ML19347A388

Chapter 7 to State Univ of Ny at Buffalo Final Hazards Summary Rept,Revision 2, Hazards

ML19347A388
Person / Time
Site:	University of Buffalo
Issue date:	09/23/1963
From:	NEW YORK, STATE UNIV. OF, BUFFALO, NY
To:
References
NUDOCS 8104080572
Download: ML19347A388 (26)

Text

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faculty members and personnel of participating industries vill frou I

time to time be appointed to this nuc1cus to provido the varied bach-

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ground and compctcace for adequato evaluation of proposM cxperiments 4

and operational changes.

t A Gcacral IIanager is responsiblo for most administr',tive I

activitics of the Center. Reporting to him is the staif responsible for maintenance, personnel, accounting, ei.c.

He is also responsible i

for all contracts involving services supplied by the Center and all i

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contracting of the Center.

The following organization chart outlines the relationships of the aforementioned functions:

1Tructeesj i

IChairman 6 PresidentI l

1 Drecutive Vice PresidentI Director Nuclear Hazards l i

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Committee i

General

' Operations 2adiological

?.esearch Manager Manager S.afety Manager Officer VIII. Ec:ards This section vill encompass all of the poenible hazarda to which the Uestern New 'lork Nuclear Research Center Research acactor might I

be subjected. Ecch hasard will be evaluated in relation to reactor design and operation.

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Four ;;;oups of hczards vill be discussed:

A.

Natural hazards B.

Hinor accidento C.

Liaximuu start-up accident

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D.

Unximura credible accident The um:imum credible accident is the worst accident that could occur under any forsceable circuustances. It will be shown that the occurence of this vorst accident will not jeopardizo the surrmindins population.

A.

Natt'rni Hazards

.I As discussed in Section II, there cre uc signific sat natural phenomena which uould contribute to a radiation hazard from the The Buffalo area appears quite favorabia, hazards-wise, reaccor.

from the point of vicw of meteorology, hydrology, geology, and seismology.

B.

Hinor Accidents In the following discussion, %inor" accidents (which produce less severo consequences than a maximum credible ac.cident), together

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with their possible results, are discussed. In ctses where hazardous l.

l conditions are predicted, a description is given of preventive or 1'].

corrective consures 1scorporated in the facility design.

1 Tocs of Ventilstion:

Loss of ventilation in the beam tubes, thermal column, and

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in the reactor room itself vill be conciderod. In the beam tubes, the major contribution to air activity r,sults from the A (n, f) A

1. _

r.acuon.

i!l 2

4 _

5

i5 13 Uith a flux at the core end of the beca tubes of 1 x 10

$_m n/cm /sec., th3 M 1 activity would be about 3.3 x 10-5

,,71,,f,,,

s f

The total volume of the six 6-inch cnd 12-inch square tubes is 1.09 x 10 cc, assuming that all but 50 centimeters of their length t

is plugged. If n'.1 tubec were opened ct once, (patently impossible),

g this would releaso 3.6 curies of [ 1 into the reactor room. If the i

volume of the reactor room is assumed to be 6.3 x 10 cc and the A'1 l

2oI -

is assumed to be uniformly mixed vich the reactor room air, the 01 resulting A concentration would be about 5.7 x 10 curics/cc.

II e

only one 6-inch tube was opened, the concentration would aimilarly

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be 5 x 10~11 curies /cc.

The activity contributed by the thermal column is siuilcrly calculated to be approximately 10~11 curies /cc of reactor room air.

The ma::inum permissible M1 concentration for a 40-hour exposu.e in contaminated air is 2.1 x 10' curies /cc. CF2 Title 10,

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Part 20 does not list the maximm pe: =issibla concentration for the 4

case of inhalation of the contaminated air. The radiation dose, a man o

receives from the gas in his lungs is considered to be negligible

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,f, compared to the dose from a large cloud of gas surrounding him.

Bovaver, if he were to inhale the total activity from one 6-inch beam l

tube, then it would not be negligible and would be a health ha:crd.

The activity of one tube is.30 curies, and it is not evident how all the activity could be inhaled. All the beam tube activity, when r.ixed uniformly with reactor roca air, presents a health hazard if personnel remain exposed for more than nine minutes,

'ti il 82-E

Loss of ventilation in the reactor room is not enough to result in a health hasard. The two possible sources of activity in the reactor roou air under normal opercting conditions are water i

dispersed cetivity, released as the tank surface evaporates, cad gases dissolved in the tank water which are cctivated, and later come out of solution. Neither of these sources releases activity in significant quantitics.

(Some gaseous activity has been noted at the LIT 2 Pool surface when the tant is opened after a one-week operating period.

The activity level is low enough for the gases to be released into the reactor room without treatment. )

h__CoolancPumpFailure:

In the event of failure of the coolant pump, or accidental

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shut-off of the water exit line, the low flow condition will auto-matica11y scram the reactor, and the flapper valve below the core will open, allowing convection cooling by upward water flow.

Calaulations have shown that the sore will not melt during the flow reversal period.

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3.

Ifaterlonninc:

An important consequence of the use of unboaded fuel elements I

is the possibility of a waterlogging fcilure. Such failure could 1

occue in the event of a defect or lenk in the fuel clodding, which would permit inlaskage of water during low power operation, or at l

j shutdown. During a subsequent pulse, pressure would be generated in

. the annulus between fuel pellata and sheath, due to the inability of the steam to essape rapidly through the defeat. A failure of the pin sould result.

i iI P.ro ccscs of such waterloggin,3 failure have been documented, i

both of which occurred in conjunction with' cuperinents in which holcs were drilled clarouch the cloddin; to accomodate thornocouples. The first case, in which a scrics of holes vera drilled in a line, a pulse having a period of 7.5 milliseconds resulted in a rupture 12 inches in lonath, which followed the, line of holes ( 1 ).

In a aiuilar case involving only one thermocouple hole, in which the expony scalant deteriorated and admitted water to the tube, a two centicoter vertical crach developed in the center of a small blister which formed after a pulse ( 2 ).

rI L,

Since the fuel is not molten at any time during the normal pulse, only the fission products contained in the annulus would be a released, and, in fact these fission product probably would have escaped during the original inleakage of water, and might serve to o,

indicate the presence of the defective pin.

,'I The possibility of a waterlo83ing accident leading to the o

destruction of adjacent pins must be considered. If it were L"

postulated that all the energy in the waterlogged pin (less than 0.26 megawatt-seconds) were released instantaneously to the water in the n,

fuel assembly, the resultant pressure surge might damage adjacent

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I,,,

pins. However, since the fuel is in pollet, rather chan powder form, it is unreasonable-to espect an instantaneous release of the q'"

contained energy.

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Waterlogging, while possible, does not appear likely, and in any event, cannot be eensidered to be the maximum credible accident.

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4 Loss of Pool Water A. complete loss of pool water in a short period of time would uncover the core while a significant amount of power is still i,

being generated in the feel pins through the fission prodoct decay i

I heat mechanism. In investigating the consequences of such an accident, I

it was assumed that the reactor had been operating at a power level of 2 megawatts for an infinite time, and that fission product con-centratinns had attained equilibrium.

Calculations, based upon experiments at the LITR in i

which loss of water was used as a shut-down mechanism, and in which I

fuel plate teoperatures were measured, were made to determine the maxi-

. mun surface temperature of a fuel pin. At the LITR, the maxisma plate temperature was reached approximately one hour after uncovering the core.

Fuel plate temperature data from the LITR at 1.0 megawatts

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and at 1.5 magavetts were extrapolated to 2 messwatts, and then were adjustad to reflect differences in hydraulic and beat transfer charac-teristics,. The resulting =awi== surface temperaturo becomes 1305*F, which is well below the molting point of aircaloy; therefore, no clad meeting will occur. Moreover, under conditions of gross pool water loss, t

l_

the installed emergency spray system would activate, and the foregoing analysis becomes academic.

5.

Maxioma Starton Accid.enti For the purpose of evaluating N =awi== start-up l

accident, it was assumed that all six rods are driven out at e=w4==

spesd, and that no corrective action is takes until the power reaches the high levet trip point (2.5 megawatts).

Itami== withdrawl rate for l

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the rods is e inches,minut.,.,1ch v. suits 1n a reacci.1=, insertion rate of 0.27%- delta h/k/s'econd, if it is assumed that the rods are at' positions of maximum differencial worth.

It was assumed that a 25 curie antimony beryllium source is used for startup indication, and that half the source neutrons escape the reactor. With a sub-g critical n:ultiplication factor of 0.764 for all rods inserted, the I

-3 source power becoucs 1.13 x 10 vates.

i During the maximum startup accident, two shutdown mechanisms l-are available; namely, the removal of reactivity by rod insertion (scram), and the Doppler coefficient.

If the total pulse time is i

I, greater than 90 milliseconds, which corresponds to the lag timo in l

the cicctronic system, the motion of the shim-safety rods will reduce the power before the pulse peak is reached. If the total time of the pulse is loss than 90 milliseconds, the Doppler coefficient will reduce the power prior to the initial motion of the shin safety rods, and the subsequent motion of the rods will serve to clip the tail of l

the pulse.

Thus, the peak power computed on the basis of a Doppler i

shutdown will be the amminum power experienced in the start-up p

l' accident.

It is assumed that the total reactivity inserted prior to 1I, i

the release of the safety rods by the high level trip (0.943% delta k/k), is inserted as a step addition. Further, the power for a step additto,s is proportional to the square of the prompt excess reactivity to the fi st power.

By comparison to the prompt excess reactivity required for a 40 megawatt-second pulse, the peak power obtained in l

) E, the maximum aeatcup accident is 336 megawatta and-the tocal energy j

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-is 17.6.megauntt-seconds. Since theso values do not c;;cced tha values (l

for the. design pulso, ute maximuut secrtup accident poses no danger l M to the safe operation of the reactor.

9E 6

_6. Transicat aod Fail; to Return to Starting Position:

JE Ac described in Section IV J, the transient rod auto-metically returns to its original position following each pulse.

li If, however, the transient rod should rentin in the fully withdraun position following a pulse, the reactor power will reach an equilib -

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rium level where the tewperature and core void coefficients will g

compensate for the reactivity worth of the e,Jected transient rod.

For the temperature risc alone to compensete for the reactivity (1.G267. delta k/k) inserted for the 90 megawatt-second a

pulce, the average exit temperature from the core must be 208.4' F.

The power level corresponding to this temperature is 2.5 megawatts.

Further, the corresponding peak heat flux, assuming a flux hot spot factor of 2.5, is then 116,000 beu/hr/ft. 7ttia heat flux is 2

considerably less than the burnout heat flux of 836,000 btu /hr/ft,

E caleslated by the liirahak correlation ( 3 ); therefore, melting of the clad will not occur. It has been calculated that the surface i

of the clad at the hot spot will be 273* F, and nucleate boiling vill take place in the asserslies of highest flux, thus providias an additional negative reactivity effccc. }breover, a credit vould be afforded by the Doppler coefficient at 2.5 megawatts. It is reason-able to conclude that the condition of the transient rod sticking in its fully withdrawn position does not represent an approach to melt-

'down conditions and is therefore an accident less than the maxismaa

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a C.

Maximua Credible /.ccident 1.

Cause:

'I Computations indicate that crroneous handling of fuel, in contravention of established administrative controls, could result in the maxirma credibl.e accident. The following assumptions must be natisfied:

(1).

.I The core has been loaded in the optimum configuration.

(2). The reactor is critical.

I (3). A fuel assembly is dropped into a grid position of cy.tremely high worth.

e (4). The high level power trip scrams the reactor.

The position the configuration which gives the greatest a

worth to an asocehly is at the base of the T' in an open core, as shown by Table A2 This positien results in a uorth of 3.87. delta h/h or $5.00. For purposes of analyzing the maximum credible accident, it is assumed that the assembly is dropped frou a height of two E

,!g feet above the core, and subsequantly the assembly enters the optimum location with a velocity sufficient to initiate a 180 megavatt-second pulse. A'n anergy release of 100 megawatt-seconds is sufficient to melt the UO2 pellets in regi as f the core corresponding to power densitics 3.3 tiaca the averoso; however, uclting of the UO doca 2

not in itsel.C cignify destruction o the cled integrity.

Instances of UO2 cca,3cratures well above the eciting temperature without incurring clad failure or deformation have been reported ( 4 ).

I Helting of the UO does indicate that a ler o temperature difference 2

exists between the full pollet and the coolant water; consequently, a high heat flux will exist. The W.T-1 code is unable to account

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for a boiling surface condition although it is evident that boiling will occur during a 180 mescwate-second transient. An inequality approach was used to appro:cimate the heat fluxes generated during a

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transient:

1.et T

=

1 Temperature of outer surface of zircaloy i

6 T

2 Temperature of inner surface of zircoloy j

=

1 T3 Temperature of surface of fuel pellet

=

Then T

> Tsat for boiling to occur 1

1+Rh T

T1 + (T3-T)

=

= T 3

1 T,,e+Rk T3 >

Also T3 _ T,yg Therfore T,,g = Ein - Ecut C1 And Elg - E,g T,,g + R S. < 1 Ein - Eout

= T A

R C

sat i

l Ilhore k.

heat flu::

=

R thermal resistance of gap and aircaloy

=

u 1

l Ein energy of pulse

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E

=

out energy which has leaked out of pin at a given time

(

=

9 lg C

=

t constant containing specific heat and 3

Cecmetry terms E

T,,g saturation tsuperature _of water

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It is assumed that the pulsa energy is absorbed adiabatically. Then 5

at time aero, E, = 0 and a number larger than the initial g/4 may be calculated.

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. E, Since the peak heat flux is greater than 3 x 10 btu /hr/ft,

burnout will probably occur at several points in the core. Some degree of mitigation will be realized from the fact that T is 3

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less then Tava, and therefore, the flux will be less than that indicated. Ibreover, in the experiments previously mentioned, conditions as severe as chose anticipated in this accident were experienced without loss of clad inte3rity ( 4 ).

It should also be noted that transient in-pile melt down experiments indicate that failure under conditions of transient haating is not violent regard-less of the cladding material used (5 ).

Other data, while not directly applicable to an oxide core due to differences of the time constants of the materials involved,

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indicate that heat fluxes considerably higher than those fluxes generated at etcady state can be accommodated without burnout daring transients ( 6 ).

Since administrative procedures prohibit leading of fuel

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at criticality, and a 'U" core normally would have an experiment installed in the "U", precluding insertion of fuel, only a combin-l l

ation of gross violation of operating procedures, with an extremely

/u improbable drop of a fuel element from a height of greater than I

l.4, two feet above the grid into a position of optimum worth, the credibility of this accident can be disputed. However, because

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retention of fuel integrity cannot be demonstrated unequivocally, the consequences of a partial meltdown will be considered.

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a 4 mm g r,

2 Results:

Coincident with the partial molting of the fuel assemblies,

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a poreton of the energy release will be converted to seca.s, which will be released to the containment. Less than 60 pounds of steam, containing fission products, will raise the building pressure to less that 0.5 psig., and the contained activity will trip the building dampers to seal the building completely. The da=per on the i-cicanup system will open, and the exhaust fan will reduce the build-ing pressure, through absolute filters and an activated charcoal bed, to the designated pressure of-0.018 psig. Approximately 6,500 ft.3 or 3.5 por cent of the building air will thus be dis-charged from the 165 foot stack in slightly more than five minutes,

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to affect this pressure reduction. Since the measured leakage at 0.5 psig. is about 105 ft.3/ min., a negligible quantity of activity will leak from the building at ground level.

Subsequent to the almost instantaneous 3.5 per cent puff, l

a continuous b1 cod-off through the cicanup system will maintain the 6

building nt-0.018 psig.; thus, no leakage at ground Icvel is possi-ble, and all activity will be released from the 165 foot stack.

I The fission products which contaminate the building air arise primarily from the molten fuel G alculated to be 2.1 per cent

' of the total core fact inventory, alrW for purposes of this analysis a conservative figure of 5 per cent has been used), and j.

from the gas in the annu11 between the cocida pellats and the sitcaloy sheath of those pins where clad failure occure (50 per cent).

f Table 24 presents the saturation activities of signifi-I_

' ' " * ' " ' " " ' * ' " * " * " ' ~ ' ' ' " ' " " * " " ' " ' " ' ' " - "

a l-

-91 i

I

I due to diffusion, after 1,950 days of operation at 2 megawatt power.

TABLE 24 i

ACTIVITIES 07 FISSION PROLTS IN HIEL MJU:2 Isotope Curies Icotopo Curico

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Kr-88 1.57 Ca.137 0.128 i ~~

Xc-133 22.6 Cs-138 1.10 Xc-138 0.81 Cs-139 45.5 I-131 10.7 Sr-89

'43.7 I-133 13.6 Sr-90 1~ 1.6 Br-83 23.9 Sr-91 1.51 Be-84 0.20 Tabic 25 shows how the significant isotopes from the gross core activityare attenuated in the U02 and the last column i

indicates the total released activity, including the contribution

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by the activity contained in the annular gas. It is assumed that 100 per cent of the moble gases, 62 per cent of the halogens, and

,.,gg 1 per cent of the non-volatiles are released from the molten fuel ( 7 ) ( 8 ).

d Tabla 27 gives attenuation factors achieved, including building control and filters for release to the environment of

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those isotopes of significant hasard to the general population.

i* ese lu_

..m l'g

.,,.-,-.-_.-.e

-...,,,.,.,,. +,. -

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Ib TABLE 25 3

FISSIO11 PRODUCT RELEASE TO COUTAIICEttr Tota'l Core Annu11 5% Heltdoun 100% Noble I

Inventory Activity

. Activity cases 50% Halogens 1% Nonvolatiles Helted Fuel Total Release, Isotop,1

_(Curies)

(Curies)

(Curies )

_ Curies curies curies and Annuli Kr-88 6.16 x 104 1.58 3.00 x 103 3.03 x 103 3

3.08 x 10 i

Xe-133 11.5 x 104 22.6 5.75 x 103 5.75 x 103 5.75 x 103 Xe-133 9.92 x 104 0.81 4.96 x 103 4.96 x 103 4.96 x 103 1-131 5.36 x 104 10.7 2.63 x 103 1.34 x 103 3

1.35 x 10 i

I-133 11.3 x 104 13.6 5.65 x 103 2.83 x 103 2.84 x 103 Br-83 0.88 x 104 23.9 0.44 x 103 0.22 x 103 0.24 x 103 Br-64 1.56 x 100 0.20 0.78 x 103 0.39 x 103 0.39 x 103 Cs-137 0.125 x 100 0.128 0.063 x 103

.63 0.76 Cs-138 9.92 x 104 1.10 4.96 x 103 49.6 0.050 x 103 Cs-139 9.51 x 1C4 46.5 4.76 x 103 47.6 0.094 x 103 Sr-89 2.80 x 104 43.7 1.40 x 103 14.0 0.058 x 103 Sr-90 1.25 x 104 11.6 0.63 7. 103 6.3 0.018 x 103 Sr-91 10.0 x 104 1.51 5.0 x 103 50 0.052 x 103 t

L T'

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- c y ~ -....--. -,,

.m_,,...,,..,.

L t w.ltea iot. of-the hazards fc;:;cact. cad :2y cl.c c6nta::zinated air resulting froci a partial meltdown and its release will be made for the follouing locations:

(1). Dounwind exposure during the 3.5 per cent " puff" releas e.

3 (2). Downwind exposure during the 5 per conc daily release.

L (3). Reactor building interior.

(4). Reactor building c::torior.

The intent in this evaluation is to assess the maximum activity which can be released from the stack under inctoro-logical inversion, either as a 3.5 per cent puff or a 5 per cent continuous biced, whichever condition is n; ore serious, without exposure of the general public at the point of maximum exposure to to doses in excess of maximum permissibic. Applying attenuation factors of the physical conditions, it will then be demonstrated that a partial core meltdown can be controlled withcut hasard to the general public.

A.

Release of 3.5 por Cent puff Prom Sutton's '*Ibaory of Atmospheric Diffusion" (9),

the total integrated dosage (TID) which may be received at a point on the ground a distanco (: ) downwind froi.: :! nonisotropic, iimcan-cancous point sourco located at a height (h) is, 2

h 20 IC C x(2-n)}

(1)

TID =

y TTCCs)1XI ~"}

y where: TID total integrated dose (micro-curies-seconds per

=

motors 3)

Q = source strength (:aicro-curies)

L.

m L

3 B.

(

(C, C ) = Sutton diffusion parc:acters (ucter2 y

z

= non-dimensionci stability paracccer n-

/1 = averago vind speed (moters per second)

'tha caimum of equction (1) may be obtained by differentiating and maximizing.

(TID) ma:: =

2@r micro-curies-seconds per caters 3 7/ejIh2Cy To obtain the total body dose (TBD) due to inhalation, the (TID) must be multiplied by the breathing rato and the faction of i

inhaled isotope which is retained by the body.

TLD = (TID) J f, l'

J = breathing rata (2.31 x 104 3

i.

meters /second) f, = fractim retained (listed in NBS #52))

4 hrefore, the source strength Q, in microcuries permissible to edieve control within *= a permissible body burden at the point of =mv 6 m dose, is given by, i

2 Cv) q,TBD (max) (6 c fh fa 2 JCs A large inversion, though a rare occurance, will be considered for 1

the 3.5 per cent " instantaneous" puff. The diffusion persmeters chosen for a large inversion are, n = 0.6 C 4 'O.22..

. 'c,= 0.12.,

u = 2 deters /sec.

y l

The maximum sourco strengths (Q) permitted for I-131, Sr-89, Sr-90, 1

(considered to be the most hazardous of the fission products) are shown in Table 2G on the assumption that TBD permitted is not I

exceeded at the distan,.e of maximum dose. The values of Q repre-I sent the inventory of tBase fission products present in the l

building air.

L TABLE 26 T3D (im )

Q Q'

g Icotope a

cuden curies curie:

I-131

.23 50 3.8 x 1010 0

1,33 x 10 10 7

Sr-09 23 40 1.4 x 10 5.0 x 10 Sr-00

.12 20 1.7 x 1010 7

1.8 x 10 Comparing the value of Q' with Q, it is apprarent that if the entire inventory of fission products in the buig air ucro to be released instantaneously from the stack under condition of larac inversion, small hazard wculd result to the general population (Icss than permitted occupctional exposure.)

It is inconceivable that the entire building inventory 4

could be instantaneously released under any circumstances. Next, attenuation factors based upon the assumed building conditions will be applied to the isotopes of interest for evaluation of source i

I strenschs which might be expected.

The stack filtering system consists of roughing filters i

followed by CUS-6 type absolute filters. These filters will provide an attenuation ~ factor of 10'4 for 0.3 micron particles (lQ 11).

It is estimated that more than 90 per cent of the particles from the

'I BORAI excursion wou18 be la.ger than 0.3 micron since the fallout from this excursion fell within a 350 foot radius. It is felt that 3

the particle size will rance from 0.1 to 1.0 micron. The minimum 3

attenuation factor for these particles would be U x 10' (12). It a

seems reasonable then to assume an overall attenuation factor of ii i'

4 x 10-3.

For iodino, the attenuation factors are not vell docu-monted. As has been indicated, there may be steam in the contain-mj ment cylinder. As the steam condenses, it is reasonable to assume the.

9 l

L some of the iodine will dissolvo, since see::m can dissolve more than 100 times the anticipated quantity of released iodine before becoming saturated. In addition, activated carbon traps in the filter system will recove aignificant quantities of iodino. The Connor Ensincering Corporation has conducted tests with activated carbon absorption units and rates of 95 eer cent efficicacy have been obtained. The three wits in series should therefore provide at 1 cast 99.9 per cent I

cfficier.cy. From these considerations, a conservative attenuation factor of 4 x 10~3 has been chosen for volatilized iodine.

l The following tabic presents the various attenuation 1Ejg factors considered herein (negiccting radioactive decay).

TABLE 27 PRACTIOIML RELEASE OF TOTAL CORE AnynT Filter Puff Total Iodine Vapor 0.004 0.035 1.4 x 104 Metallic 0xides 0.004 0.035 1.4 x 104 In evaluating the seriousness of airborne hazards based ii 1-on these parameters, several points of conservatism should be l

13

noced, 1 E, (1). C marvative values were used for filter efficiency.

(3. No attenuation u=s assumed due to dissolution, chemical combination, vash-out or fall-out in the reactor room before materials are released from tho lj stack.

Applying the attenuation factors in Table 27 to the lI 1 E, values of Q' in Table 26, and then comparing the result with values of Q in that table, it is clearly indicated that the initial 3.5 por cent puff will'not result in a hasard to the general population by a safety factor in excess of 1 x 10,

5

-97 m

,,, - -. ~

-,.-y 4

3.

continuous nelease i

For the cacc of the continuous release, it is assumed that the fission products are instantencously dispersed in the reactor l

building, and after' the initial 3.5 per cent puff, are continuously released at the rate of 5 per cent per day from the stack at an f

cicvation of 165 fccc.

l The Sutton formula for the ground concentration (K) at a t

l distanco (x) meters downwind and (y) meters crossuind from a continu-l ous, nonisotropic point source located at a height (h) ir, S

a(z 2 + C 2) x -2 Y

n 2

K(x) =

y (2)

C C 5 X ~"

yz 1

K(x) = concentration (micro-curics por meter 3)

= average rate o' energy releasc (micro-curies per second)

The other quantitics are the sar e as those defined before l-in equation (1). In calculating (K), the crosswind (y) is set equal to zero. This gives the concentration along the centerline of the

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source. The maximum of equation ( 2 ) may te obtained by differen-tinting and maxialsic, The maximum concentration is given by, S

2 '1C K max =

C 2

7feh KCy The maximum permissibic airborne concancration for

.]

exposure to the critical organ after exclusive inhalation of contami-I~

nated air from time (t) is,

]

3 x 10-2 qg2 ne..r g, g -O.69t/T) 7 t

I

== -

L

Where, i

).

NPC = maximum permissibic concentration (nicro-curics per meter )

3

^

q = micro-curies in totoal body to giva 0.3 rem /wk cxposure to the critical organ f2 = fraction in critical orgen of total activity f, = fraction retained by inhalation a

T = effective half-life (days)

'l The expression for K max is set equal to that for ZGC and dQ/dt (max), is obecined, 3x10-2,f7,,3fC 2

~2 7

'a dQ/dt (max)=

jg 2 C T fa (1-o-t/T) i Table 28 lists the constants chosen for the isotopes (13).

Fc,e a moderata lapse, Cy = 0.3 C, = 0.25, and 5 = 6 meters /sec.

TABIE ta l

l d Q/dt (max) dQ/dt (max) e g

1 day 10 days gotope L

-?

fa _

T 4.c/ day v c/dev _

I131 50- 0.2 0.23 7.6 1.59 x 105 2.16 x 104 09 Sr 40 0.99 0.23 50.4 6.7 x 105 4

5.1 x 10 Sr' 3

20 0.99 0.12 6.4 x 10 5.6 x 105 5.7 x 104 Tabla 28 lists the =earimum' permissible leakage race, l

dQ/dt, to deliver no more than )cc at the point of maximum dose rata downstrean.

To assoas the ability to control hasards associated with a continuous release, it is postulated that the fission products are practically instantaneously dispersed throughout the reactor room air, and then allowed to leak out the stack at a controlled rate I

-,9

L t

3 l-(L uctors 73cc,), -The averego chc. age in the rate of caergy release is given by, da L Q de' 3"Y (3)

L = Icahage rato (maters /sec.)

3 V = volume of reactor room (ceters )

The value of Q (c) at cny time is given by:.

fe"AC l~ 3.12 x 1019 (1-e-AC)

P q,

(4) 3.7 x 1010 G ero t = seconds after reactor shut down P = power of reactor hilovatts T = operating tic:a of racetor seconds f = fractional yield of j

A=decayconstant(sec"givenisotope

)

However, aLuce Q continually decreases with time due to decay and repeated dilution of tha source strength remaining in the room, it is more realistic to obtain an avarage rate of energy release, (T

,g q,

,/Q (e) dt j 2%

7-where [ = time of exposure 1!-g Substituting equation ( '4 ) for q (t) in the above a=pression, l Q and integrating, AI M

' 3g Q=

4 (1-o

)

y where Q,= cora inventory at shutdown.

Substituting the above expression in' equation ( 3

),

b=SL (1

m AI)

, gs dT y

lt.

Af l

Ly applying the atton'uation factors listed in Tables 25 and 27,

j the values for the last colutut of Tabla 29 are obtained.

f Pera!.ssible' release rates for one, and ten day exposures are q

listed for ocuparative purpooss in the third column of Table 29

.1

'11

-100-

it i

TABLE 2)

Isotope Exposure dG/dt (cax) 43/dt

.uc/ day

)<c/ day 131 I

1 day 1.59 x 105 1.27 x 102 10 days 2.16 x 104 6.7 x 101 09 5

2 Sr 1 day 6.7 x 10 1.93 x 10 4

10 days 5.1 x 10 1.93 x 10 90 Sr 1 day 5.6xlof 3.39x10N 10 days 5.7 x 10 3.39 x 10 comparison of dq/dt (max) with the anticipated release

~

rate dQ/dt clearly indicates that the continuous release over a 10 day period of 5 percent / day of the building contents trould not constitute a hazard to the general public. A miMmnm safety factor of about 10 is apparent.

C.

Mc:ards I!ithin the Reactor Building In order to estimate the gross gamma sourca existing within the reactor building, the method of IMP-72-2 (24) will be used. The method consists of analyzing tha volatile fission products which included the isotopos of Xe, Kr I, and Br, and eliminating from consideration those isotopes which have long half-lives (>l year),

no assmnas, low yicids, or very short half-lives (<1 min.). In this way, the fission yield M-a? ? = 47.3 per cent and the average

-1 energy per fission becomes 0.297 Mav. The fission rate is given

~

theoretically as, Q = f e~' " 3.12 x 10D (1-e

) disintegrations /sec.

~

P Since the only fission products considered are those with, 1yeaT1/2>1uinuta than)

I and for.'-

r (operating time of roastor),

A

~

-101-

t 1-c' 2:.1 and if only a sho.t time (t) after the accident is considered,

.l t 1, c~ A tal so that the disintegration rate of the fission product is independent of the isotope and becones, Q = 3.12 x 10D f?

Conservatively cssuming that the fission products associated with the 5 por cent of the core that is considered to have melted era eject into the building. air, cud for P = 2.0 x 103 kilowatts, f = 0.473, E = 0.297 acv., the enern flux is, S = 3.12 x 1013 x 0.473 x 2.0 = 103 10 x 0.05 = 4.38 x 10 mv/sec.

Assuming the volatilo fission products arc distributed uniformly throughout the air of tha reactor contain=cnt, (assu=ad 3

^

voluna 5.27 x 10 cm ), the source strength per unit volu=a is

=

0 :ev/sec-cm. The integration can be simplified by 1

3 S,= 8.3 x 10 approximating the reactor containnant by a cylinder uith a radius of 35 feet snd a height of 50 fcot. sinec the scif-cbcoiption in the 53 feet of air is negligible, the unshicided energy flux at a point on the axis of cylinder cad 0.5 feet from the and of the cylinder is given by,

~

q b

Il S dx 2 TT y dy

=,f y

1 O

X 0

4 p (x2,7)2 I

2 2

(x in (x2.y) 7 X I

= _Sy.

2 4

x2 y1 tan-1xjX

+2 g

7

i. !_

1 I

l Evaluating this expression with xg = 0.5 feet x2 = 58 feet and yt si 35 fact gives, 3

2 I =

(2.59 x 10 ) = 5.4 x 107 mev/en,,,,.

o

-102-

1 i

L i

i converting this energy flux to a dose rate, DO = 2 :: 10-6 x 5.4 x 107 = 107 r/hr.

This is casentially the dose rato for points on the inside surface of the containracnt cylinder ecll. Evacuatir,n fi:no~ for the building (appro::imately 2 minutes) restricts the exposure of operators (1.8r) to substantially less than permitted emergency enposures for occupational personnel of 25 r.~

D.

E::to: mal IIazards To calculate the dose at the outside wall of the containment cylinder, nauinal point of closest approach (20 feet),

and the nearest residence (6out 500 feet), the method outlined in the Reactor Shielding Design. Hanual is used (15). f.ssuming a line source at the center lina of the containment building, the iatensity i

of radiation'at a point against the building wall at ground level can be represented by,

.f' = 2TTa L

F /Sb) where h = mev/cm -sec.

I g

L5 3

=

build up factor T/

S

=

g line source strength distance from the line source to point outside building a

=

G angle of clevation of line source from point outside

=

building b

= ~T/ ~,6 A

The value S is calculated assuming total flux (4.38 a: 10 mov/sec.)

10 g

is corcentrated along the axis of the cylinder, N

Sg = 4.38 x 10 (58) (30.3) = 2.46 n 10Ll.sv/ u-sec.

y

-103-it

Uith the value a = 37 feet cr.d

= 37.1, 7 (?r. b) = 1.07 n 10~3 and 0

I,

(

f6)f2.45 :: 10") (1.07 :: 10'3)

= 2.25 x 105 p,yf,2,,,,

=

(2)(jQ (3.7)

(30.0) 5 (2.25 x 10 )

(2 x 10 )

D

=

452 cr/hr

=

In the case of a point tucnty feet frc= the wall, the car.e treatcent con be used if ground shiciding is assuned to be negligible. This, of course, introducas an overcstircts of dose rate; he' co, it is a n

I conse:vative estinato. The doso 10 thus calculated to be,

.g D = (2 :: 10'01 (6) (2.46 :: 10") (9 x 10'12) = 240 mr/hr.

3_

(2)

( rr )

(57)

(30.8)

For a distance greater than 100 fact from the building, only half the line source can be "seen" cnd the flux can thus be o

expressed,

~

3 8 /2 P (8,b)

= 535 5

=

4 iT c' i"-

(2x10%

(585) = 1.07 x 10~3 D

=

Thus, the dose ratos as points off the em:r. sus are less then the maximum permissible for the general population. The dose rato is cufficiently low at the point of closest approach to the building to permit roping of or otherwise restricting approach to the building before serious exposure cca be accumulated.

a-3.

conclusion:

i It has been shown thac tk: consequcaces of a us::imum credibic accident can be controlled so that exposure of the general public will be below merhen permissible and that of the operating

, s-staff vill not be excessive; consequently, the reactor can be operated safely.

s-d T 'J

-104-

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