1

~·0.0862(44) + 3000[1-e-0.0862(14)Je-0.0862(0)

. = ___ . --· 3 ....... 6.-....51...__ __ = 1. 7 4 ~o + 0.0225 +.2103

5 .;.1 For Cs-137 <l = 6.29 x 10- day )

......

F1<cs-137 > = __ 3~6 ..... 5_1 __ c 1 ...... --·-~9_::.::_6_: 2_9_E_..:_5_n_o_9_5_> __ 1 ---1000c1_9-6. 29E-5 ( 60>ie.;.6.29E-5C254 >

'• .··, r" •:.:

0059E 0125b

· .. · .· ..

+2000[1-e-6.29E-5(200)Je-6.29E-5(44)

+ 3ciOO(l~e-6.29E-5(14)) 9-6.29E-5(0).

; ... ·

= ______ 2 __ 4 __ 3 .... ~.-.1_..6_____ = 7 • 77 3.74 + 24.93 + 2.64

- ' ....... -· ... _: ..

... ,: .

·. -,·,:··

17 of 11 {FINAL)

APPROVED

APR 30 '85

o.o.s. R.

. ; ·,,·

... --.

., .. _,:,.·-

.... _ ·, .

": , .

''.·.-'

.... .... '

·­I·'· .

. : .··-~

... · .. · ..

- ··.

. r'.

.. ~ .. · .... ·i·_ ·., ~-.·. l

.i­.:3

~1. :f -'l ;l

-.- ..... -. -. ,..,;·:·--::·----.--. - ·----:~·'':° ......... ~--·-. .... ·:-:-·--·---: .. ··----· ~- -· .... -------.. - ---·-.. ~.__ _____ ; ____ ~- __ ; .. _----·-~.;__~~--·---· ·--·-----· ---· -- ····--··--~