I J~ Y 3. At7 A ECI I I I I2Nw -, R RESEARCH -RPOJT

AEC RESEARCH AND HW- 81964DEVELOPMENT REPORT

BETA-GAMMA DOSE RATES FROM U2 3 2 IN U2 3 3

F. E. OWEN

APRIL 1964

IRRADIATION PROCESSING

HANFORD ATOMIC PRODUCTS OPERATION

RICHLAND, WASHINGTON

GENERAL E LECTRIC

UNIVERSITY OFARIZONA LIBRARY

Docitments ClectionSEP 9 1964

metadc100648

LEGAL NOTICE

This report was prepared as an account of Government sponsored work. Neither the United States,

nor the Commission, nor any person acting on behalf of the Commission:

A. Makes any warranty or representation, expressed or implied, with respect to the accuracy, com-

pleteness, or usefulness of the information contained in this report, or that the use of any information,apparatus, method, or process disclosed in this report may not infringe privately owned rights; or

B. Assumes any liabilities with respect to the use of, or for damages resulting from the use of

any information, apparatus, method, or process disclosed in this report.

As used in the above, "person acting on behalf of the Commission" includes any employee or

contractor of the Commission, or employee of such contractor, to the extent that such employee or con-

tractor of the Commission, or employee of such contractor prepares, disseminates, or provides access to,

any information pursuant to his employment or contract with the Commission, or his employment withsuch contractor.

AEC-GE RICHLAND. WASH.

HW -81964

UC -41, Health and Safety(TID-4500, 31st Ed. )

BETA-GAMMA DOSE RATES FROM U2 3 2 IN U2 3 3

By

F. E. Owen

Process and Reactor DevelopmentResearch and Engineering

Irradiation Processing Department

April 1964

HANFORD ATOMIC PRODUCTS OPERATIONRICHLAND, WASHINGTON

Work performed under Contract No. AT(45-1)-1350 betweenthe Atomic Energy Commission and General Electric Company

Printed by/for the U. S. Atomic Energy Commission

HW-81964

INTRODUCTION

The gamma and beta dose rates encountered with U233 are createdby the daughters of U2 3 2 which is invariably present as an impurity.

Large scale production of U2 3 3 with low concentrations of U2 3 2

(less than 10 ppm) will create a need for a simple method of predicting

dose rates to which personnel could be exposed. This report defines in

detail the source of the dose rate and describes a method by which they

may be predicted.

It would be impossible to predict all significant dose rates because

of the variety of shapes, sizes, and chemical compositions that U233 can

assume during production and use. However, the method is based on a

fully developed set of dose rates for a single finite mass of U2 3 3 that can

be easily adjusted to an actual situation. Since the exposure received by

personnel working with Pu239 is familiar to many who may be concerned.U 233 exouete 239*

with U exposures, the Pu dose rates for a comparable source are

included to be used as a comparison for assistance in predicting total

personnel exposures.

In contrast to Pu2 3 9 , the other items of radiological concern are

far less significant for U2 3 3 of low U232 concentration. Neutron dose rates

do not become significant. Its hazard value(1) for contamination control and

for internal emitters falls in the intermediate group whereas Pu2 3 9 is in the

very hazardous group.

Information calculated from Arnold's(2) work indicated that it would

not be necessary to develop fully the neutron dose rates. The spontaneous

fission neutron production in U233 contaminated with as much as 100 ppm

U232 is insignificant compared to Pu239. In the presence of most light

elements, the (a, n) reaction neutrons from U233 are less than 10% of the

neutrons from Pu2 3 9 U 2 3 2 contamination would have to exceed 10 ppm

and its age would have to be several years before it and its daughters could

Isotopic composition: 93. 5% Pu239, 6% Pu240, and 0. 5% Pu241

-2-

HW -81964

develop enough alpha activity to cause (a, n) neutrons equal to U233. Under

these conditions, the gamma dose rates would be intolerably high before the

neutron dose rate is high enough to be significant. Only if the source is

heavily shielded for gamma radiation are the neutrons likely to become

significant.

SUMMARY

The significant external dose rate from U2 3 3 comes from the daughters

of the U2 3 2 that is present as an impurity. For this reason, the dose rates

increase with time as the daughter activity increases. The rate of increase

is governed by the 1. 9-year half-life of Th228. The radiation consists pri-

marily of very penetrating gamma (2. 6 Mev) from T1208 and energetic beta

from both Bi212 (2. 26 Mev) and T1208 (1. 8 Mev). At 1 ppm U232, the dose

rate starts well below that of Pu2 3 9 * but is equal to it within a month or two

and reaches a level an order of magnitude higher within a year. The dose

rates will be in direct proportion to the amount of U232 present, increased

amounts of which will significantly reduce the age at which U2 3 3 can be

handled at a reasonable dose rate. For example, at 10 ppm U232, the U233

dose rate exceeds that of Pu239 in 10 days and is 2 orders of magnitude

higher in a year.

DISCUSSION OF CALCULATIONS

The calculations in this report are based on a thin disc of U233 metal

whose physical characteristics are:

Weight = 1. 0 kg 3Density = 18.7 g/cmThickness = 0. 7 cmDiameter = 10 cm2Projected Area = 78 cm 232 ( g kImpurity = 1 ppm U = (1 mg per kg of U2 3 3 )

If a Pu239 source is used for comparison, it also is a 1. 0-kg thin disc 10 cm

in diameter.

* Isotopic composition: 93. 5% Pu 2 3 9 , 6% Pu2 4 0, and 0. 5% Pu 2 4 1

-3-

HW-81964

Decay Activity

The U 2 3 2 decays by alpha disintegrations with a 74-year half-life. (3)

One mg has an activity of 20. 94 mc.

232The complete decay scheme for U and its daughters is shown in

Figure 1.(' The activity during a 10-year period was calculated for

each of the daughters from this decay scheme. For the first 380 days, the

activity of each individual nuclide was computed at 0. 5-day intervals for

20 days, and at 10-day intervals for 380 days through a computer program.

After this, only the Th2 2 8 (1. 9-year half-life) activity was computed. The

rest of the daughters which have a relatively short half-life are in equilibrium

with Th228 and were assumed to have equal activity. T12 0 8 and Po212 are

exceptions, being on branch chains and having activities of 36% and 64% of

Th228, respectively. The Th228 activity was calculated as follows:

( ) C -x t -x Th)CTh = T(e U - e

Th (ATh ~ U

CTh = Th2 2 8 activity (mc)

CU = Initial U232 activity (20. 94 mc)

T 0. 693 =0.363'Th 1. 9 yr

= 0. 693 = 0. 00942U 74yr

t = Total decay time for Th228 in years

' C = (21. 5) (e-0.009ret- d-0. 363t)''CThiona65)cet-d

Tl20 is on a 36% branch in the decay chain

.'' T1= (21. 5) (0. 36) (e-0. 00942t _ -0. 363t)

-4-

HW-81964

_ _ I __ __ I __ __ _ _

100

50

10

5

U Rad

Mode of Decay

U2 3 2

74 yr Alpha$

Th228

1. 9 yr Alpha

a224

3. 64 day Alpha

Rn2 2 0

54 sec Alpha

PO 216

0. 16 sec Alpha

Pb12

10. 6 hr Beta (0. 355)

Bi212

64% 36%60 min Beta (2. 26) 60 min Alpha

po 2120. 3 psec Alpha

Ti20 8

3. 1 min Beta (1. 8)

Pb2 08

K'

ioactive Decay Chain

Gamma Radiation

Photon Yield

Energy, Mev (Photon/Disintegration)

0.0580. 131

0.0840. 1370. 1690. 2080. 217

0.240.290.410.65

0.54

0.00210.00075

0.0160.00160.00130.00030.0030

0.037

0.0003

0.1150. 2390.30

0.040.730. 1241.622. 2

0. 2770.5100. 5820. 8592. 6

0.81

0.25

0.06

0. 10.250.80. 151.00

S I I I I 1 111

50 100 200 1 2

Age After Separation

FIGURE 1

Change in U233 Dose Rate Due to U232 Daughters

-5-

1 $ i

.0

.r,

0

0~

0

00

a

0. 5

0. 1

0.05

II I

2

I III5 10

Days

5 10

Years

AEC-GE RICHLAND. WASH.

-

_

" " " 1 1 1 1I 1 I,

2

I I

HW-81964

The computed activity for each of the U 2 3 2 daughters, which emit

a significant amount of gamma radiation, is tabulated at representative

time intervals in Table 1. The complete reference set of gamma photon

values given in Figure 1, used to compute dose rates in Table 1, were

consolidated for a single nuclide if the energies were about the same and

omitted if the yield was insignificant.

The total gamma radiation emitted by U233 and its daughters amounts

to 0. 2 mr/hr at 1 ft at 1 year and 2 mr/hr at 1 ft at 10 years; this is insig-

nificant compared to the gamma from even 1 ppm U232. Therefore, it has

been disregarded in the dose rate totals.

Gamma Dose Rates (Table 1)

The gamma dose rates were determined from each photon energy of

each gamma-emitting U232 daughter at the same representative time inter-

vals. The gamma activity (gamma curies) for each energy was determined

by multiplying the gamma photon yield(3' 4) in photons /disintegration by the

decay rate activity (curies) of each radionuclide. The gamma activity is

converted to dose rate at a finite distance (r/hr at 1 ft) by the dose rate

factor(3) for each particular gamma energy. Since all of the decay rate

activities were determined and tabulated (Table 1), the gamma photon yield

and the dose rate factor were combined into one multiplying factor, dose

rate conversion, which converts the decay rate activity in curies directly to

dose rate in r/hr at 1 ft.

Dose Rate Conversion(r/hr at 1 ft/curie)

Dose Rate _ Dose Rate Factor Gamma Photon Yield DecayRate(r/hr at 1 ft) (r/hr at 1 ft/curie gamma) X (Photon/disintegration) X Activity

(curie)

In Table 1, this operation is shown for the daughters of 1 mg of U2 3 2 .

(1 mg U232 = 1 ppm U232 in 1 kg U233) and the resulting individual dose rates

are given in mr/hr from millicurie values of the decay rate activities. The

total dose rate from 1 mg U232 and daughters is equal to the sum of the individ-

ual dose rates for the various energy levels of each of the radionuclides. In

Table 1 the dose rates have been individually totaled and recorded in five energy

groups to be available for use where shielding is introduced.

-6-

HW-81964

C

to

W-

C-4-

0W

Cd

rCdV

c

4-

4-' C(1

o

cd

Cd

d) -

Q--

C

U-

0)--

1I 41I_ _I_ _1_4 1

TABLE 1

ACTIVITY AND GAMMA DOSE RATE OF U232 AND DAUGHTERS

(Calculated for 1 mg U232)

Decay Rate Activity -(mc)

4 Days

me mr/hr

7 Days 14 Days

me mr/hr mc mr/hr

20 Days

m c mr/hr

40 Days

m c mr/hr

Gamma Dose Rate(mr/hr at 1 ft)

60 Days

me mr/hr

120 Days

me mr/hr

280 Days

mc mr/hr

2 Years 10 Years

mc mr /hr mc mr /hr

74 yr I

I

1.9 yr I

I

0.06

0. 13

0. 2

0.084

0. 002

0.00075

0.006

0.016

0.3

0.7

1.2

0.45

1 x 10

1. 4 x 10 2

20. 9 0.02 20.9

0.08 ---- 0.145

0.02 20.9

0.290

0.02 20.9 0.02 20.9

0.412 0.01

0.02 20.9 0.02 20.9

0.815 0.01 1.21 0.02 2.35

---- 2.08

5.0

20. 6 ---- 19.1

0.1 10.7 0.2 18.8 0.3

3.64 day II 0.24 0.037

10.6 hr

60

II 0.25 0.81

min I 0.04 0.25

IV 0.73 0.06

3.1 min II 0.3 0.10

III 0.57 1.05

IV 0.86 0.15

V 2.6 1.00

EnergyGroup

I

II

III

IV

V

1.3

1.5

0. 15

4. 7

1.8

3.7

5.5

12.5

0. 048

1.2

0. 027

0.28

0. 18

3.9

0. 83

12.5

0.025 ---- 0.066 ---- 0.191 0.01 0.312 0.01

0.018 0.02 0.057 0.07 0.176 0.21 0.294 0.35

0.018

0.0065

RepresentativeEnergy, Mev

0. 1

0. 25

0. 57

0.80

2.6

---- 0.055

0.01

---- 0.020

0.03

0.01

0.08

0.02

0.02

0.03

0. 02

0.08

0. 17

---- 0. 175

0.02 0.05

0.293 0.010.08

---- 0.063 0.01 0.105 0.02

0.08

0.02

0.25

0.02

0.07

0.08

0.04

0. 25

0.46

0.250.05

0. 79

0.02

0. 23

0. 25

0. 10

0. 79

1.39

0.40

0.09

1.31

0. 04

0. 39

0. 40

0. 17

1.31

2.31

0.721 0.03 1.12 0.05 2.29 0.1 5.0

0.698 0.84 1.09 1.31 2.25 2.7 5.0

0.697 0.03 1.09 0.04 2.25 0.1 5.0

0.20 0. 31 0.a6

0.2 10.7 0.5 18.8 0.9

6.0 10.7 12.8 18.8 22.6

0.2 10.7 0.4

1.4

0.250 0.05 0.394 0.07 0.807 0.2 1.79 0.3

0. 97

0.21

3. 13

0.06

0. 92

0. 97

0.41

3. 13

5.49

1.53

0. 33

4. 93

0.08

1.43

1.53

0. 64

4.93

8. 61

3.1

0. 7

10. 1

0. 1

3.0

3. 1

1.3

10. 1

17.6

7.0

1.5

22.4

0. 3

6.5

7.0

2.9

22.4

29. 1

3.0

3.85 0.6

15.0

3. 2

48. 1

0. 6

13.9

15.0

6. 3

48. 1

83.8

6.77 0.7

5.3

1.2

26.4

5. 6

84. 6

1.o

24. 7

26. 4

10. 9

84. 6

147. 6

Total Gamma Dose Rate at 1 ft (mr/hr)

AEC-GE RICHLAND, WASH

C)

z

Pb2 1 2

Ti2 0 8

-7-

Th2 28

HW-81964

Self-Shielding (Table 2)

The self-shielding of the gamma radiation created by the uranium is

quite significant and was calculated by the following formula:

D(p/) (1 - e~4X) bD = b

x xp

DX = Self-shielded dose rate (mr/hr)

Do = Calculated gamma dose rate of a finite energy with

no self-shielding (mr/hr) (Values from Table 1)

p/p = Mass attenuation coefficient for uranium (cm2Ig)

xp = Mass thickness of uranium slab (g/cm2)

b = Buildup factor

The contribution to the total dose rate from each successive layer of

a homogeneous slab source is reduced by self-shielding at a rate propor-

tional to e where p/p is the mass attenuation coefficient (cm2Ig).

Ultimately, the infinitely thick slab would have a dose rate equal to that

from an amount of source material, not self-shielded, contained in a layer

whose thickness is the reciprocal of the mass attenuation factor.

Although a buildup factor (b) is given in the equation, actually there

will be little buildup since the major contribution to the total dose rate

comes from the surface layers of the slab. An estimated buildup of 10%

(factor = 1. 1) was used for all cases.

-8-

HW-81964

Energy Group

Self-ShieldingReduction Factor

Age AfterSeparation, Days

4

7

14

20

40

60

120

280

2 years

10 years

TABLE 2

GAMMA DOSE RATE - SELF-SHIELDED

(at 1 ft from 1 mg U2 3 2 Contained in

1. 0-kg U2 3 3 Disc 0. 7 cm Thick)

Dose Rate at 1 ft

II III IV

0. 12

(mr /hr)

<0.01

0.01

0.03

0.05

0. 11

0. 17

0. 4

0. 8

1.7

3.0

0.48

(mr /hr)

0.02

0.04

0. 12

0. 19

0.47

0.73

1.5

3.4

7.2

12.7

0. 62

(mr/hr)

0.01

0.03

0.06

0. 11

0.25

0.40

0.8

1.8

3.9

6. 8

V Total

0.85

(mr/hr)

0.07

0.21

0.67

1. 11

2. 66

4. 19

8. 6

19. 0

40. 9

71.9

(mr /hr)

0. 10

0.29

0.88

1.46

3.49

5.49

11.3

25. 0

53. 7

94. 4

Example

From Table 1, Energy Group IV, at 10 years:

Dose rate Do = 10. 9 mr /hr

Gamma energy = 0. 8 Mev

Mass absorption coefficient - uranium (pip) = 0. 1 cm2Ig

Slab thickness (x) = (0. 7 cm) (18. 7 g/cm3) = 13 g/cm2

Dose buildup factor (b) = 1. 1

D = (10.9)(1. 1) (1 - e(0. 1)(13)x (0. 1)(13)

D = (10.9)(0. 62) = 6.8 mr/hr

-9-

HW -81964

The self-shielding factor for 0. 8 Mev gamma from the 0. 7-cm-thick

uranium source is 0. 62. The self-shielding dose rate reduction factors for

each energy group are shown in Table 2. The dose rates given in Table 2

are the results of the Table 1 values reduced by the self-shielding factors.

By totaling the resulting dose rates for the individual energy groups, the

total dose rate from the 1. 0-kg disc was determined.

Shielded Dose Rate, 1/4-in. Lead Filter (Table 3)

The low energy dose rates are significantly reduced when shielded

by a 1/4 in. of lead. The amount of reduction was calculated for each

shielding group. The reduction factor for each shielding group was calculated

from the following equation:

D

x

where

D = originating dose rate from Table 2,

b = dose buildup factor for lead,

4/p = mass absorption coefficient for lead (cm2/g),

xp = mass thickness of the lead, (1/4 in)(2. 54)(11. 34) = 7. 2 g/cm2, and

D -7.2 p/pRF = D beb/.x

The reduction factor for each energy group is shown in Table 3. The dose

rates given in Table 3 are the Table 2 self-shielded dose rate values reduced

by the 1/4-in. lead shield reduction factors.

-10 -

HW-81964

TABLE 3

GAMMA DOSE RATE AT 1 ft THROUGH 1/4-in. LEAD FILTER

(for1 mg U233 in1. Okg U233 0. 7cm thick)

Energy Group

1/4-in. Lead ShieldReduction Factor

Age AfterSeparation, Days

4

7

14

20

40

60

120

280

2

10

years

years

II

0. 025

mr /hr

<0. 01

<0. 01

<0. 01

<0. 01

<0. 01

<0. 1

<0. 1

<0. 1

0. 1

0. 1

Dose

III

0. 5

mr/hr

0.01

0.02

0.06

0. 10

0.24

0.37

0. 8

1. 7

3. 6

6. 4

Rate at 1

IV

0. 6

mr /hr

0.01

0.02

0.04

0.07

0. 15

0.24

0. 5

1. 1

2.4

4. 1

V Total

0.8

mr/hr

0.06

0. 17

0.54

0.89

2. 13

3.35

6. 9

15. 2

32. 7

57.5

mr /hr

0.08

0.21

0. 65

1.07

2. 53

3. 97

8. 3

18. 1

38. 8

68. 1

Beta Dose Rates and Total Dose Rates

In 1958, Heid and Keck(5) measured dose rates from a thin metal

disc of U 2 3 3 about 2 months after separation:

Weight

Density

Diameter

Area

Thickness

Impurity U 2 3 2

= 500to 600 g

= 18. 7 g/cm3 (assumed)

= 7 cm

= 39 cm 2

= 0. 7cm

= 60 ppm (~ 30 mg)

They used a HAPO extended CP (TCP), which has a chamber 3 in. in

diameter and 5-11/16 in. long with 7 mg/cm2 end window and 440 mg/cm 2

walls, to make the measurements. Their measurements are shown in Table 4.

-11-

TABLE 4

ABSORPTION STUDY(5)

233Mati: 'al Identification U CBM-8, MPM-2 15, 256

Chamb. Metal TCP MeterDiStanc t. mr/hr

00. 51. 01.52. 02. 53.03.5

41003150260021501850155013501200

D-1/2

0.01570. 01780. 01960. 02160. 02320. 02540. 02740. 0289

2Thick Lead, 2. 041 g/cm Per Layer

Layers mr/hr % Penetration

0 4100 1000-Lead + Acetate 3900 95.21-Lead + Acetate 1200 29.32 -Lead + Acetate 1050 25.63-Lead + Acetate 950 23.24-Lead + Acetate 900 22.05-Lead + Acetate 800 19.5

Thin Lead, 0. 136 g/cm2 Per Layer

Layers mr /hr

00 -Lead + Acetate1-Lead + Acetate2-Lead + Acetate3-Lead + Acetate4-Lead + Acetate5-Lead + Acetate

4100390022001650140013501300

By A. H. Keck Date 7-14-58

Dimensions 2-3/4-in. -Diameter ButtonWeight 500-600 g

Aluminum, 0. 02536 g/cm2

Per Layer

Layers

01

23456789

10152030405060

mr /hr % Penetration

41003800355033503200300028502700255025502300190016001450140014001400

100. 692. 686. 681. 878. 173.269. 565. 962. 262. 256. 246. 339. 135. 434. 234. 234. 2

Cellulose Acetate, 0. 00976 g/cm2

Per Layer

Layers

% Penetration 01

100 295.2 353.6 440.2 534.2 633 731.7 8

9

Thick Brass, 0. 1057 g/cm2 Per Layer

Layers mr/hr % Penetration

0 4100 1001 3000 73.22 2250 553 1800 43.94 1550 37.85 1400 34.2

Thin Brass, 0. 023 g/cm2 Per Layer

mr/hr

4100375035003300310029002700

mr/hr % Penetration

41003900380037003600350034503400330032003150285026002150180017001600

10152030405060

10095. 292. 790. 387. 885. 484. 082. 980. 578. 276. 869. 563.452. 543. 941. 539.0

% Penetration

10091.585. 580. 575. 670. 865.9

1S

Layers

0123456

Co

COaP

s

w

b

s

r

s

0

HW-81964

The distance was measured nominally from the surface of the disc to the

chamber end window, the maximum being 3. 5 in.

Figure 2 shows the calculation of the surface dose rate from their

measurements with a resulting dose rate of 72 rad/hr. The surface dose

rate of U2 3 3 metal depends on the thickness of the metal and the concentra-

tion of beta-gamma emitters (ppm U2 3 2 and age of the metal). The basic

1. 0-kg thin disc of U233 was purposely established to have the same thick-

ness as the one measured by Keck and Heid, so at the same age (60 days),

the dose will be in direct proportion to the ppm of U2 3 2 and have a dose rate

of (72 rad/hr)(1 ppm) = 1. 2 rad/hr at the surface, which is the dose rate(60 ppm) 233 232

of a thin metal disc of U containing 1 ppm U

Using the values from Table 4, Heid calculated the total dose rate

(beta plus gamma) at 1 ft to be 475 mrad/hr by the DR2 method. (4) Converting

this to a dose rate for the basic 1. 0-kg U233 thin disc with 1 ppm U232 yields

(475) (1) (78) 15. 8 mrad/hr at 1 ft. In this calculation, the dose rate was(60) (39)reduced by the ratio of U2 3 2 and increased by the ratio of the projected areas.

The thickness of both discs is the same. At 60 days, the difference between

this total dose rate 15. 8 rnrad /hr, and the calculated gamma dose rate, 5. 5

mr/hr (Table 2), amounts to 10. 3 mrad/hr. This is assumed to be the beta

contribution and it amounts to 65% of the total. This is in close agreement

with the value determined by Heid and Keck, which is slightly in excess of

60%.

Heid calculated a spectrum of effective energies from the measured

data in Table 4 by a method Helgeson documented. (6) The energy distribu-

tion measured and calculated by Heid for 60-day-old material compared

favorably with the theoretical distribution.

-13-

s

A

i1

n

S

r

s

v

s

N

Z

U2 3 3

Metal Button Diameter 2-3/4 in.

60 Days \60 ppm U

2 3 2 J

A =n 2.7)=5.9in. 2

(2 n) (l r) 2Ds ATlY D-1

D = (2r) (5.2)2

s(5. 9) (0. 02) (103)

D = 72 rad/hr Surfaces Dose Rate

0. 024

0. 020

0. 016

0.012

C. 008

0. 004

0

-3 -2 -1 0 1 2 3 4

Relative Distance, in.

FIGURE 2

Surface Dose Rate Determination by DR2 Method(4 )

'I

5

0.032

0. 028

-r = 5. 2 in.

-4 6

Co

N

O

O

It

.- 4

I

1

I

HW-81964

Beta and Gamma Energy Distribution

A Comparison Between Measured and Calculated Values

(5) iFrom Ta.ble IIIFrom Heid-Keck Theoretical

Energy Range, Mev Measurements, % Calculations, %

1.5 to 2. 6 Gamma 30 26. 1

0.5 to 1.5Gamma <5 7

0. 15 to 0.5Gamma 6 1

<0. 15 Gamma 2 <0. 1

2. 0 - 3. OMev Beta 60+ 65

212Bi is a primary beta emitter in the decay chain and it is in equilib-

rium with the other beta emitters, so it was assumed that the beta dose rate

remains in proportion to the Bi212 activity. The calculated beta dose rate of

10. 3 mrad/hr at 1 ft and the total surface dose rate of 1. 2 rad/hr at 60 days

were used to determine the dose rates at the other decay ages, by proportion-

ing them to the Bi212 activity (millicuries) in.Table 5.

Pu 239*Dose Rates

The dose rates for Pu239 given in this document are from measure-

ments of production metal and represent not a single measurement but the

long term average. The measurements were adjusted to the 1. 0-kg thin

metal disc. These are 1. 2 r /hr at the surface, 13 mr/hr at 1 ft from the

bare metal, and, when the metal is shielded by'1/4 in. of lead, 2 mr /hr at 1 ft.

Although the bare metal dose rate will build up with time as the Am241 activity

increases, this has not been considered in the dose rates given here and

shown in Figures 3 and 4. The main contributors to the Pu239 dose rate are

Pu2 3 9 and U2 3 7 and fission products. The plutonium isotopes and daughters

all have energies less than 0. 4 Mev and because so much of it is below 0. 04 Mev,

the 1/4-in. lead shield reduces total dose rate to insignificance.

* Isotopic composition: 93. 5% Pu239 6% Pu240 and 0. 5% Pu241

-15-

HW -81964

TABLE 5

Age, Days

4

7

14

20

40

60*

120

280

2 years

10 years

TOTAL DOSE RATE FOR BARE U 2 3 3 METAL

Dose Rates*

Activity Beta Total Beat and Gamma

Bi212 at 1 ft, at 1. 0 ft, Surfacemc mrad/hr mrad/hr rad/hr

0.018 0.2 0.3 0.02

0.055 0.5 0.8 0.06

0.175 1.7 2.6 0.19

0.293 2.8 4.3 0.32

0.697 6.6 10.1 0.78

1.09 10. 3* 15.8 1. 2*

2.25 21.2 32.5 2.45

5.0 47 72 5.5

10.7 110 154 11.8

18.8 117 271 20.6

* Basis: Calculated beta dose rate at 60 days. Other dose rates

are in proportion to the Bi212 activity.

-16-

-17-

- U233

Bare Metal

239 F. P. Carryover .23U'Shielded byPu Bare Metal 1/4 in. Lead Filte

239-Pu Shielded by1/4 in. Lead Filter

-233 Thin Metal Disc

1 kg 1 ppm U2 3 2

0.7 cm Thick

78 cm Area

Dose Rate

F. P. Carryover Bare Metal 65% Beta35% Gamma

Shielded Metal 100% Gamma

- -- ii50 10.0 2.00 1 2 5 10

Years

5 10

Days

Age After Separation

FIGURE 3

U233 Dose Rate at 1 ft Due to U232 Daughters

AEC-GE RICHLAND. WASH.

HW -81964

100

50

10

5~0Q

1

0. 5

0. 1

-18- HW-81964

10

5

1

10 50 100 200

Days

1 2 5 10

Years

Age After Separation

FIGURE 4

U233 Surface Dose Rate Due to U232 Daughters

AEC-GE RICHLAND. WASH.

UP233 Bare Metal

Pu239 Bare Metal

U2 3 3

Thin Metal Disc

1 kg 1 ppm U2 3 2

0.7 cm Thick

78 cm2 Area

Dose Rate: 65% Beta35% Gamma

F. P. Carryover

0. 5a)

0

0.1

0.05

0.01

5

HW-81964

Fission Product Contribution

All of the gamma radiation greater than 0. 4 Mev comes from fission

products that have carried over from the irradiated fuel separations process.

The 2 mr/hr at 1 ft dose rate (1/4-in. lead filtered) shown in Figure 3 can

be attributed entirely to fission product carryover. This dose rate will

reduce with time as the fission products decay away.

In U233 production, some fission product carryover can be expected.

On the basis of Pu239* production experience, it might be expected to con-

tribute as much as the 2 mr/hr at 1 ft and 0. 1 to 0. 2 r/hr at surface. Since

there is no sound basis for this judgment, it has been depicted in Figures 3

and 4 as gray zones marked "F. P. Carryover. "

DISCUSSION OF RESULTS

U233 metal increases in radioactivity with time because of the U232

impurity whose daughters emit both gamma and beta radiations in significant

quantity. The dose rate increases as the U2 3 2 daughters are produced as

decay products after chemical separation has taken place. The activity

buildup is determined by the 1. 9-year half-life Th228, the first daughter of

U232. The remaining daughters come into equilibrium with the Th228 veryquickly, because their total half-life amounts to only 4 days. By the end of

10 years, a maximum is reached as the Th228 reaches equlibrium with the

74-year half-life U2 3 2

Figure 1

The dose rate buildup calculated in Table 2 was converted to percent

buildup versus time growth and is presented in Figure 1. This is based on

100% for the maximum dose rate which will be reached at about 10 years,

when the Th228 comes into equilibrium with the U232 Time of growth

starts with the chemical separation step in the U233 production process.

The daughters of U232 are elementally different from uranium, so although

U 2 3 2 remains with the U233, the daughters can be removed by a chemical

separation process.

Isotopic composition: 93. 5%P239, 6% Pu240, and 0. 5% Pu241

-19-

HW-81964

The dose rates which build up 1 ft from the basic 1. 0 kg thin metal

disc are shown in Figure 3, both from the bare metal and from metal

shielded with 1/4 in. of lead. This is a graphical display of values obtained

from Tables 3 and 5.

Since there is a sizeable amount of beta and some lower energy

gamma radiations present as well as the very penetrating 2. 6 Mev gamma

from T1 2 0 8 , a 1/4-in. lead filter was used as a method for showing and

comparing the penetrating radiations with those more readily shielded.

Average dose rates encountered in handling Pu239* metal of the same size

and shape are included for comparison.

The shaded area in Figure 3 indicates the increased dose rate that

might result from fission product carryover from the separation process.

If anything, this estimate is high. With Pu239*, for example, the shielded

dose rate of 2 mr/hr at 1 ft comes entirely from fission products carried

over. It will be noted that the dose rate from bare metal U233 aged about

60 days equals that of Pu2 3 9 *. The penetrating dose rates, however, are

equal when the U233 is less than a month old because of the predominately

high energy gamma from U233 and low energy gammas from Pu239*

Figure 4

The U233 surface dose rate which builds up with time is shown in

Figure 4. These values are taken from Table 5. As a comparison, the

average surface dose rate found on Pu239* metal is shown on the chart.

It will be noted that the dose rates are about equal when the U2 3 3 has aged

about 60 days.

CONCLUSIONS

The information presented here may be used to determine or predict232 228 208 208the dose rates when U is present or whenever the Th--4j-T12 > Pb

chain occurs in a wide variety of situations for U233 processing and handling.

The information is also adaptable to situations other than those of U2 3 3

material.

Isotopic composition: 93.5% Pu2 3 9 , 6% Pu240 and 0. 5Pu241

-20-

HW-81964

1. ppm of U2 3 2

All beta and gamma dose rates from U233 are directly proportional

to the ppm of U.2 32

2. Dose Rate Buildup

The change in dose rate for a U232 created source can be deter-

mined over any time period from the curve in Figure 1.

3. Shielding

In Table 1, the dose rate from each energy of gamma radiation

has been determined at significant time intervals so the effect of

shielding can be accurately calculated.

4. Uranium Self-Shielding

An equation for determining the self-shielding for a homogeneous

source is given in the Calculations and values for the 1. 0-kg U2 3 3

disc are given in Table 3.

5. Beta Dose Rates

a. Beta radiation is a surface phenomenon so that changes in a

field beta dose rate are directly proportional to the projected

area of the source. Table 5 gives the field beta dose rates

from an area of 78 cm2

b. The beta radiations are so energetic (1. 8 to 2. 26 Mev) that

the beta-to-gamma ratio remains constant out to distances of

a foot or more from the source.

c. The surface beta dose rate is independent of the area and mass

and depends only on the concentration of U232

6. Approximations

a. If the mass of metal is larger or smaller than 1 kg, the field

dose rates from the higher energy gammas, those filtered

by 1/4 in. of lead (Table 3 and Figure 3), can be assumed to

change in direct proportion to the mass.

-21-

HW-81964

b. If the area (78 cm2) of the disc is changed, the total dose

rate 1 ft from bare metal (Table 5 and Figure 3) will

change in direct proportion to the area.

c. The beta-to-gamma ratio is about 2:1 for the 1. 0-kg thin

disc considered; this will reduce to about 1:1 for a 1-kg

sphere.

7. Th228

If Th28 is present in other materials, the dose rates can be

estimated by comparison of the Th228 curies /kg of material

to the corresponding dose rates given in this report.

ACKNOWLEDGMENTS

I wish to express my appreciation to R. H. Meichle and

R. O. Gumprecht, for their preparation and computation of individual U232daughter radionuclide activities during the initial period of buildup, and to

K. R. Heid whose measurements in 1958 made presentation of the beta dose

possible.

-22-

HW-81964

REFERENCES

1. K. Z. Morgan, W. S. Snyder, and M. R. Ford. "Relative Hazard of

the Various Radioactive Materials, " Health Physics, vol. 10, pp. 151-169.

March 1964.

2. E. D. Arnold. "Radiation Limitations on Recycle of Power Reactor Fuels, "

p/1838, Proceedings of the Second United Nations International Conference

on the Peaceful Uses of Atomic Energy, vol. 13, pp. 237-250. United

Nations, Geneva, 1958.

3. "Radiological Health Handbook, " edited by S. Kinsman, U. S. Dept. of

Health, Education and Welfare, Public Health Service, Washington 25, D. C.

September 1960.

4. Landolt-Bornstein Numerical Data and Functional Relationships in Science

and Technology. New Series, K. H. Hellwege, editor, Group I. Nuclear

Physics and Technology, vol. I. Energy Levels of Nuclei: A-5 to A-257.

Berlin: Springer-Verlag, 1961.

5. K. R. Heid and A. H. Keck. Unpublished Data, "Exposure Rates from

U3 Source Two Months after Separation. "October 8, 1958.

6. G. L. Helgeson. Surface Dosimetry and Effective Energy Calculations,

HW-41439. September 8, 1956.

7. C. M. Unruh. The Radiological Physics of Plutonium, HW-SA-2740.

August 30, 1962.

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Y

Standard Distribution

ABERDEEN PROVING GROUND

AEROJET-GENERAL CORPORATION

AEROJET-GENERAL NUCLEONICS

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AIR FORCE INSTITUTE OF TECHNOLOGY

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1

UC-41

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325

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1

1

.i: ,j -,

I

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=

4

|