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Potential Radiological Consequences of Accidental Nuclear Criticality in Exxon Nuclear UO2 Plant
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a XN-NF-486 ISSUE DATE: 12/05/78 1

1 1

POTENTIAL RADIOLOGICAL CONSEQUENCES OF ACCIDENTAL NUCLEAR CRITICALITY IN THE EXXON NUCLEAR UO PLANT 2

1 1

November 1978 l

12 ass I

t XN-NF-486 POTENTIAL RADIOLOGICAL CONSEQUENCES OF ACCIDENTAL NUCLEAR CRITICALITY IN THE EXXON NUCLEAR UO PLANT 2

Prepared by:

AI'I'

^

"NN/5 C. O. Brown Date Licensing Engineer m

Accepted by: 'k

/1,_</JIrv

//

7[

W. S. Nechodom, Manager Date Licensing and Compliance ht 7

}

Approved by:

[ [lj f%

R.Idikson Date Manager, Licensing

=

XN-NF-486 POTENTIAL RADIOLOGICAL C0t4SEQUEf4CES OF ACCIEENTAL NUCLEAR CRITICALITY IN THE t

EXXON NUCLEAR UO PLANT 2

Distribution C. O. Brown (5)

D. E. Clark D. L. Cornell H. P. Estey L. E. Hansen R. L. Miles W. S. Nechodom R. Nilson T. C. Probasco R. H. Purcell R. H. Schutt M. L. Smith R. B. Stephenson Document Control (10)

XN-NF-486 TABLE OF CONTENTS Page LIST OF TABLEn.

11 LIST OF FIGURES iii PREFACE iv INTRODUCTION.....................

1 ENC MODEL CRITICALITY ACCIDENT.

1 FISSION-PRODUCT SOURCE INVENTORY.

4 PROMPT GAMMA AND NEUTRON DOSE CALCULATIONS.

4 FISSION-PRODUCT PLUME TRANSPORT AND DISPERSION...

7 FISSION-PRODUCT PLUME EXTERNAL DOSE CALCULATIONS..

11 TOTAL INTEGRATED EXTERNAL DOSE CALCULATIONS 12 CRITICAL ORGAN DOSE CALCULATIONS...............

19 CONCLUSIONS 19 APPENDIX A.....

A-1

.j-

XN-NF-486 LIST OF TABLES Table No.

Page I

UO Plant Vessel Descriptions and Dimensions..

3 2

II Radioactivity of Important Nuclides Released by a Single Fission Burst of 1018 Fissior,s....

5 III Radioactive Decay of the Fission-Product Plume..

6 IV Prompt Gamma and Neutron Doses to an Individual as a Function of Distance from the Source 8

V Short Term Average Centerline Values of the Ground Level Concentration.........

11 VI Beta and Gamma Dose Rate Calculation Results 12 VII Whole Body, Genetic, and Skin Dose Calculation Results 13 VIII Total External Dose Summary 14 IX Total Whole Body Dose vs Time 17 X

Critical Organ Dose Calculation Results 20 XI Total Two-Hour Whole Body Exterral and Thyroid Dose Summary.

21 A-I Average Rate of Radioactive Material Released per Fission Burst A-1 A-II Effective Decay Energy A-3

-ii-

XN-NF-486 LIST OF FIGURES Figure No.

Page 1

Total Whole Body External Dose vs Distance from Release (Twelve Inches Concreto) 15 2

Total Whole Body External Dose vs Distance from Release (No Concrete).....

16 3

Total Whole B^dy External Dose vs Time....

18 w

- iii -

i XN-NF-486 PREFACE This report summarizes the results of an assessment of the potential radiological consequences of an accidental nuclear criticality accident in the Exxon Nuclear UO Plant. Guidelines for this analysis are as 2

described in USNRC Regulatory Guide 3.34, " Assumptions Used For Evaluating the Potential Radiological Consequences of Accidental Nuclear Criticality In A Uranium Fuel Fabrication Plant," April 1977.

Throughout the text of this report references are made to comments found in Appendix A.

These comments provide supportive information relative to assumed conditions, calculational methods, etc. used in that section.

- iv -

XN-NF-486 POTENTIAL RADIOLOGICAL CONSEQUENCES OF ACCIDENTAL NUCLEAR CRITICALITY IN THE EXXON NUCLEAR UO PLANT 2

INTRODUCTION The Exxon Nuclear UO Plant located at Richland, Washington is designed 2

to manufacture fuel assemblies for light water reactors.

The plant is 235 ) UF, which is converted to licensed to receive low-enriched (< 5 wt.%

U 6

UO2 p wder, pressed into pellets and loaded into metal rods.

Finished fuel rods are then bundled, temporarily stored, and subsequently shipped to respective reactor sites.

This report summarizes the results of an assess-ment of~the risk to personnel health and safety of a postulated nuclear criticality accident within the facility and the extent of the acute radio-

}

logical consequences of such an event to personnel both on and offsite.

The guidelines provided in USNRC Regulatory Guide 3.34, " Assumptions used for Evaluating the Potential Radiological Consequences of Accidental Nuclear Criticality in a Uranium Fuel Fabrication Plant," April 1977, were used to derive the results in this report.

However, certain of the assump-tions in the Regulatory Guide, specifically those dealing with release fractions of the noble gases and iodines, are unduly conservative.

Hence,

~

the results presented must be used with caution when planning the emergency responses to a detected accidental criticality.

ENC MODEL CRITICALITY ACCIDENT One of the major safety considerations in the design of a fuel fabrication plant is the potential for inadvertant criticality.

To date, four nuclear excursions have occurred in fuel fabrication and scrap recovery facilities in the United States.II)

The most recent occurred in 1964. All of these involved fissile solutions accumulating in " unsafe"

~

containers, and none of the events produced significant environmental consequences.( )

XN-NF-486 With these facts in mind, a review of the Exxon Nuclear UO Plant was 2

made to determine those areas within the facility that, given the right set of circumstances, could lead to a possible criticality accident.

The UF ~

6 UO Conversion Area, where wet chemistry processing takes place, was deemed 2

to be the most credible location for such an event.

The total fission rate, being a strong function of fissile material volume, suggested a review of those vessels in the area containing large volumes of fissile material.

Table I lists the three largest vessels in the area and gives a description of each.

In reviewing these vessels, it is noted that the dryer and calciner represent " safe geometries" for fissile material at optimum conditions associated with each vessel.

Tank-163, however, represents an " unsafe geometry" should a uranium concentration violation occur.

(Tank-163 is a series of five waste hold tanks located in the Waste Tank Gallery between Lines 1 and 2.) The volume of a single tank in the series is used to estimate the total number of fissions that could occur in an eight-hour period. As discussed by Olsen, et al.(3), the total number of fissions produced by solution-type excursions is both a function of the volume of fissile material present and the duration of the excursion.

F F

F T

B+

P T

=

total fissions

=

F B

number of fissions produced

=

in initial burst F p number of fissions during plateau

=

period For a 438 liter container the total number of fissioris over an eight-hour period is calculated (see C0" MENT 1, Appendix A) to be:

17 F

4.33 x 10 fissions for V

= 438 t

=

B B

IO F

2.51 x 10 fissions for t = 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months

=

p 18 F

2.94 x 10 fissions

=

T _.

M M

M e

TABLE I UO PLANT VESSEL DESCRIPTI0f45 AND DIMENSIONS 2

Fissile Material Maximum Uranium Vessel Description Dimensions, cm Volume, t Form Concentration, g/t Tk-163 Vertical Cylinders d = 30.23 438 U(4)0, + H O 350 2

(Line 2) h = 610.0 Calciner Horizontal Cylinder d = 25.4 417 ADU(5) + H 0*

850 2

(Line 1)

L = 823.0 w

Dryer Horizontal Cuboid w = 36.20 662 ADU(4) + H O 850 2

(Line 2) h = 37.47 L = 488.0

Assumed worst credible accident condition; ADU - ammonium diuranate

$i e

XN-NF-486 The above results indicate that for the vessel evaluated, the maximum 18 total number of fissions occurring in an eight-hour period is 2.94 x 10 fissions.

FISSION-PRODUCT SOURCE INVENTORY For the purposes of this analysis it is assumed that the total fissions during the plateau period occur at ten minute intervals over an eight-hour period, i.e., 47 additional bursts occur after the initial burst. At this rate each secondary burst would account for an 16 average of 5.34 x 10 fissions.

Fission-product inventory data based on ORIGEN(4) computer code calculations were used to estimate the Xe, Kr, and I radionuclide sources produced from the excursion.

Twenty-five percent of the iodine radionuclides and all of the noble gas fission-products escape to the ventilated room atmosphere.(5) As summarized in Table II, the maximum quantities of fission-products produced over a 18 twenty-four hour period are listed for a single fission burst of 10 fissions. Table III lists total curie amounts and release rates assumed as source terms for dose rate calculations. As explained in COMMENT 2 of Appendix A, the dose rate source terms are conservatively based on the maximum fission-product amounts released.

PROMPT GAMMA AND NEUTRON DOSE CALCULATIONS The dose to an individual from the prompt gamma and neutron radiation resulting from the ENC model criticality accident was examined as a function of distance from the facility.

For these calculations, the semi-empirical equations (b) used are as follows:

Prompt Gamma Dose D = 2.1 (10-20) Nd-2 exp (-3.4d)

Where:

Y D = gamma dose (rem)

Y N = number of fissions d = distance from source (km)

XN-NF-486 TABLE II RADI0 ACTIVITY OF IMPORTANT NUCLIDES RELEASED 18 BY A SINGLE FISSION BURST OF 10 FISSIONS (Ci)

(Origen(6) Calculation)

Time After Maximum Quantity (a)

Burst of Maximum Nuclide Maximum Quantity Quantity Escaping Building 83m 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months 4.43-00 4.43-00 Kr 85m 20 minutes 1.39+01 1.39+01 gp 85 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months 1.85-04 1.05-04 Kr 87 5 minutes 9.64+01 9.64+01 Kr 88 2 minutes 6.54+01 6.54+01 Kr 89 1 minute 3.68+03 3.68+03 Kr 131m 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months 2.92-03 2.92-03 Xe 133m 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months 6.86-02 6.86-02 Xe 133 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months 1.32-00 1.32-00 Xe 135m 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months 1.22+01 1.22+01 Xe 135 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months 1.48+01 1.48+01 Xe 137 1 minute 3.58+03 3.58+03 Xe 138 2 minutes 1.02+03 1.02+03 Xe 129 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months 2.84-10 7.10-11 y

131 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months 6.67-01 1.67-01 1

132 0

3.16-00 7.90-01 1

133 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months 1.43+01 3.58-00 y

134 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months 1.90 -02 4.75+01 y

135 5 minutes 4.68+01 1.17+01 7

TOTAL 8.74+03 8.55+03 (a) 25% of the iodine radionuclides resulting from the excursion are released to the room atmosphere. -

umme mums usum umms uuss usur' un ' ' auer ' ummer umur ammu ummus ' umme - unus ' - ~ umuur - umuur -' imus-TABLE III RADI0 ACTIVE DECAY OF THE FISSION-PRODUCT PLUME Fission-Product Inventory (b), Ci After(c)

After After After After Number of Fissions No Decay _

3 minutes 5 minutes 7 minutes 9 minutes 17 minutes 4.33 x 10I7 (Initial Burst) 3.70+03 2.24+03 1.65+03 1.27+03 9.90+02 4.46+02 5.34 x 1016 (Initial Burst) 4.57+02 2.77+02 2.03+02 1.56+02 1.25+02 5.50+01 i

cn l

(b) Assumes no decay of iodine radionuclides ( }.

(c) Decay times representing the approximate distances of the fission-product plume frcm the point of release, per assumed meteorological conditions, are as follows:

Distance, m Decay Time, min.

100, 150 3

200, 230 5

370 7

500 9

x 1000 17 A8

XN-NF-486 Prompt Neutron Dose D = 7.0 (10-20) Nd-exp (-5.2d)

Where:

D = neutron dose (rem) n N = number of fissions d = distance from source (km)

Table IV summarizes the results of calculated prompt gamma and neutron doses as functicns of distance from the source.

Dose estimates are given for radiation attenuation through six inches and twelve inches of concrete to account for dose reduction, when applicable, through facility walls.

Gamma and neutron radiation reduction factors of 5.0 and 4.6, respectively, for the first twelve inches (5) of concrete and 2.0 and 1.8, respectively, for the first six inches (6) were used.

FISSION-PRODUCT PLUME TRANSPORT AND DISPERSION In order to evaluate external and internal exposure to personnel located outside the UO facility, an assessment of tne excursion fission-2 product yield, migration, and dispersion was made.

In addition, factors relevant to the operation of the Exxon Nuclear plant specifically were included in the evaluation.

For the ENC model nuclear excursion, it is assumet that all of the noble gases and 25% of the iodine radionuclides are released directly to the room atmosphere of the Waste Tank Gallery.

The air exchange rate of the gallery is fifteen to thirty air changes per hour, and exhaust air released to the outside atmosphere passes out of the 11.0 meter high K-33 stack. At this ventilation rate, the room is vented through the exhaust stack on an average of once every three minutes. Hence, fission-products released during an excursion would be exhausted from the callery in not longer than three minutes.

The noble gas and iodine release fractions given are extremely con-servative.

These release fractions were originally used in early guidelines by the industry and the Atomic Energy Commission to evaluate the adequacy

XN-NF-486

\\

TABLE IV

^

PROMPT GAMMA AND NEUTRON DOSES (REM) TO AN INDIVIDUAL AS A FUNCTION OF DISTANCE FROM THE SOURCE Total Dose Through Concrete Distance, m Gamma Dose Neutron Dose None 6 Inches 12 Inches 17 FOR INITIAL BURST (4.33 x 10 Fissions) 100 6.47-01 1.80-00 2.45-00 1.32-00 5.21-01 150 2.43-01 6.18-01 8.61-01 4.65-01 1.83-01 200 1.15-01 2.68-01 3.83-01 2.06-01 8.13-02 230 7.86-02 1.73-01 2.52-01 1.35-01 5.33-02 370 1.89-02 3.23-02 5.12-02 2.74-02 1.08-02 500 6.64-03 9.00-03 1.56-02 7.81-03 3.28-03 1000 3.03-04 1.67-04 4.70-04 2.45-04 9.69-05 16 FOR SUBSEQUENT BURST (5.34 x 10 Fissions) 100 7.98-02 2.22-01 3.02-01 1.63-01 6.42-02 150 2.99-02 7.62-02 1.06-01 5.73-02 2.25-02 200 1.42-02 3.30-02 4.72-02 2.54-02 1.00-02 230 9.70-03 2.14-02 3.11-02 1.67-02 6.59-03 370 2.33-03 3.99-03 6.32-03 3.38-03 1.33-03 500 8.19-04 1.11-03 1.93-03 9.64-04 4.05-04 1000 3.74-05 2.06-05 5.80-05 3.02-05 1.20-05

XN-NF-486 I

of reactor site boundaries.

To approach these release fractions, it was necessary to assume that the reactor core was molten and that both the reactor pressure boundary and the reactor containment vessel were breached, thus exposing the molten core directly to the environs.

These assumptions do not apply to the slightly-enriched uranium fuel fabrication plant situation because:

e The criticality would have to occur in a uranium-bearing solution.

17 e

The prompt burst of about 4 x 10 fissions would only heat I

the solution, even assuming a limited volume of a cubic foot, a few tens of degrees farenheit--far short of the boiling point, much less the melting point of any uranium compound.

The plateau period would maintain the solution at an elevated temperature, but short of boiling, until evaporation or corrective action stopped the reaction.

I e

The total volume of noble gas fission products created by the burst would be less than 0.1 cc (STP), which is several I

orders of magnitude less than the solubility of the noble gas in such a solution volume (on the order of a few hundred cc of gas).

Iodine is similarly soluble.

In view of these facts, it is important that the reaction to a criticality accident not be based on the hypothesized release of gaseous fission products, but rather on the direct radiation dose.

As will be shown below, discounting the effects of an assumed plume reduces the cal-culated dose by an order of magnitude, and would certainly minimize the need for hasty evacuation beyona the existing staging areas.

For example, the initial burst contributes a direct radiation dose of.183 rem at the UO staging area.

The total direct radiation dose for the first two hours 2

would be.431 rem at the UO staging area.

Everyone who evacuates to the 2

U0 staging area would have initially been closer to the initial burst, 2

I g

-e-

XN-NF-486 I

and would therefore have received an initial dose of 0.183 rem or greater.

Remaining at the staging area would, at most, roughly double the initial dose from direct radiation.

If, however, a dose from a plume release is considered, the whole body dose at the UO staging area would be 6.46 rem, I

2 clearly calling for a different response and making the UO staging area 2

location inappropriate.

It is conservatively assumed that all airborne fission-products released to the outside environment are released at ground level.

The atmospheric diffusion model for the release is based on a Pasquill Type F I

meteorological condition at a windspeed of 1 meter /sec.( )

Thus, the diffusion equation (5) to calculate the ground level concentration as a function of distance for this model release is given as:

Oi

x. =

where: x.=

the short-term average I

" " "y I

I z

centerline value of the ground level concentration (Ci/m )

Q.=

average rate of material I

released (Ci/sec) p windspeed (m/sec)

=

I horizontal standard deviation

=

o Y

of plume (m) vertical standard deviation o =

z of lume (m)

I Table V gives the results of x-value calculations as a function of distance from the point of release.

These values assume an average fission-product inventory as described in COMMENT 2 of Appendix A and summarized in I

Table III. Building wake factors, horizontal and vertical standard deviations of the plume used to calculate x-values are as discussed in COMMENT 3 of Appendix A.

I I

I

XN-NF-486 TABLE V SHORT TERM AVERAGE CENTERLINE VALUES OF THE 3

GROUND LEVEL CONCENTRATION (Ci/m )

(Building Wake Factors Included) 3 x, Ci/m Distance, m Initial Burst Subsequent Burst 100 4.50-02 5.57-03 150 1.62-02 2.00-03 200 8.79-03 1.08-03 230 6.08-03 7.47-03 370 3.30-03 4.06-03 500 1.82-03 2.29-04 1000 4.04-04 5.00-05 FISSION-PRODUCT PLUME EXTERNAL DOSE CALCULATIONS With the results from Table V, beta and gamma dose rate calculations (5) were performed. For beta radiation the receptor is assumed to be exposed to an infinite cloud at the maximum ground level concentration.

For gamma a

radiation the dose is calculated for a semi-infinite cloud owirg to the presence of the ground.

Under these conditions, the rate of energy absorption per unit volume is equal to the rate of energy released per unit volume.

The average beta and gamma energies per disintegration are assumed to be 1.41 and 1.20 Mev respectively (see COMMENT 4, Appendix A).

A summary of the dose rate calculations for a uniform cloud containing x curies of radioactivity per cubic meter as a function of distance is given in Table VI (see COMMENT 5, Appendix A).

XN-NF-486 TABLE VI BETA AND GAMMA DOSE RATE CALCULATION RESULTS B =, rad /sec D', rad /sec in Air Distance, m in Air Surface Y"

I7 FOR INITIAL BURST (4.33 x 10 Fissions) 100 2.90-02 1.46-02 1.35-02 150 1.04-02 5.25-03 4.86-03 200 5.66-03 2.85-03 2.64-03 230 3.92-03 1.97-03 1.82-03 370 2.13-03 1.07-03 9.09-04 500 1.17-03 5.89-04 5.46-04 1000 2.60-04 1.31-04 1.21-04 16 FOR SUBSEQUENT BURST (5.34 x 10 Fissions) 100 3.59-03 1.81-03 1.67-03 150 1.29-03 6.49-04 6.00-04 200 6.96-04 3.50-04 3.24-04 230 4.81-04 2.42-04 2.24-04 370 2.62-04 1.32-04 1.22-04 500 1.48-04 7.43-05 6.87-05 1000 3.22-05 1.62-05 1.50-05 hhole body and genetic dose calculations were performed assuming a gamma radiation dose at a depth of 5 centimeters and 1 centimeter respec-tively (see COMMENT 6, Appendix A).

The skin dose was calculated as the sum (5) of the surface gamma dose and beta dose at a depth of 7 mg/cm Results of these calculations are summarized in Table VII.

TOTAL INTEGRATED EXTERNAL DOSE CALCULATIONS The total dose is the summation of the dose received due to both direct radiation and the fission-product plume (see COMMENT 7, Appendix A).

Table VIII summarizes +% results of the expected total external dose, both whole body and skin, to an individual as a function of distance (100-1000 meters) from the facility over a two-hour period.

Figure 1 graphically depicts the expected total whole body external dose to an individual located _ _ _ _ _.

XN-NF-486 at the UO Staging Area (150 meters), M0 Staging Area (200 meters), or the 2

CTF (370 meters).

Figure 2 shows graphically the expected total whole body dose at the nearest site boundary (230 meters).

In addition, the expected whole body dose to an individual as a function of time for distances of 100, 150, 200, 230 and 500 meters has been estimated and these results are summarized in Table IX and Figure 3.

TABLE VII WHOLE BODY, GENETIC, AND SKIN DOSE CALCULATION RESULTS (Rad /sec)

Distance _1_m WB Genetic Skin 17 FOR INITIAL BURST (4.33 x 10 Fissions) 100 1.18-02 1.31-02 2.78-02 150 4.26-03 4.73-03 1.00-02 200 2.31-03 2.57-03 5.43-03 230 1.59-03 1.77-03 3.75-03 370 7.96-04 8.85-04 1.96-03 500 4.78-04 5.32-04 1.12-03 I

1000 1.06-04 1.18-04 2.49-04 16 FOR SUBSEQUENT BURST (5.34 x 10 Fissions) 100 1.46-03 1.69-03 3.50-03 150 5.26-04 5.84-04 1.24-03 200 2.84-04 3.16-04 6.67-04 230 1.96-04 2.18-04 4.61-04 370 1.07-04 1.19-04 3.59-04 500 6.02-05 6.69-05 1.42-04 1000 1.31.05 1.46-05 3.09-05 I _ _.... _

XN-NF-486 TABLE VIII TOTAL DOSE

SUMMARY

(REM)

Assumed Conditions:

17 1)

Initial Fission Burst of 4.33 x 10 Fissions 2)

SubsequentFissionBursts(PlggeauPeriod)

No. of Fissions:

5.34 x 10 Occurrence: Once every ten minutes after initial burst 3)

Total Dose Calculated over a 2-Hour Time Period 4)

Fission Product Inventorj: As described in text and Table II Direct Radiation Plume Total Dose Distance, m (Dy + Dn)

Whole Body Skin Whole Body Skin FOR NO CONCRETE SHIELDING 100 5.77-00 1.67+01 3.98+01 2.25+01 4.56+01 150 2.03-00 6.03-00 1.42+01 8.06-00 1.62+01 200 9.02-01 3.26-00 7.66-00 4.16-00 8.56-00 230 5.94-01 2.25-00 5.29-00 2.84-00 5.88-00 370 1.21-01 1.18-00 3.55-00 1.30-00 3.67-00 500 3.68-02 6.84-01 1.61-00 7.21-01 1.65-00 1000 1.11-03 1.50-01 3.53-01 1.51-01 3.54-01 FOR 6 INCHES CONCRETE SHIELDING 100 3.11-00 1.67+01 3.98+01 1.98+01 4.29+01 150 1.10-00 6.03-00 1.42+01 7.13-00 1.53+01 200 4.85-01 3.26-00 7.66-00 3.75-00 8.15-00 230 3.19-01 2.25-00 5.29-00 2.57-00 R.61-00 370 6.46-02 1.18-00 3.55-00 1.24-00 3.61-00 500 1.84-02 6.84-01 1.61-00 7.02-01 1.63-01 1000 5.77-04 1.50-01 3.53-01 1.51 -01 3.54-01 FOR 12 INCHES CONCRETE SHIELDING 100 6.77-01 1.67+01 3.98+01 1.74+01 4.05+01 150 4.31-01 6.03-00 1.42+01 6.46-00 1.46+01 200 1.91-01 3.26-00 7.66-00 3.45-00 7.85-00 230 1.26-01 2.25-00 5.29-00 2.38-00 5.42-00 370 2.54-02 1.18-00 3.55-00 1.21-00 3.58-00 500 7.74-03 6.84-01 1.61-00 6.92-01 1.62-00 1000 2.29-04 1.50-01 3.53-01 1.50-01 3.53-01

XN-f&-486 FIGURE 1 100.0 70 Total Whole Body External Dose (Rem) vs Distance from Release (m) 40 2 Hour Time Period

- 12 Inches Concrete UO Staging Area 2

M0 Staging Area 10.0 7.0 WB Plume 4.0

- WB (Total)

CTF 2.0 1.00

^

0.7 Total WB 0.4 External

- Dose, Rem 0.2 0.10

.07

.04 WB (Direct)

.02 0.01

~

.007 6

.004

.002 0.00i l

I I

I I!

100 200 500 1000 Distance, m XN-NF-486 FIGURE 2 100.0 Total Whole Body External Dose (Rem) 70 vs Distance from Release (m) 40 --

- 2 Hour Time Period

- No Concrete 20 -

WB (Plume) 10.0 7.0 Site Boundary 4.0 2.0 WB (Total) l.00 0.7 Total WB 0.4 External

Dose, Rem 0.2 0.10 a

.07 WB (Direct)

.04

\\c

\\

.02 0.01

.007

.004

.002 0.001 I

I 100 200 500 1000 Distance, m

M XN-NF-486 TABLE IX TOTAL WHOLE BODY EXTERNAL DOSE (REM) VS TIME (HOURS)

Total Whole Body

Time, Direct Radiation, Rem

Plume, Dose, Rem Distance, m Hours 12" Concrete No Concrete Rem 12" CCT No CC_T 150 (U0 0.5 2.28-01 1.07-00 3.19-00 3.42-00 4.26-00 StagibgArea) 1.0 2.96-01 1.39-00 4.13-00 4.43-00 5.52-C0 2.0 4.31-01 2.03-00 6.03-00 6.46-00 8.06-00 4.0 7.01-01 3.30-00 9.81-00 1.05+01

1. 31 +01 6.0 9.71-01 4.57-00 1.36+01 1.46+01 1.82+01 8.0 1.24-00 5.84-00 1.74+01 1.86+01 2.32+01 200 (M0 0.5 1.01-01 4.77-01 1.73-00 1.83-00 2.21-00 Staging Area) 1.0 1.31-01 6.19-01 2.24-00 2.37-00 2.86-00 2.0 1.91-01 9.02-01 3.26-00 3.45-00 4.16-00 4.0 3.11-01 1.47-00 5.31-00 5.64-00 6.78-00 6.0 4.31-01 2.04-00 7.35-00 7.78'Q0 9.39-00 8.0 5.51-01 2.60-00 9.39-00 9.94-00 1.20+01 230 (Site 0.5 6.65-02 3.14-01 1.19-00 1.26-00 1.50-00 Boundary) 1.0 8.63-02 4.08-01 1.54-00 1.63-00 1.95-00

_~

2.0 1.26-01 5.94-01 2.25-00 2.38-00 2.84-00 4.0 2.05-01 9.67-01 3.66-00 3.87-00 4.63-00 6.0 2.84-01 1.24-00 5.07-00 5.35-00 6.41-00 8.0 3.63-01 1.71-00 6.48-00 6.84-00 8.19-00 370 (CTF) 0.5 1.35-02 6.38-02 6.06-01 6.20-01 6.70-01 1.0 1.75-02 8.28-02 7.99-01 8.17-01 8.82-01 2.0 2.54-02 1.21-01 1.18-00 1.21-00 1.30-00 4.0 4.14-02 1.97-01 1.95-00 1.99-00 2.15-00 6.0 5.74-02 2.72-01 2.72-00 2.78-00 2.99-00 8.0 7.33-02 3.48-01 3.50-00 3.57-00 3.85-00.

XN-NF-486 FIGURE 3 50.0 Total Whole Body External Dose (Rem) 40.0 ys Time (Hours) 20.0 150 m (12" CCT) ptA g3 gG\\

o s,

200 m (12" CCT) 230 m (No CCT) g 9

7.0 go

, - go'50*

, - Es6 4.0 370 m (12" CCT) cd i

2.0

/

1.00

.70 Total WB

.40 External

Dose, Rem

.20 0.10

.07

.04 NOTE:

The dose estimate at the site boundary assumes no prompt gamma and neutron radiation attenuation through concrete (CCT).

.02 0.01 0

2'. 0 4.b 6l. 0 8l.0 10'.0 Time, Hours

XN-NF-486 CRITICAL ORGAN DOSE CALCULATIONS The " Critical Organ" dose from inhaled radioactive materials was also estimated.

In particular, the dose to a man's thyroid due to the presence of iodine radionuclides in the airborne fission-product plume was calculated.

-4 The breathing rate of the man was assumed to be 3.47 x 10 cubic meters persecond(5), and the thyroid dose was calculated as a function of discar.ce from the facility over a two-hour period (see COMMENT 7, Appendix A).

Results of these calculations are given in Table X.

CONCLUSIONS This report has conservatively estimated doses and dose rates versus I

both time and distance from the point of release for a postulated criticality accident occurring in the Exxon Nuclear UO facili ty.

Atmospheric conditions, 2

fission-product source terms, diffusion and dose rate equations, etc. used in this analysis were as defined per USNRC guidelines (5)

Calculations were performed for, but not limited to specific locations near the UO2 pl nt, i.e.,

UO and M0 staging areas, nearest site boundary, and the CTF. As a matter 2

of convenience, whole body external and thyroid dose results estimated at these locations over a two-hour period are summarized in Table XI.

I I

I

>M r '

XN-NF-486 TABLE X CRITICAL ORGAN DOSE CALCULATION RESULTS Thyroid Dose (Rem) vs Distance from Release (m)

Assumed Conditions:

1)

Thyroid dose to an adult from plume inhalation over a two-hour period.

2)

Inhalation rate - 3.47-04 m /sec(5) 3 3)

Fission-product inventory assumed to consist of maximum released quantities (no decay) of 131I,1331,134I, and 1351, 4)

Dose rate conversion factors as outlined in reference (7).

Thyroid Dose, Rem From Plume Inhalation Dose 31 Distance, m 7

133; 134 135 7

I TOTAL 100 1.06-00 6.08-00 5.08-00 6.20-00 1.84+01 150 4.60-01 2.66-00 2.22-00 2.71-00 8.05-00 200 3.10-01 1.78-00 1.49-00 1.82-00 5.40-00 230 1.94-01 1.11-00 9.31-01 1.14-00 3.38-00 370 1.37-01 7.88-01 6.58-01 8.04-01 2.39-00 500 5.86-02 3.37-01 2.82-01 3.44-01 1.02-00 1000 4.76-02 2.74-01 2.29-01 2.80-01 8.31 -01 _ _. _ _ _..

XN-NF-486 I

TABLE XI TOTAL TWO-HOUR WHOLE BODY EXTERNAL D THYROID DOSE (REM)

SUMMARY

Dose, Rem Distance, m Whole Body Thyroid 150 (UO Staging Area) 6.46-00 8.05-00 200 (MO Staging Area) 3.45-00 5.40-00 230 (Site Bourdary) 2.84-00 3.38-00 370 (CTF) 1.21-00 2.39-00 I

I I

I

-i---um.

XN-NF-486 APPENDIX A M

m__-..

XN-NF-486 COMMENT 1 The total number of fissions was calculated as follows:(3) jg 15.47 + 0.82 log VBf p

=

B V = 438 t B

17 4.33 x 10 fissions

=

3.2 x 1018 (1 - t-0.l')

t = 28,800 sec.

F

=

p (8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months )(5) 18 2.51 x 10 fissions

=

18 F

FB+Fp = 2.94 x 10 fissions

=

T COMMENT 2 Table II summarizes the time-dependent maximum radioactivity over a twenty-four burst of 10 gour period of important nuclides released by a single fission fissions. The total amount of material, A.

released per fission burst was calculated as the sum of the maximum yle,ld of each radio-nuclide, reduced to reflect fission burst size and radioactive plume in-transit decay.

The calculated average rate of radioactive material released, Q, is therefore conservatively high and results are summarized inTableAibelow:

TABLE A-1 AVERAGE RATE OF RADI0 ACTIVE MATERIAL RELEASED (Ci/Sec) PER FISSION BURST Noble Gas (a)

A (b), C1 Q (c), Ci/sec Decay Time, 5

9 Distance, m min.

Tnitial Subsequent Initiri Subsequent 100, 150 3

2.24+03 2.77+02 3.73-00 4.62-01 200, 230 5

1.65+03 2.03+02 2.75-00 3.38-01 370 7

1.27+02 1.56+02 2.12-00 2.60-01 500 9

9.90+02 1.25+02 1.65-00 2.08-01 1000 17 4.46+02 5.50+01 7.43-01 9.20-02 (a) The noble gas radionuclides decay as the airborne plume migrates from the point of release.

(b) A is the total decayed amount of radioactive material released per 4f1ssion burst.

(c) Q is the average rate of radioactive material released over a ten-minute 4

interval.

A-1

---i-misses

XN-NF-486 COMMENT 3 ghort term average centerline values of the ground level concentration (5)

(Ci/m ) were calculated as follows:

Ni

=

xj y py a where, for Pasquill Type F meteorological conditions, the horizontal (o

and vertical redu)ction factors (k)z()b), and average windspeed (u) are given as: standard d Distance, m

_a,m

-z, m b

u, m/sec 100 4

2.2 3

1 150 7

3.5 3

1 200 8.3 4

3 1

230 10 4.8 3

1 370 15 6.8 2

1 500 20 8.5 1.7 1

1000 38 14 1.1 1

Values of Q assumed for the x-value calculations are as given in 5

COMMENT 2.

COMMENT 4 The average beta and gamma decay energies were estimated to be 1.41 and 1.20 Mev per disintegration 2ccording to the simple weighting relation-ship:

(A )(E )

j j

I E (ave.)

=

^i i

where A and E are the radionuclide source activities and the q

3 effective average energies (beta or gamma), respectively,for radionuclides listed in Table II.

Table A-II below lists effective decay energies (9) for nuclides of interest.

4 A-2

XN-NF-486 TABLE A-II EFFECTIVE DECAY ENERGY (9)

(Mev/ disc.)

Nuclide f;

E 83m 0

0.009 Kr 85m 0.22 0.15 Kr 85 0.30 0.002 Kr 87 1.40 1.50 Kr 88 0.39 0.84 Kr 89 1.40 2.10 Kr 131m 0

0.018 Xe 133m 0

0.047 Xe 133 0.098 0.046 Xe 135m 0

0.42 Xe 135 0.30 0.25 Xe 137 1.60 0.18 Xe 138 0.96 1.50 Xe 129 0.039 0.024 7

1 31 0.18 0.38 7

132 0.46 2.60 7

133 0.44 0.48 7

134 0.66 2.50 7

135 0.32 1.80 7

COMMENT 5 Beta and gamma dose rate calculations (5) made as folicws:

B;

0. 6 E x Beta dose in air (rad /sec)

D

=

g

[

g; 0.23 E x Beta surface dose (rad /sec)

D

=

g D'

0.25 E x Gamma dose in air (rad /sec)

=

Y" Y

(semi-infinite Cloud)

where, E

1.41 Mev/ disc

=

g E

1.20 Mev/ disc

=

Y The fission product inventory decay energy spectrum is dominated by 89Kr, 137Xe, and 13eXe (see COMMENT 4).

A-3

XN-NF-486 COMMENT 6 Whole body and genetic dose estimations from the plume were calculated assuming a gamma radiation dose at a depth of 5 cm and 1 cm respectively.

Hence,the99mn)adoseisattenuatedbythetransmissionfactor(T)givenby the equation \\b:

-mtp T

= e

where, T

transmission factor

=

density of tissue (10),

=

0.876 for 5 cm p

T

=

0.974 for 1 cm thickness, cm t

=

mass attenuation coefficient -

m =

2 50) 0.0288 cm /gm @ 1.25 Mev Skin dose estimates were calculated as the sum of the gamma dose rate attenuation factorg5qurface dose (corrected by D /D in air and the betp

- an energy-dependent d B

,),

COMMENT 7 At a distance of 100 meters, the total unattenuated external dose over the initial two-hour time period is calculated as follows:

For the total whole body external dose -

From Direct Radiation:

=

(1) (2.45) + (11)(0.301) = 5.77 Rem From the Plume:

1.18 x 10-2 Rem /sec x 600 sec = 7.08 Rem (from initial burst)

+ 5.77 x 10- Rem /sec x 6000 sec = 9.64 Rem (from subseq. bursts)

= 16.7 Rem Total WB Dose:

5.77 + 16.7 = 22.5 Rem NOTE: Dose estimates summarized in Table IX were calculated in a similar manner.

A-4 4

XN-NF-486 COMMENT 8 The inhaled thyroid dose estimates were calculated for maximum iodine buildup in the fission and decay product plume over a two-hour period according to the following equation (8):

f)(k)f(At+e

-A t b - 1) where, D

=

j a

b b

D = dose to organ (rem) j f = fractional uptake by organ a

P = rate of intake (pCi per unit time)

A = biological half-life (time unit) b t = time following initial intake k = dose conversion factor (rem per pCi)

For the purposes of this report, no iodine decay was assumed; hence, the equatien reduces to, D = (f )(k)(P)(t) j a

where

= 0.23 for iodine (8) fa k = 6.45-00 for 1311 + 131mXe from reference (8) 1.73-00 for 1331 + drs.

1.09-01 for 134I 5.37-01 for 13sl + drs.

(P)(t) calculated as the total pCi inhaled over a two-hour time period due to both the initial and subsequent burst fission and decay product plumes.

A-5

XN-NF-486 6.

i REFERENCES I

1.

W. R. Stratton, " Review of Criticality Incidents," LA-3611, Los Alamos Scientific Laboratory (January 1967).

2.

J. M. Selby, et al., " Considerations in the Assessment of the I

Consequences of Effluents from Mixed 0xide Fuel Fabrication Plants,"

BNWL-1697, Rev. 1, Battelle Pacific Northwest Laboratories (1975).

I 3.

A. R. Olsen, R. L. Hooper, V. O. Uotinen, C. L. Brown, " Empirical Model to Estimate Energy Release from Accidental Criticality," ANS Trans., Vol. 19, p. 189-91 (1960).

4.

M. J. Bell, "0RIGEN - The ORNL Isotope Generation and Depletion I

Code," ORNL-4628, May 1973.

I 5.

Reguletory Guide 3.34, " Assumptions Used for Evaluating the Potential Radiological Consequences of Accidental Nuclear Criticality in a Uranium Fuel Fabrication Plant," USNRC, Washington, D.C., April 1977.

6.

The Effects of Nuclear Weapons, Revised Edition, S. Glasstone, Editor, U. S. Department of Defense (1962).

7.

F. A. Gifford, Jr., "Use of Routine Meteorological Observations for Estimating Atmospheric Dispersion," Nuclear Safety, Vol. 2, No. 4,

p. 48 (June 1961).

8.

E. C. Watson, " Dose Conversion Factors," HW-71171, Part 3, August 1963 (unpublished).

9.

" Safety Report - Nuclear Fuel Fabrication Plant Lingen, Niedersachsen, Federal Republic of Germany," Exxon Nuclear GmbH, XN-EU-1.002, Rev. 8 (April 1976).

10.

Report of the Task Group on Reference Man, ICRP Publication 23, Pergamon Press (1975).

_.

ML19281A940

Text

SUMMARY

SUMMARY

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