Text

Chemistry Procedure 78.000.15, Determination of Extent of Core Damage.
	ML20052E374
Person / Time
Site:	Fermi
Issue date:	04/29/1982
From:	Eberhardt R, Mcneil W; DETROIT EDISON CO.
To:	;
Shared Package
ML20052E370	List: ML20052E369; ML20052E374;
References
	78.000.15, TAC-44467, TAC-44468, NUDOCS 8205110065
	Download: ML20052E374 (20)
	v • d • e

78.000.15 e e ENRICO FERMI ATOMIC POWER PLANT UNIT NO. 2 Type: CHEMISTRY PROCEDURES - SPECIAL TESTS L W O R M K O 'N O ~Y Y

Title:

DETERMINATION OF EXTENT OF CORE DAMAGE RECORD OF APPROVAL AND CHANGES Prepared by Robert Eberhardt/W. McNeil April 23, 1982 Date

~

Submitted by [ e_ i b. M / 2 f /22

't Section Head Date Recommended by // ,/ 4 2Gr 'L .

USRO Cha an /

Dite Recommended by 4 F 2.

Supervisor, Date Operational Assurance p f 7 L Approved by (( .-

Superintendent - Nuclea Date Production Supervisor, Operational Supt. -

Revision OSR0 Assurance Nuc. Prod.

Recommended Date Recommended Date Approved Date No.

1 2

3 4

5 6

7 8

8205110065 820429 -

DR ADOCK 05000

.a

4 _ _

78.000.15

' Rev. O i.

I TABLE OF CO:iTENTS Page 1.0 Pu r po s e . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . I l

2.0 D i s c us s i o n . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . I I 3.0 Re f e r e n c e s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2 4.0 Equipment Re qu i r e d . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2 2

4.1 Apparatus..................................

4 . 2 Re a g e n t s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2 5.0 Precaut ions and Limit a tions . . . . . . . . . . . . . . . . . . . . 2 6.0 Prerequisities................................. 3 7.0 P r o c e d ur e . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 8.0 Ac ce pt anc e Cr it e r ia . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6 Enclosures Core inventory of Major Fission Products in a Reference Plant Ope rated at 3651 MWt f o r Th r e e Ye a r s . . . . . . . . . . . . . . . . Encl o sur e 1 Fission Product Concentrations in Reactor Water and Drywell Gas Space During Reactor Shutdown Under !brmal Conditions ... Enclosure 2 Ratios of Isotopes in Core Inv ento r y a nd Fuc i Ga p . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . Enc lo sur e 3 Plant Pa r a m e t e r s . . . . . . . . . . . . . . . . . . . . . . . . . . . En c l o su r e 4 Relationship Between 1-131 Concentration in the Primary Coolant (Reactor Water +

Pool Water) and the Ex te nt of Core Damage in Re f e rence Plant . . . . . . . . . . . . . . . . . . Enclo sure 5 Relationship Between Cs-137 Concentration in the Primary Coolant (Reactor Water +

Pool Water) and the Extent of Gore Damage in Re f e r enc e Plan t . . . . . . . . . . . . . . . . . . . . . . . . . Enclosure 6 l

Relationship. Between Xc-133 Concentration in the Containment Gas (Drywell + Torus

~

l

' Cas) and the Extent of Core Damaje in Re f e re nce Pl an t . . . . . . . . . . . . . . . . . . . . . . . . . . . . Enclo sur e 7 a

(

1

- - 78.000.15 l Rev. O TABLE OF CONTENTS Enclosures (continued)

Relationship Between Kr-85 Concentration in the Containnent Gas (Dr ywell + Torus Gas) and the Extent of Core Damage i n Re f e re nc e Pl an t . . . . . . . . . . . . . . . . . . . . . . . . . Enc lo su r e 8 Best-Estimate Fission Product Re l e a se Fr ac t io n s . . . . . . . . . . . . . . . . . . . . . . . . . . Enc lo su r e 9 Samples !-bst Re pre se nta tive of Core Canditions Daring An Accident For Estimat(on of Core Da m ag e . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . En c l o s u r e 10

b 78.000.15 Rev. 0 ,

Page 1 i

1.0. Purpose The purpose of this procedure is to determine the degree of reactor core damage from the measured fission product concentrations in either water or gas samples taken from the primary system under accident j-conditions. The procedure involves calculations of fission product ;

inventories in the core and the release of these inventories into the primary system under postulated loss-of-coolant accident (LOCA) ;

cond itions . The fuel gap fission products are assumed to be released

' upon the rupture of fuel cladding; the majority of fission products in the core will be released when the fuel is melted at higher ;

?

tempe ra tures .

i I

2.0 Discussion The estimation of core damage will be calculated by comparing the ;

measured concentrations of major fission prod uc ts in either das or :

liquid samples, af ter appropriate normalization , with reference plant data fron a SJR-6/238 aith a turk 3 containment. Fission produc t i inventories in the primary system were calculated based on postulated l loss of coolant accident (LOCA) conditions af ter three years (1095 days) of continuous operation at 3651 MWt. or 102% of rated p3 war by ,

i using a computer code developed at Los Alamos and adapted to the GE computer system. The inventories of major fission products in the ,

core at the time of reactor shutdown are given in Enclosure 1. l The pertinent reference and EF2 plant parancters are given below: i Reference Plant EF2

Rated Reactor Thermal Power 3579 MWt 3292 tJe Number of Fuel Bund les 748 Bundles 764 Bundles 4 .

Total Primary Coolant thss (Reactor Water plus Suppression i Pool water) 3.92 x 109 g 3. 51 x 10 9g I Total Containment and Dr ywell Gas Space Volune 4.0 x 1010 cc 8.35 x 10 9cc ,

Cas/ water samples taken from the Post Acc id ent Sampling system are .

analyzed for major fission product concentrations by gasma ray spectrometry. If the concentration of a fission product in reactor .

water or drywell, decay corrected to the time of reactor shutdown , is !

measured to be higher than the baseline concentration shown in 4 Enclosure 2, the extent of fuel or cladding daca;e can be determined directly f rom Enclosures 5-8 . based on isoto res I-131, Cs-137, 'xe-133, and Kr-85. Measurements of Cs-137 and Kr-85 are not very likely until the reactor has been shut down for longer- than a few weeks and most of ,

the shorter-lived isotopes have decayed.

I i

s e w y + -

~

. 3 t

' 78.000.15 Rev. 0 :

Page 2 = ,

P If the concentration falls into a range where the release of thefuel cannot be defini 6

fission product from the fuel gap or moltento determine the source of determined, additional data may be neededFor example , if less volatile fission fission product release. found to tuve products such as isoto pes of Sr , Ba , la , and Ra aresample as compared to unusually high concentrations in the water baseline reactor water concentrations, a fuel meltdown may beSr-92 (1.38 assum ed . The presence of 27hr identif y and measure from a (1.597:teV) will be relatively easy to gamma ray spectrun.

In addition to the longer-lived isotopes , some sample shorter-lived

. 'Ihe isotope ratios of isotopes significantly concentrations may be measured in thefuel gap or the molten fuel are released from either the source (fuel or gap) of dif ferent as shown in Enclosure 3, thus the release may be identified .

from Samples acquired for the estimation of core damage shall be takenwi locations that are'Ihis consistent will ensure the viability of results reported (Enclosure 10).

and provide the best estimation of core damage.

3.0 References 3.1 Lin , Chien C, Procedures for the D2 termination of Core Ihmage Conditions , Ceneral Electric Co ., 1982 Under Accident Ib st Acc ident Sampling 3.2 Chemistry Procedure tb . 78.000.14 3.3 Chemistry Procedure ib . 70. 000. 05, Operation of the Gemistr y ND6685 3.4 Chemist r y Proced ure Ib . 76.000.06, Operation of the @ccistr y ND680 4.0 Equipment Required _

4.1 Apparatus 4.1.1 Camma Spectroscopy system.

4.2 Reagents l

i ibne 5.0 Precautions and Limitations None

. . .. . -r 1 - r

78.000.15 Rev. O Page 3 6.0- Prerequisites ,

6.1 Accident condidions exist and a decision has been made to take a .

sample by the General Supervisor of Chemistry or designee.

6.2 Specific location and additional instructions for the acquisition of samples have been given to Operations and Chemistry consistent ,

with the break case and systen conditions as described in Enclosure 10.

i 7.0 Procedure 7.1 Estimation Procedure . .

7 .1.1 Ontain samples , consistent wi t h Ehclosure 10, from tne Po st Acc ident Sampling System. C

- 7.1. 2 Perform gasna spectroscony and determine the concentra-tion of fission products 1-131, Cs-137, Ne-133, and Kr-85.- ( Cwi in water or Agi in gas .)

i a

7.1. 3 Correct the measured fission products for decay to the i time of reactor shutdown.

7.1. 4 If the temperature and pressure of the gas sample vial

' are different f ron that in the containment , correct the '

measured gaseous activity concentration for temperature and pressure per Sec tion 7. 3. l 4

7.1. 5 Calculate the fission product inventory correction f actor Fit per Section 7.4.

I 4 7.1. 6 Calculate the plant parameter correction factors (F3 or I Fw) per Section 7. 5.

I 7.1.7 Calc Uf or Cj,ulate f for I-131, f theCs-137, normalized Xe-133,concentration

'and Kr-SS per , C ? or ,

l' Section 7. 6 i u

7.1. 8 Interpretation of C[tf or Ch{f :

1. If the normalized concentrations , CU{f orc $1, f ;

obtained in Sec tion 7.1.7 are higher thaa the [

baseline concentrations shown in Enclosure 2, the extent of fuel or cladding damage can be -

I determined directly f rom Enclosures 5-8. 3 t

I S

. e t

v w -v * , + ~v ems v y e-~s- - n.<~e.- p e ,, - , y

- - 78.000.15 .

Rev. O Page 4 (2) If the normalized concentrations fall into a range where release of the fission product from the fuel

% gap or the molten fuel cannot be definitely determined , the presence of Sr , Ba , la and Ru should be established . Fission products 27hr St-92 (1.385 !!cV) and 40hr 12-140 (1.597MeV) are relatively easy to identify and measure from a gamma ray spectrum and are indicative of fuel mel tdown . These results should be compared to baseline reactor water concentrations.

7.2 Identification of Pelease Source 7.2.1 Determine the concentrations of the following short-lived isotopes by gamma spectroscopy:

Kr-87 1-134 Kr-88 1-132 Kr-75m I-135 Xe-133 1-133 1-131 7.2.2 Correct the measured fission products to the time of reactor shutdown.

7.2.3 Calculate isotopic ratio's per Section 7.7.

7.2.4 Determine release source by comparing results obtained in Section 7.2.3 to ratio's supplied in Enclosure 3.

7.3 Temperature and pressure correction for gas sample vial.

C =C P T gi gi (vial) x 2 1 E T l 2 where Cgi (vial) = Samp.le vial isotopic concentration Cgi = Containment isotopic concentration ,

- (P,T) = Sample vial pressure and temperature 3 3 (P , T ) = Containment pressure and temperature 2 2 i

i I . t

. t'

78.000.15 Rev. O Page 5 '

7.4 L Fission Product Inventory Correc tion Factor F = Inventory in reference plant 11 utventor y in u4

-1095 )L i

= 3651 (1-e, s) _.

i

{jl Pj(1-e ~Ai'j ) e Ai jl where: Pj = steady reactor power operated in priod j (?Mt)*

Tj = duration of operating period j (days)*

time j Tj = time between the end of operating period j of reactor shutdown (days)*

4 For a particular short-lived isotope , i, a calculation for only a period of 6 half-lives of reactor operation time before reactor ;

shutdown should be accurate enough. It should be pointed out i that the computer _ calculation of core inventory takes into account the fuel burnup, plutonian fission and neutron capture reactions. The correction factor calculated from this equation may not be entirely accurate , but the error is insignificant in comparison to the uncertainties in the fission product release i

f ractions (Enclosure 9 ) and other assur.ptions .

7. 5 Plant Parameter Correction Factors 4

l Fy = EF2 coolant mass (c) _

reference plant coolant mass (3.92x107 g)

. Fg = EF2 gas volume (cc) reference plant containment gas vol. (4x101V cc)-

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 Cgt or Cgt would be averaded from the separate measurement s :

. C,1 = (Conc. in Rx water)(Rx water mass) + (Conc. in pool)(Pool water mass)

Reactor water mass + pool water mass l C gi = (Conc. in drywell)(Dryeell gas vol) + (Conc. in Torus)(Torus gas vol)

Drywell gas volume + Torus 3as volume

)

I

ln each period the variation of steady power should be limited to + 20%.

~

a -

y 'es- r - ,

i l

78.000.15 -I Rev. O Page 6 7.6 Calculation of !brmalized Concentration C,i and C81 it 5 Cwyf = Cwie xFuxF w Chf=Cgie g

it FuxFg where C f = concentration of isotope 1 in the reference plant reference plant coolant (Ci/g)

CMf = concentration of isotope i in the reference E plant containment gas (Ci/cc)

= measured concentration of isotope i in Cwi EF2 coolant at time, t (Ci/g)

C gi = measured concentration of isotope i in EF2 containment gas at time, t (Ci/ cc) t i = decay correction to the time of reactor shutdown ki = decay constant of isotope 1 (day)

> t = time between the reactor shutdown and the sample time (day)

Fit = inventory correction factor for isotope i F

E

= containment gas volume correction factor Fw = primary coolant mass correction factor 7.7 Calculation of isotopic ratios Noble gas ratio = Noble gas isotopic concentration Xe-133 Concentration lodine ratio = Iodine isotopic concentration 1-131 Concentration 8.0 Acceptance Criteria None

" " W d .D. 4 " p-: g e

78.000.15 i Rev. 0 +

CORE INVENTORY OF tlAJOR FISSIO!! PRODUCTS IN A REFERENCE PLAhT OPERATED AT 3651.Wt FOR THREE YEARS t

MAJOR GA!!'!A RAY ENERGY INVENTORY (INTENSITY)

CHEMICAL CROUP ISOTOPE

11ALF-LIFE 106 Ci kev (7/d)

Noble gases Kr-85m 4.48h 24. 6 151(0.755)

Kr-85 10.72y 1.1 514(0.0043)

Kr-87 76. m 47.1 403(3.494 )

Kr-88 2. 84h 66.8 196(0.203),1530(O.109) s Xe-133 5.25d 202. 81(0. 371 )

Xe-135 9.09h 26.1 250(0.906)

Halogens I-131 8.04d 96. 364(O.824) 1-132 2.29h 140 668(0.99),773(0.702) 1-133 20.8 h 201 530(0.87) 1-134 52.6 m 221 847(0.954),884(9.653) 1132(0.231),1260(0.293) 4 1-135 6.59h 189 Alkali Metals Cs-134 2.06y 19.6 605(0.98),796(0.88)

Cs-137 30.17y 12.1 662(0.85) .

Cs-138 32.2 n 2990.** 463(0.267),1436(0.75)

Tellurium Group Te- 132 78. h 138 228(0.88) ,

2 Noble Metals tb-99 66.02h 183 740(0.138)

Ru-103 39.4 d 155 497 (0.9)

Alkaline Earths St-91 9.52h 115 750(0.24)

, Sr-92 2.71h 123 1385(0.9)

Ba-140 12.8 d 173 537(0.238)

Rare Earths Y-92 58. 6 d 118 934(0.137)

La-140 40. 2 n 184 487(0.453),1597(0.953)

Ce-141 32. 5 d 161 145(0.49) ,

Ce-144 284.4 d 129 134(0.108) -

i Refractories Zr-95 46. d 161 724(0.435),757(0.543)

Zr-97 16.8 h 166- 743(0.933).

0nly the representative isotopes Wich have relatively large inventory and considered to be easy to measure are listed hele.

- 1 hr after shutdown ,

Enclosure 1

' Page 1 of 1 i

__ _ -y- = m % -

w-

78.000.15 * '

4 R2v. 0 1

FISS10N PRODUCT CONCENTRATIONS IN REACTOR WATER AND DRYJELL CAS SPACE DURING REACTOR SHlTIDOWN UNDER NOM 1AL CONDITIONS DRY 4 ELL GAS ISOTOPE REACTOR WATER, uC1/g uCi/cc UPPER LIMl_T, . NOMINAL UPPER LIMIT NOMINAL I-131 29 0.7 -- -

Cs-137 0.3* 0.03** -- --

Xe-133 -- -- 10-4* 10-5**

Kr-85 -- --- 4x10-5* 4xto-6

i

0bserved experimentally, .in an operating BWR-3 with .11K I containment ,

l data obtained from GE unpublished document, DRF 268-DEV-0009.

- Assuming 10% of the upper limit values.

Enclosure 2 Page 1 of 1 ,

a s 9n, wham" A9 .gg _

A -

Ma - - _ , , _ , _.

_ _ _ [ ._. ,, , , , , ._

78.000.15 Rev. 0 RATIOS OF ISOTOPES IN CORE INTENTORY A*.B FUEL GAP ACTIVITY RATIO

IN ACTIVITY RATIO

IN liALF-LIFE CORE INVENTORY FUEL GAP ISOTOPE 0.233 0.0234 Kr-87 76 m Kr-88 2. 84h 0.33 0.0495 4.48h 0.122 0.023 Kr-85m Xe-133 5.25d 1.0* 1.0*

I-134 52.6 a 2.3 0.155 I-132 2.28h 1.46 0.127 I-135 6. 59h 1. 97 0.364 20.8 h 2.09 O.685 I-133 1-131 8.04d 1.0* 1.0*

s

Ratio = noble gas isotope concentration for noble gases Xe-133 concentration

,. Iodine isotope concentration for iodines 1-J31 concentration Enclosure 3 Page 1 of 1

- - _ ~ . _.

l 78.000.15-j Rev . 0 t

. )

i.

PLAhT PARAMETERS t

PRIMARY COOLANT

CONTAINtENT GAS

TORUS /

RATED REACTOR SUPPRESSION DRWELL CONTAI?BIENT POWER WATER MASS POOL WATER CAS VOL. GAS VOLIDIE

- REACTOR TYPE / CON-PLANT- TAIN!ENT DESIGN ( MWt) (108g) (10 9g) (109ec) (109cc)

BWR 4 3292 2.77 3.23 4.64 3.71 EF2 MKI l

T

, I

Total Primary Coolant Mass = Reactor Water + Suppression Pool Water s J

Total Containment Gas Voltene = Drywell Gas + Torus (or Primary Containment in MKIII) f gas i

4 Enclosure 4 Page 1 of 1

4. ._ m_... - ,. .

Rev. 0 Reistionship Between 1-131 Concentration in the Pricary Coolant (Reactor

. Water + Pool Water) and the Extent of Core Dacage in Reference Plant g . , . . . . . .

10 ,. . _ -

-- 1 I - _ .

Fuel Meltdown t _ _ _ . . .

1 . _ _ . _ _ _ _ . _ . . ,

-- T ' -

_ _. . . . Upper Release Limit . .

=_ _ ;

i -

.K____..., ....

g m .. . .. '

. Best Estimate '-

. y s

% s'/_

_. : ._u.; .p. .

j:: Lower Release Limit / . I.

102

_ Kqf.* ;r./ . ._. .-

\

, ,9%. . ._, . . ,

. ,- ,7 ,'

. _ /< ...,.s. -.

-i --I _ _ _ .

.._ _. ./.,'.. . .- _ .... .

l _ _ .s .< : ...,. / ' . -- . . ,

. . . . _ ,a _., s.. . . . . . . . ._.._ . .. .

w I o i e 3: _

.s. )g # .

j t.

a. .i

.. =-_ = = ..

s., .i , -- / -

1

._ Jy_ ; . . . . ,

/ /

.c -. _._...._.,: _.. . . .

e . .

3 ,

. _. , . _ . ,jf_ ,c.

w r

'.N ... . ,/ _ ...___..

y f. .a t 7 10 ;,

i. ,

.s - . .

x ,

a

, s ,

, , _ . . . _ . _ _ _ _ _ .. _ . . . . . ... 1,

..c. -

,- p ,

c ,-. . . . . .

- _ .4 ._ __ _. . , ,...___ f. -

s

..c, a

m , ,

i s....w i

s. .

/ . , N '/- ,

.q,,.' i .

e .

c . , . .

, . ; Cladding Failure --------4 8

c,10 c 1

{ .

-. ,/ g -- . .

=-

[

# Upper Release Limi t - ._-- ---.

u ,..- ,r

,- , x _

2 ! s x Best Esticate

. 1

7 i / . , s

- i / N

..{

. a,

, j_ , _ _ _ _,... < Lower Release Limit _ _

~

.j ' .

. _ _ . ._ . _ . . . _ . . .._._e

.t .i . /4 . 1 . ,f' 6 i ! .

=- I. . .

a f. . . - - - 1 .4 -~ '

_. ..._ _ J J :P.-j-5f. : i . p q .7 . ,'. . i

. j !

1. 0- -

1 z r . , _.

I y ./.

/i Norral Shutdown Conc. in Reactor Water --

l

f. s - ) ; ,1 , s .

l

%- - i a_ 2 ,' Upper Limit : 29.0 uti/g _ . . . . . .

( > ~ww- -

w i

3,',./

t. _ ',

s h.ominal : 0.7 uti/g -- - - i 1 e . , - .. . ar..1 - - . ..- . . , . . . , , , .

m :_A .: 1 r= n 2.e : : - : : -t u - 1 m . i .i' s = ! :+ ! -: ...- .:

9

_ . .i.?_d 2 -

_.i1. .E. ,M. _U. r. i_. s :. t. . .i .9. 31.i.i .._~_

, . . _._M_i. .i. i. H. . H. . .'

.! .i + J 11 . .i i. i t

I O' 1 Q.1 -

1.0 10 100 h % Cladding Failure 4 1.0 10 100 e 1 Fuel Heltdown -

Enclosure 5 1

Page 1 of.1

^

= _ - -._---__.-w ~ n , _

, , , -- - 1

.._._._.-.m_. ~_ _ _ _ _ - _ _... ..a .__.._ ___..~ .._. . n.. .. . _ L . _ p - r-"

\.

78.000.15 Rev. 0 Relationship Between Xe-133 Concentration in the Containment Gas (Drywell s 7 4 + Torus Cas) and the Extent of Core Danage in Ref erence Plant 10-1 _ _ . ..,

4 Fuel Meltdown

- .s ~' -

t i _ _ .....__ _ - _ Upper Release Limit s - r/ --

i

.f

! - ---- Bes t Es tima te - -

7' - - - - :i I : 4 Lower Release Limit '

-- - . l 1 j . _. ,i '. .. .. . . ... . ._ .f .j -

i. .. . . .

1 , !

39, _.

y_,y..v _.- -

. -. , 27 ,

<,/ ,

. _ . _ . - . -- . _..a., 4_ - -. . - - . . . - . ,

,< 7 ,

p.-_.-..-..... --.s

.' o ,,< . . - . . . . . . .

u l

y , '. j l s,

v1.:,7 , . _..

.~- ,17. _

__w -

u , ,' - -

4 ' ;, . ; /

, - . ~s / ,- / .

^*

i g *; . . - . . .. .. . -.

, . , .,' ,r . . . . . , . . - - - . . .. ,

c. -

.s . -- - s- ,_ .__. . - __ - - ...

o & >% * /

C

._--.,.e..

l c l . o . , . :

-10 m _ . . . _ . . . -

.s.--.__

. s. - --- '

..........i.

c __-_....s -

-%-\

S .

,' Cladding Failure -

- - - - -m

. .- 8

.C

--. , 4 \ . .

, p . _ . . . . . . .,r. . - - - . . . . - ,_f. .q Uppe r Rel e a s e Li mi t _

.9 p____...,,._

.. .. .-_....,<.. . N Best Esti m te . - _ _ - . . . .

o , N ~~

i 0 ,'

/ <- ' Lower Release Limit a .

c ;' . . Ir . . . . . . . . 8-u 1. 0,2 - -

e ,

7 '

o _

u .a ,,/ _ . , . . .

n 8 n ..-- ,/. -- _- ,.

_ -_ -. . . 9 e.

- f f 4 .

n ,

./, ; g . . .< _

. .. . i , ; ; i3..? -

.; ,s. -

_.1-

__ _. l\ .

.i.

a /.

.....=,:.. . .. - --..:. .

0

1 l ,.

f Mornal Operation Conc. in Drwell -

y

,i , -

1 l . l . .

. . . . , . . . .. . Uppe r Li mi t : 10'4 uti/cc _ _ _ ._

7 3 ,

.r _1 r

.. : i.

9 Nom. na) : 10 uti/cc i_

U

,4 ..

t

. 6

i4 *I

'N*. '~.iI N l k Id *I *d *8*

h. - i, .l

9 '

-- l j

m

, t

~~

m- -_. .. :-a .:.: m: . :2. - - - - - : :..: :. u. :_: -

--r--- -

- '. - . . . -:~1. -a- .: *r;:- --

)).2_._; _d _ '. :,.-.--..-. ...-n- .-.,_--*.-.._--'- .4.-

0.1 1.0 10 100 t ,

x.

r ! Cladding Failura 1.0 10 103 ,

Enclosum q l

l

- n _. n - ram R q_3 l

. > - _ . . - - . . _w z.a 3.. a4 s: .,

.a:._- a._ _ , ~

u.

\

78.000.15

, Rev. 0

+ ~

a

- Relationship Between Xe-133 Concentration [in the Containment Gas (Drywell 4 + Toruy. Gas)-and the Extent of Core Dama;;e in Ref erence Plant ..

10- 1

.._ __ . . 4

.( .-

Fuel Meltdown ,

J__ _ Upper Release Linit s if --

i

.i

-- --- Best Estimate 7- - - - - ;

. -?I .

Lower Release Limit . \, ,

5 c

. _ . :, . _ ..: .- ., . ... . . . j- i

.O . . ..

._ . ._.,s,,

19),v.c- _

~ ;

__y, - ,

f fj " / -

___..i

.,a 4_ _. . - - . . _

-i. _ . _ _ . . . . . . . . . _ . _ . - . . . . . __..,#af .* ,/

f/

...,I

/

~

v i__-....._......._-..-_..._._...

7 ,a y,'j / i, v1 '., r .

._ .x .. c. _ . . .

.~- .._ _ _ .w . _ - . . . . . . . . . . . . .. _ . . .

=_

, / __ '

v _i a ../

/

~

c5 %, i.

,' */

, /

', - / /

o , --

q3- * . _ . . . . _.

. , . .., .,r . . . . . , . . . _ . _ .. ,

- c. ,

. _ . - . . . _ _ _ . . .___...s.___e___-

5 a s% /

\

c. ,

, c !

l _ . . . _ . .

.-,.,s._'.. *

' ,~10 ^ - - ._ -- _ -

m..

e. .

8 * .s Cladding Failure - - - - - - -

, .7,' -i..x

,c .

._..,_ - - s Upper nelease Limit

...e.. , s

.f. . .

~ ' ,

o 'p _... , ' . _ .. .. . . _ ..../. . A _' Be s t E s ti ma te . _ _ . _ _ . . . .

o ,

- N . _i E i

/ <- ' Lower Release Limit j

~ .

e st . 1 . .a .

1 0,4 u .' ' -

c o

7 '

v .s -

f __- ,.' '

,_.L. a n i

-~ ,,/

n *- ,--- r - -- -

n, ,

i , ,

t v -

7 ; , . . . . . _ . . _ _ -

x /, ; r, ; i . , _!..

.1 % t s. I , ;

/ j ~._-t - - . ! i

. / ; ,

. . i s. :- : .- .

. :s -

t. ; --. i. :- . - .::: -.: - . .-

g s 0.1:; -

g, ; ,- .!ornal Operation" Conc. in Drwell .

b -,

i . . . .. . ].4 Upper Limit : 10"4' uti/cc z-i ; _,'

=+, Nominal

10.s uti/cc -

-_i 4 2 , . . . . -

2 . t. a: w .4.

.3 a i i :, .->.a= . 4. n .

Wu.d>_ 1_.: . .L.: ; --- .-t r ? .2 .a441 di d 3-- O 2iilD T-- C-.2 5 T4 W 3..

e . , - .;:. . ..m. - __ ; ..:.re .: : : u: . :.:. - - .- e n.: u: -----:--

. . . . , - . . . . . _ . . . . . . . _ , -- ' . . .- . . . - . = _7.- .: . :. : .: .:- 2 _ -3.:-__ - -

. 3 3 2 . _:- _. 2:

.d.

0.1 1.0 '10 100 F ! Clading Failura 1.0 10 Enclosure 7 s 103

> - - - .~

n. . O Pegg_K,_gR J

78.000.15 'I Rev. 0 Relationship Between Kr-85 Concentration in the containnent Cas (Drywell +

Torus Gas) and the Extent of Core Datage in Ref erence Plant -

7 -- . ;

10.,. - - - . . _ . ,

Fuel 14eltdo,n Upper n,elease Limit .

' Sest Estimate # -.

- i ; - -i . .

-N .b/_ _ - . . - - .i s

" ;' . ~*.

ower e ease i mi t

.4 .

~~ \ ' !

10 1 y x .> - -_.._

..y

.. m .

s/

./

f 1 1_ -.

. 3< <

7 .- . ,.y: _ q- . - _ - . . _ _ . . . . . .

3 - .

. . . . . . .. _ ...,s,' ..,L... . -- . . . . -

-}

,i

. s .s . . .

rr /

s / .. .. .. .

),h . / > '. .

.n,.,....,.]_...

= - ... .. ,

W .I ,/ / j u ,J

~ '. ;, r s.-

f / _

g .,

_ . . . - .. .. .. . . - ...- ..,r. .

..f-:K .y' ,.- _ -

'I '

/ > . -..-.....s e

. s. -. .. . . . .

. . r s '.: . . . a: s .s . l m

n% .. . . ..,

- - . . . . ~ . ..

u 3 0-)' ... ..

o ;. ..-....... .

--....1 C. .

2 p.

!-f <

/ -_. _ . - . . ,

.c-

. .. . . . . .. . . , .. ....,<--..-F- Cladding Failure.

o < . . . . . _ _ _ _ . . . '

c

_s.. s.

O .

/ . _ _.C .

Upper Release Limit e

.- /. ./ .x x

- i .

. . est s tw.. ate .

e o l un ."-

/ ,' LcNer Release Limit -

. . , -y i

., m

,= ,/ ,/ .

a , < .

/ ,- '

E.

t e

T

s

-f .-

,,v .

., . - - - - t.

c ,,- . .. , .;, t f /. a

u. , ;

- +

y: 9 -t - n .

.- .--------t cc

'i - i l~~. -1 -

=i I l'.

-L: = . .

!/

u )0 ;,, p . i . - ~ '

' .~

orral Operation Conc. in Dryv: ell .

1.-'-

. y !

Upper Limit : 4 x 10-5 uti/cc l,

[1

1 .;

j Nominal : 4 x 10-6 uCi/cc w,1. . 2 .-!. - ,i . . ,o I . '.

l -, .. .

4 .. 'i i. ..

i. ;

0:nvin r- m .1: .- , .: . . - -- . . : A = E 2 1m f 4 ,.

. , p -

- .; . ! n ; . . . ,

3

- - rs.

. :2 :._: . .: .: : .1. -

. :---- a =: . .: , p._ -

u m::: _..:-:-:.;-.-: . . .::.: _. ..:-.w :.; . . a. , - -1 .33:y g 4,,= : .. ,.: : = .: a a : 100 0.1 1.0 10 !

Cladding Failure 10 100

1. "

m r : Fuel Meltd:Nn Enclosure 8 Page 1 of 1

y- .

. . L l

[

BEST-ESTLMATE FISSION IRODUCT RELEASE FKACTIO!!S [

1

. t Gap Helease d Meltdown Release oxidation Release Vaporization Release Lo wr Uppe r lo w r . [5per tow r Upper lo w r Uppe r {

Nominal Limit Limit flominal Limit Limit Nominal Limit Limit Nominal Limit Limit (

. t Noble Cases 0.030 0.010 0. 12 0.873 0.485 0.970 0.087 0.078 0.097 0.010 0.010 0.010 (Xe ,Kr) ;

Haloc,cns 0.017 0.001 0.20 0.885 0.492 0.983 0.088 0.073 0.098 0.010 0.010 0.010 \

(1, lE)

Alkali thtals 0.050 0.004 0.30 0.760 0.380 0.855 --- -- --

O.190 0.190 0.190

! ( Cs , Rb) '

0.250 0.340 0.340 0.340 ;

TaLlurium O.0001 3x10-7 0.04 0.150 0.05 0.510 0.340 0.680 Group (Te ,se ,sb)

Noble Ihtals --- --- --- 0.030 0.01 0.10 0.873 0.776 0.970 0.005 0.001 0.024 I i

(Ru ,Rh ,Pd ,ilo ,Tc) 1 I

A1 ka1 ine 1x10-6 3x10-9 0.0004 0.100 0.02 0.20 -- --- ---

0.009 0.002 0.045 f Earths 1

'(Sr,Ba) !

' Rare Earths --- --- ---

0.003 0.001 0.01 --- -- ---

0.010 0.002 0.050 (Y , la ,Ce f41, t

. Pr , Di , Pm , Sn , 6

, lip 1%) ;

e

-- --- --- 0.003 0.001 0.01 -- --- -- --- --- ---

[

,Refractories '

fZr,tb) {

A EP

< B.f

. .O &

c? I a >

{

9 SAMPLES tiOST REPRESE?TfATIVE OF CORE CONDITIOf!$ DURING AN ACCIOCNT FOR Tile ESTIMAT10:3 0F CORE In!! AGE Break Ottegory/ System C4,nditions Sample Location Other Inst r uc tions Supp. Espp .

Je t ibo t abol l Ptimp Liquid A t rios . Kl!R Drywelll Small Liquid Line fireak, Heactor Power >l % Ye s --

Yes I --

Yes2 Small Itquid Line Break, Nactor Po we r (1 % -- --

Yes I Yes Yes 2 1. RitR minst be in shutdown cooling nale .

2. Reactor mter level must be raised and flow f roa moisture se pa ra to r s .

Sm al) Steam Line Break, Heactor Power >1 % Yes -- Yes l ---

Yes2 Small 5 team Line Break, Heactor Power (1% -- --- Yes I Yes Yes2 1. RilR must he in shutdom cooling mode.

2. Reactor mter level mtst be raised and flow from noisture separators.

Lardo Liquid Line fireak, hactor Po wer >- 1 % Ye s3 ygg 4 Yes I --

Yes2 1. Suppression pool must be in suppression cooling male .

large Liquid Line Break, Heactor Power <1 % --

Yes' Yes I Yes 3 Yes 2 1. RitR iats t be in shutdo m cooling mtale ,

g' a 2. Suppression pool must be in ff y

,0 suppression cooling mode , f-4@S 3. @ actor mter level cust he cg

raised and flow from moisture _

se pa r.ito r s . w' Mg

_4

e i

SAMPLES MOST REPRESENTATIVE OF CORE CONDITIONS 111 RING AN ACCIDENT FOR THE ESTIMATION OF CORE IW!AT Breck Category / System Conditions Sample Location other Instructions Supp. Supp.

Jet Ibol Ibol l Pump Liquid Atmos. RilR Drywelll Large Stea:n Line Break, Reactor Power >1 % Yes3 ye 3 4 ___ ___ yes large Steam Line Break, Reactor Power <1 % -- --

Yes I Yes Ye s 2 1. RilR must be in shutdown cooling mode. !

2. Reactor unter level must be raised and flow from moisture se pa rato r s .

1.. the if SRV's are vented to the suppression pool.

2. Use if SRV's are not vented to suppression pool.

3. Use if makeup water flow is <50% of core flow present.

4. Use if makeup water flow is >50% of coce flow present .

r G

2g -

?u 4

O "o,a, .

oo*

8

' m fD w

n g

.___

ML20052E374

Text

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