ORAU Team

NIOSH Dose Reconstruction Project

Technical Basis Document: Basis for the Development of an 

Exposure Matrix for Tennessee Valley Authority, Muscle Shoals, 

Alabama, Period of Operation: 1951–1955

Document Number:

ORAUT-TKBS-0020

Effective Date: 03/30/2004

Revision No.: 00

Controlled Copy No.: ______

Page 1 of 10

Subject Expert: Jeri L. Anderson

Document Owner

Approval: Signature on File Date: 03/29/2004 Jeri L. Anderson, TBD Team Leader

Approval: Signature on File Date: 03/30/2004 Judson L. Kenoyer, Task 3 Manager

Concurrence: Signature on File Date: 03/29/2004 Richard E. Toohey, Project Director

Approval: Signature on File Date: 03/30/2004 James W. Neton, OCAS Health Science Administrator

Supersedes:

None

TABLE OF CONTENTS

Section

Record of Issue/Revisions ...................................................................................................................

Page

2

1.0 Site Description and Operational History.....................................................................................3

2.0 Estimation of Internal Exposure ..................................................................................................4

3.0 Estimation of Radon Exposure....................................................................................................4

4.0 Estimation of External Exposure ................................................................................................. 4

References ..........................................................................................................................................9

LIST OF TABLES

Table

1 Annual organ doses due to submersion in air contaminated with UF4 .........................................

Page

5

2 Annual organ doses due to exposure to ground surface contamination ...................................... 6

3 Results of calculations of the dose rates from a drum of UF4 and a drum of yellowcake ............. 7

4 Annual organ doses due to exposure to the container of UF4...................................................... 7

5 Annual organ doses due to the assumed annual diagnostic chest X-ray..................................... 8
Effective Date: 03/30/2004 Revision No. 00 Document No. ORAUT-TKBS-0020 Page 2 of 10

RECORD OF ISSUE/REVISIONS

ISSUE 

AUTHORIZATION 

DATE

EFFECTIVE 

DATE

REV. 

NO. DESCRIPTION

Draft 01/05/2004 00-A New Technical Basis Document for the 

Tennessee Valley Authority Site. Initiated by Jeri 

L. Anderson.

Draft 01/18/2004 00-B Incorporates internal comments. Initiated by Jeri 

L. Anderson.

Draft 02/20/2004 00-C Incorporates internal and NIOSH review 

comments. Initiated by Jeri L. Anderson.

03/30/2004 03/30/2004 00 First approved issue. Initiated by Jeri L. 

Anderson.
Effective Date: 03/30/2004 Revision No. 00 Document No. ORAUT-TKBS-0020 Page 3 of 10

1.0

The Tennessee Valley Authority (TVA) in Muscle Shoals, AL had a cooperative agreement with the 

Atomic Energy Commission (AEC) to research the recovery of uranium during the production of 

fertilizer from phosphate ore. The TVA performed this research and development from 1951 to 1955. 

The facilities involved in the project were laboratories 20 and 21 of the Research and Engineering 

Office Building and the Pilot Plant Building 411. Also, two laboratories (T-283 and L-275/276) at the 

National Fertilizer Development Center were used for analytical work associated with uranium 

production. The TVA owned and operated the laboratories and pilot plant and still maintains 

ownership of the facility (US DOE 1980, US DOE 1987).

SITE DESCRIPTION AND OPERATIONAL HISTORY

In 1951, TVA began researching processes for producing fertilizer from leached-zone phosphate ore. 

This ore is a low-grade, high-alumina phosphate ore that contains very small amounts of uranium. 

The leached-zone ore is normally discarded as waste during phosphate mining processes, but TVA 

was interested in conserving this material and developing economical methods of producing fertilizer 

from it. A pilot plant was constructed and a process was developed to extract the phosphate with a 

mixture of nitric and sulfuric acid. The principal steps of the process were to calcine the ore at 2000°

to 2100°F, and then extract the phosphorus pentoxide (P2O5) with a mixture of nitric and sulfuric 

acids. Over 90% of the uranium in the ore was extracted with the phosphate, making it available for 

recovery. The extraction slurry was then filtered and the filtrate concentrated by evaporation. The 

uranium was recovered at this point. The concentrated filtrate was ammoniated and granulated, and 

potassium chloride was added to produce a three-component fertilizer (nitrogen, phosphorus, and 

potassium). The product was then dried (Hignett et al. 1957).

The research and development on uranium recovery at TVA involved using solvent extraction to 

recover the uranium from the acid extract, a process that was originally developed by Dow Chemical 

Company of Walnut Creek, California (Stolz 1958). The process involved using alkyl pyrophosphoric 

acids to extract uranium from the phosphoric acid. Essentially, the phosphoric acid containing the 

uranium from the acid plant was reduced with iron. Then a pyrophosphate ester was added which 

complexed the uranium. The organic complex was then separated from the phosphoric acid solution, 

which was sent back to the plant for continuation of the fertilizer production process. Sulfuric acid was 

used to remove calcium, iron and other ions, and the uranium was recovered by reacting with 

hydrofluoric acid. The recovered uranium in the form of uranium tetrafluoride, UF4, was dried and 

shipped to the AEC for further processing (Greek et al. 1957).

Very little actual work was done on uranium recovery from the acid extracts because it was first 

necessary to ensure that the fertilizer production by this method was marketable and economically 

feasible. It appears that only about 5.5 pounds of uranium concentrate was produced which 

contained an average of 2% (range of 0.3-25%) uranium by weight. No accountability records were 

retained so the disposition of the material and the laboratory and pilot plant equipment is unknown. 

There was also no information found concerning the radiological status of the facilities at the 

termination of the project (US DOE 1980, US DOE 1987).

Because no monitoring records have been found for the AEC work at the TVA facility, exposures at 

the facility are estimated using the Blockson Chemical Company exposure matrix (Anderson 2003). 

The uranium recovery process at the Blockson facility was a much larger process that also involved 

recovery of uranium from wet phosphoric acid. The process was slightly different in that the uranium 

was precipitated using sodium hydrosulfite, and the uranium concentrate was a sodium uranous 

phosphate as opposed to uranium tetrafluoride (UF4). Blockson produced 50,000 pounds of U3O8 per 

year compared to the 5.5 pounds produced by TVA from 1951 to 1955. For the purpose of dose 

reconstruction, the Blockson exposure matrix is scaled down proportionately using the ratio of the 
Effective Date: 03/30/2004 Revision No. 00 Document No. ORAUT-TKBS-0020 Page 4 of 10

estimated quantity of U3O8 produced daily at Blockson to the quantity of UF4 produced daily at the 

TVA facility.

2.0

The uranium produced at the TVA was the result of process research and development. The nature 

of research and development work is that it is not necessarily constant and ongoing as is the case 

with production work. Thus, potential intakes would most likely result from a series of acute 

exposures rather than a chronic exposure. However, because specific information regarding periods 

of exposure is unknown, to simplify the dose reconstruction, a single chronic intake is used to 

approximate multiple acute intakes.

ESTIMATION OF INTERNAL EXPOSURE

The TVA uranium recovery process was a wet process. Therefore, the primary potential inhalation 

hazard would have been once the uranium had been recovered from the acid in the form of UF4 and 

dried and packaged for shipment. Blockson produced an estimated 50,000 pounds of U3O8 per year, 

or 137 pounds (62.1 kg) per day. The total amount of uranium concentrate reported to be produced at 

TVA from 1951 to 1955 was 5.5 pounds (2.5 kg), which averages out to 0.003 pounds (1.37 g) per 

day. Thus, it is assumed that the quantity aerosolized at the TVA facility was a factor of 45,300 lower 

than the quantity aerosolized at Blockson.

Urinalysis data were available for 25 Blockson Chemical employees. These data were used to fit 

chronic intakes for each of the workers to determine exposure from inhalation. The resulting chronic 

intake rates were lognormally distributed with a median of 24 pCi/d and a geometric standard 

deviation of 1.6. Using the breathing rate of 1.2 m3

/h and an assumed 8-hour workday, the air 

concentration was estimated to be 2.5 pCi/m3

. The air concentration during operations at TVA is 

estimated by reducing the air concentration at the Blockson facility by a factor of 45,300, which results 

in an estimated air concentration of 5.5E-05 pCi/m3

. If the median chronic exposure rate at Blockson 

of 24 pCi/d is reduced by a factor of 45,300, the resulting chronic exposure rate is 5.3E-04 pCi/d. 

Assuming a continuous exposure to 5.3E-04 pCi/d for 5 years gives a total inhalation intake of 1 pCi. 

This level of exposure is considered insignificant and is not included in this dose reconstruction.

Ingestion intakes were estimated using guidance provided by NIOSH/OCAS (NIOSH 2004). The 

amount of uranium ingested daily is based on the average activity air concentration and is estimated 

to be 1.1E-05 pCi/d. This quantity is insignificant is therefore not included in this dose reconstruction.

3.0 ESTIMATION OF RADON EXPOSURE

RESERVED

4.0

The primary radionuclides of interest for potential external exposure are U-238 and daughter 

radionuclides Th-234 and Pa-234m. For the purpose of this dose reconstruction, it is assumed that 

there was a potential for external exposure from five sources: submersion in air contaminated with 

ESTIMATION OF EXTERNAL EXPOSURE
Effective Date: 03/30/2004 Revision No. 00 Document No. ORAUT-TKBS-0020 Page 5 of 10

UF4 dust, exposure from contaminated surfaces, exposure from contamination on the skin, exposure 

to drums of UF4, and the assumed annual occupationally-related diagnostic x-ray. 

For estimating external exposure due to submersion in air contaminated with UF4 dust, the air 

concentration level was estimated to be 5.5E-05 pCi/m3 (see Section 2.0). This estimated air 

concentration and an assumed 2000-hr work year were combined with dose coefficients for U-238, 

Th-234 and Pa-234m from Federal Guidance Report No. 12 (US EPA 1993). Table 1 shows external 

annual organ dose estimates due to submersion of workers in air contaminated with UF4 dust. The 

cumulative dose from 1951 to 1955 is less than 1 mrem and is therefore not included in this dose 

reconstruction.

Table 1. Annual organ doses due to submersion in 

air contaminated with UF4. 

Organ Annual organ dose (rem)

Adrenal 1.1E-12

U bladder 1.2E-12

Bone surface 3.4E-12

Brain 1.5E-12

Breast 1.8E-12

Esophagus 1.1E-12

Stomach wall 1.3E-12

Small intestine 1.1E-12

Upper large intestine wall 1.1E-12

Lower large intestine wall 1.1E-12

Kidney 1.3E-12

Liver 1.3E-12

Lung 1.4E-12

Muscle 1.4E-12

Ovaries 1.1E-12

Pancreas 1.1E-12

Red bone marrow 1.3E-12

Remainder 1.2E-12

Skin 8.0E-11

Spleen 1.3E-12

Testes 1.5E-12

Thymus 1.4E-12

Thyroid 1.5E-12

Uterus 1.1E-12

There was also the potential for external exposure from surface contamination in the area where the 

UF4 was dried and packaged. To be claimant favorable it was assumed that a certain amount of UF4-

contaminated dust was allowed to build up between cleaning of the area. Although it was likely that 

the experimental work took place in a ventilation hood, to be claimant favorable, it was assumed that 

no hood was used. To estimate the extent of contamination on surfaces, the air concentration 

determined from the urinalysis results (5.5E-05 pCi/m3

) was multiplied by the indoor deposition 

velocity and the assumed deposition time. The indoor deposition velocity is dependent on the 

physical properties of the room (air viscosity and density, turbulence, thermal gradients, surface 

geometry) and the particles (diameter, shape, density). Because these characteristics are unknown, 

the terminal settling velocity was calculated for an aerosol with ICRP 66 default particle size of 5 μm 

activity mean aerodynamic diameter (AMAD) (ICRP 1994). The calculated terminal settling velocity of 

0.00075 m/s was used as an estimate of the velocity of deposition to surfaces in the rooms. This 

value is within the range of deposition velocities (2.7E-06 to 2.7E-03 m/s) measured in various studies 
Effective Date: 03/30/2004 Revision No. 00 Document No. ORAUT-TKBS-0020 Page 6 of 10

(US NRC 2002) and is considered claimant-favorable. Also, room air exchange rates, ventilation, and 

facility housekeeping practices are unknown so it was assumed that there was a steady state air 

concentration and that surface contamination was the result of 365 days (1 year) of settling.

The estimated surface contamination is multiplied by the dose coefficients for U-238 and daughter 

radionuclides Th-234 and Pa-234m for contaminated ground surfaces from Federal Guidance Report 

No. 12 (US EPA 1993). Table 2 shows the external organ dose estimates due to exposure to ground 

surface contamination. The cumulative dose from 1951 to 1955 for each of the organs is less than 1 

mrem and is therefore not included in this dose reconstruction.

Table 2. Annual organ doses due to exposure to 

ground surface contamination. 

Organ Annual organ dose (rem)

Adrenal 2.5E-10

U bladder 2.8E-10

Bone surface 7.1E-10

Brain 2.6E-10

Breast 3.6E-10

Esophagus 2.3E-10

Stomach wall 2.8E-10

Small intestine 2.6E-10

Upper large intestine wall 2.7E-10

Lower large intestine wall 2.7E-10

Kidney 2.8E-10

Liver 2.8E-10

Lung 2.9E-10

Muscle 3.4E-10

Ovaries 2.6E-10

Pancreas 2.5E-10

Red bone marrow 2.9E-10

Remainder 2.7E-10

Skin 1.3E-07

Spleen 2.8E-10

Testes 3.5E-10

Thymus 2.7E-10

Thyroid 3.0E-10

Uterus 2.6E-10

It was also assumed that there was a potential to receive a shallow dose from electrons due to skin 

contaminated with UF4. The amount of skin contamination was calculated by using a measured 

deposition velocity for 4-μm particles to skin of 0.012 m/s (Fogh et al. 1999, Andersson et al. 2002). 

For simplification, it was assumed that the material deposited on the skin during an 8-hour period was 

deposited at the beginning of the shift and the worker took a shower at the end of the shift. The 

estimated amount of skin contamination was combined with electron dose-rate conversion factors for 

U-238, Th-234, and Pa-234m for skin in contact with radionuclides (Kocher and Eckerman 1987). 

The worker was assumed to receive exposure from skin contamination only during the hours worked. 

Based on these assumptions, the annual dose due to the skin from skin contaminated with UF4 was 

estimated to be 1.9E-07 rem. Consequently, this skin dose is not included in this dose reconstruction.

There was also the potential for exposure to the container the uranium concentrate was stored in. It 

was assumed that all 5.5 pounds of the uranium concentrate (containing 0.3-25% uranium) produced 

at the site was stored in one container. NIOSH/OCAS provided results of a survey of measurements 
Effective Date: 03/30/2004 Revision No. 00 Document No. ORAUT-TKBS-0020 Page 7 of 10

of partially filled 55-gallon drums of UF4 at the DOE facility at Fernald. These drums were assumed to 

be half-full, which suggests that a drum contains approximately 1500 pounds of UF4. Measurements 

were taken at the center side and at the bottom side of the drum. For the Blockson Chemical 

exposure matrix, dose rates were also calculated for 55-gallon drums of yellowcake using 

MicroShield© (Grove Engineering 2003) and MCNP (LANL 2003). The mean measurements for the 

center and the bottom of the Fernald drums were averaged together to get a dose rate of 1.3 mrem/h 

at the surface. The drum measurements at 1 meter from the surface were below the instrument’s 

detection limit of 0.5 mrem/h. So, to get an estimate of the dose rate at 1 foot and 1 meter from the 

UF4 drums, the surface dose rate was divided by the average ratio of the surface-to-1-foot- and 

surface-to-1-meter-calculated dose rates obtained with MicroShield© and MCNP. Table 3 shows the 

results of the calculations for the UF4 and yellow cake drums. 

Table 3. Results of calculations of the dose rates from a drum 

of UF4 and a drum of yellowcake. 

Dose rate

Surface (side) 30 cm (1 ft.) 1 m

MicroShield (mR/h) 5.5E-01 8.4E-02 2.2E-02

MCNP (mrem/h) 5.6E-01 1.3E-01 3.6E-02

UF4 (mrem/h) 1.3E+00 2.4E-01 6.8E-02

The container of UF4 at the TVA was assumed to contain a maximum of 5.5 pounds, so using the 

dose rate from a 55-gallon drum containing around 1500 pounds of UF4 to estimate annual dose is 

very claimant favorable. It was assumed that the worker was 1 meter from the container of UF4 for 1 

hour per day, 5 days per week, and 50 weeks per year. The resulting annual dose due to exposure to 

the container of 1500 pounds of UF4 was calculated to be 0.017 rem.

The organ doses were calculated by multiplying the estimated annual dose of 0.017 rem by the 

“Ambient Dose Equivalent (H*(10)) to Organ Dose (HT)” photon dose conversion factors found in 

Appendix B of the NIOSH External Dose Reconstruction Implementation Guideline (NIOSH 2002a). 

The exposure geometry was assumed to be anterior-posterior (AP) and the dose rate was divided 

equally between photons with E=30-250 keV and photons with E>250 keV. Table 4 below shows the 

annual organ doses due to the potential exposure to the container of UF4. 

Table 4. Annual organ doses due to exposure to the 

container of UF4. 

Organ

Annual organ dose (rem)

Photons

E=30-250 keV

Photons

E>250 keV Total

Bladder 8.0E-03 7.7E-03 1.6E-02

Red bone marrow 4.1E-03 6.3E-03 1.0E-02

Bone surface 7.8E-03 6.7E-03 1.4E-02

Breast 8.1E-03 8.2E-03 1.6E-02

Colon 6.8E-03 7.4E-03 1.4E-02

Esophagus 4.4E-03 6.5E-03 1.1E-02

Eye 8.0E-03 7.7E-03 1.6E-02

Ovaries 6.2E-03 7.2E-03 1.3E-02

Testes 9.2E-03 8.3E-03 1.7E-02

Liver 6.8E-03 7.5E-03 1.4E-02

Lung 6.3E-03 7.3E-03 1.4E-02

Remainder 5.7E-03 6.9E-03 1.3E-02

Skin 5.7E-03 7.3E-03 1.3E-02

Stomach 8.1E-03 7.8E-03 1.6E-02
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Thymus 9.0E-03 7.8E-03 1.7E-02

Thyroid 9.2E-03 8.5E-03 1.8E-02

Uterus 6.5E-03 6.9E-03 1.3E-02

The organ doses in the second and third columns of Table 4 are entered into the NIOSH IREP 

program assuming a chronic exposure and a constant distribution. The organ doses in the second 

column are attributed to photons with E=30-250 keV and the organ doses in the third column are 

attributed to photons with E>250 keV.

The assumption was also made that workers received an annual occupationally related diagnostic Xray. The exposure geometry was assumed to be posterior-anterior (PA) (NIOSH 2002a). Table 5 

below shows the annual organ doses due to the assumed annual diagnostic chest X-ray (Kathren et 

al. 2003). The values in Table 5 are entered into the NIOSH-IREP program as the annual dose due to 

an acute exposure to photons (E=30-250 keV). The distribution is assumed to be normal with a 

standard deviation of +/- 30%. 

Table 5. Annual organ doses due to the 

assumed annual diagnostic chest X-ray. 

Organ Annual dose (rem)

Thyroid 3.48E-02

Eye/Brain 6.40E-03

Ovaries 2.5E-02

Liver/Gall Bladder/Spleen 9.02E-02

Urinary Bladder 2.5E-02

Colon/Rectum 2.5E-02

Testes 5.0E-03

Lungs (male) 8.38E-02

Lungs (female 9.02E-02

Thymus 9.02E-02

Esophagus 9.02E-02

Stomach 9.02E-02

Bone Surfaces 9.02E-02

Remainder 9.02E-02

Breast 9.80E-03

Uterus (embryo) 2.5E-02

Bone Marrow (male) 1.84E-02

Bone Marrow (female) 1.72E-02

Skin 2.70E-01

5.0

According to the NIOSH residual radioactivity report, there is very little potential for radioactive 

contamination at the TVA facility beyond the period of AEC operations (NIOSH 2002b). Therefore, 

exposures due to residual radioactivity were not included in this dose reconstruction.

Estimation of Exposure to Residual Radioactivity
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REFERENCES

Anderson, J.L., Basis for the Development of an Exposure Matrix for Blockson Chemical Company, 

Technical Basis Document for the ORAU Dose Reconstruction Project, ORAUT-TKKBS-0002, 

December 2003

Andersson, K.G., Fogh, C.L., Byrne, M.A., Roed, J., Goddard, A.J.H., Hotchkiss, S.A.M., “Radiation 

dose implications of airborne contaminant deposition to humans,” Health Physics, 82(2), pp. 

226-232

Fogh, C.L., Byrne, M.A., Andersson, K.G., Bell, K.F., Roed, J., Goddard, A.J.H., Vollmair, D.V., 

Hotchkiss, S.A.M., Quantitative measurement of aerosol deposition on skin, hair and clothing 

for dosimetric assessment—final report, Roskilde, Denmark: Risø National Laboratory, RisøR-1075(EN), 1999

Greek, B.F., Allen, O.W., and Tynan, D.E., Uranium Recovery from Wet Process Phosphoric Acid, 

Staff-Industry collaborative report, Industrial and Engineering Chemistry, Vol. 49(4), 1957

Hignett, T.P., Siegel, M.R., Kelso, T.M., and Meline, R.S., Fertilizer from Leached-Zone Ore, 

Agricultural and Food Chemistry, Vol. 5 (8), p. 587, 1957

International Commission on Radiological Protection, “Human Respiratory Tract Model for 

Radiological Protection,” ICRP Publication 66, 1994 

Kathren, R.L., Shockley, V.E., Thomas, E.M., Taulbee, T.D., Technical Information Bulletin: Dose 

Reconstruction from Occupationally Related Diagnostic X-ray Procedures, Revision 2, 

ORAUT-OTIB-0006, ORAU Dose Reconstruction Project, Cincinnati, Ohio, 2003

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materials,” Health Physics, Vol. 39(3): pp. 269-277, 1979

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