ORAU Team 

NIOSH Dose Reconstruction Project

Technical Basis Document: Basis for Development of an 

Exposure Matrix for Blockson Chemical Company, Joliet, Illinois; 

Period of Operation: March 1, 1951 through March 31, 1962 

Document Number: 

ORAUT-TKBS-0002 

Effective Date: 06/29/2004 

Revision No.: 01 

Controlled Copy No.: ________ 

Page 1 of 19

Subject Experts: Cindy W. Bloom 

Document Owner 

Approval: Signature on File Date: 06/24/2004

 Cindy W. Bloom, TBD Team Leader

Approval: Signature on File Date: 06/29/2004

 Judson L. Kenoyer, Task 3 Manager

Concurrence: Signature on File Date: 06/28/2004

 Richard E. Toohey, Project Director

Approval: Signature on File Date: 06/29/2004

 James W. Neton, OCAS Health Science Administrator 

Supersedes: 

Revision No: 00

TABLE OF CONTENTS 

Section Page

Record of Issue/Revisions......................................................................................................................3

1.0 Introduction ....................................................................................................................................4

2.0 Site Description, Operational History, and Process.......................................................................4

3.0 Estimation of Internal Exposure.....................................................................................................6

4.0 Estimation of Radon Exposures.....................................................................................................9

5.0 Estimation of External Exposure..................................................................................................10

6.0 Estimation of Exposure to Residual Activity.................................................................................15

References ...........................................................................................................................................17
Effective Date: 06/29/2004 Revision No. 01 Document No. ORAUT-TKBS-0002 Page 2 of 19 

LIST OF TABLES 

Table Page

1 Calculation of the quantity of U3O8 aerosolized per day from the estimated 

amount of yellowcake produced each year ...................................................................................7

2 Estimation of the daily intake of natural uranium based on source term .......................................7

3 Comparison of Blockson Building 55 uranium air concentrations calculated 

from urinalysis results and source term estimates to air concentrations 

measured in uranium mills.............................................................................................................9

4 Annual organ doses due to submersion in air contaminated with yellowcake 

dust..............................................................................................................................................10

5 Annual organ doses due to exposure to ground surface contamination......................................12

6 Annual organ doses due to exposure to contaminated surfaces.................................................13

7 Results of calculations of the exposure rate from a drum of yellowcake and 

a drum of UF4 ..............................................................................................................................14

8 Annual organ doses due to exposure to drums of yellowcake.....................................................14

9 Annual organ doses due to the assumed annual diagnostic chest x-ray.....................................15

10 Annual organ doses due to external exposure to residual radioactivity.......................................16

LIST OF FIGURES 

Figure Page

1 The Blockson process for the recovery of uranium from phosphoric acid .....................................5

2 Photograph of Building 55 of the Blockson Chemical Company....................................................8

3 Beta dose rate on the surface of yellowcake ...............................................................................11

4 Beta dose rate from yellowcake separated from ore for more than 100 days 

as a function of distance from the surface...................................................................................11
Effective Date: 06/29/2004 Revision No. 01 Document No. ORAUT-TKBS-0002 Page 3 of 19 

RECORD OF ISSUE/REVISIONS 

ISSUE 

AUTHORIZATION 

DATE 

EFFECTIVE 

DATE 

REV. 

NO. DESCRIPTION 

Draft 04/14/2003 00-A New document to establish the technical basis for 

the development of a radiation exposure matrix 

for Blockson Chemical Company. Initiated by Jeri 

L. Anderson. 

Draft 06/12/2003 00-B Draft revision to incorporate internal and NIOSH 

review comments. Initiated by Jeri L. Anderson. 

Draft 07/02/2003 00-C Draft revision to incorporate internal additional 

data. Initiated by Jeri L. Anderson 

Draft 09/26/2003 00-D Draft revision to incorporate NIOSH review 

comments. Initiated by Jeri L. Anderson 

Draft 10/08/2003 00-E Draft revision to incorporate NIOSH review 

comments. Initiated by Jeri L. Anderson 

Draft 10/10/2003 00-F Draft revision to incorporate NIOSH review 

comments. Initiated by Jeri L. Anderson 

10/10/2003 10/10/2003 00 First approved issue of revision 00. Initiated by 

Jeri L. Anderson, Subject Expert. 

Draft 10/27/2003 01-A Page change to correct typographical errors. 

Initiated by Jeri L. Anderson. 

Draft 12/04/2003 01-B Revision to incorporate discussion to resolve 

radon issue. Initiated by Jeri L. Anderson. 

Draft 01/20/2004 01-C Revision to incorporate DOL comments. Initiated 

by Jeri L. Anderson. 

Draft 02/19/2004 01-D Revision to incorporate additional NIOSH 

comments. Initiated by Jeri L. Anderson. 

Draft 03/29/2004 01-E Revision to incorporate additional NIOSH 

comments. Initiated by Jeri L. Anderson. 

Draft 04/27/2004 01-F Revision to incorporate additional NIOSH 

comment. Initiated by Jeri L. Anderson. 

Draft 06/04/2004 01-G Revision to incorporate additional NIOSH 

comment. Initiated by Judson L. Kenoyer. 

06/29/2004 06/29/2004 01 Approved Issue of Revision 01. Initiated by 

Judson L. Kenoyer. 
Effective Date: 06/29/2004 Revision No. 01 Document No. ORAUT-TKBS-0002 Page 4 of 19 

1.0 Introduction

Technical Basis Documents and Site Profile Documents are general working documents that provide 

guidance concerning the preparation of dose reconstructions at particular sites or categories of sites. 

They will be revised in the event additional relevant information is obtained about the affected site(s). 

These documents may be used to assist NIOSH in the completion of the individual work required for 

each dose reconstruction. 

In this document the word “facility” is used as a general term for an area, building or group of buildings 

that served a specific purpose at a site. It does not necessarily connote an “atomic weapons 

employer facility” or a “ Department of Energy facility” as defined in the Energy Employee 

Occupational Illness Compensation Program Act of 2000 (42 U.S.C. § 7384l (5) and (12)). 

2.0 Site Description, Operational History, and Process

In 1950-1951, the U.S. Atomic Energy Commission (AEC) approached several phosphate rock 

consumers about the possibility of recovering the uranium from the phosphate rock they processed. 

At the Blockson Chemical Company plant, the AEC was interested in the uranium that could be 

separated from the phosphoric acid, so in early 1951, the research staff at Blockson began an 

evaluation of the available research data and preliminary experimentation that the AEC made 

available to them. They determined that the only economically feasible approach applicable to the 

Blockson process would be to make the uranium recovery a by-product process (Stolz, Jr. 1958). 

In 1951, the AEC signed a letter contract with the Blockson Chemical Company (contract number 

AT(49-1)-606) to develop a process to extract uranium from wet phosphoric acid (US DOE 1985, US 

DOE 2002). The Blockson research staff began the research and it was eventually determined that 

by controlling pH and reducing conditions, the uranium could be precipitated from the phosphoric acid. 

After an economic evaluation of the process, a pilot plant was constructed to further test and refine 

the process. Meanwhile, laboratory investigations continued on another possible method of 

precipitating the uranium from the phosphoric acid. This method involved using chlorine as an 

oxidizing agent and then adding sodium hydrosulfite to cause precipitation. This process was much 

more successful and economical, so the pilot plant was shut down and converted. Then work began 

to upgrade the process, and a recovery plant was designed and constructed. 

The letter contract was later replaced by another contract (contract number AT(49-1)-611) that was 

signed October 15, 1951. Under this second contract, Blockson constructed, at its own expense, a 

facility (Building 55) to house uranium recovery equipment at their plant in Joliet, Illinois. The AEC 

furnished and installed the uranium recovery equipment (US DOE 1985). On August 15, 1952, 

Blockson began production and delivery of uranium concentrates to the AEC (Stolz, Jr. 1958). 

According to the contract, production was limited to not more than 50,000 pounds of uranium per year 

(US DOE 1983, US DOE 1985). 

In 1955, Blockson was sold to the Olin Mathieson Chemical Corporation who assumed the liabilities 

and obligations of Blockson under all contracts, as stated in contract number AT(49-1)-611 

Amendment 1. The Olin Corporation continued the uranium recovery program under contract with the 

AEC. The actual amount of uranium produced for the AEC is not known. However, according to the 

contract, the amount of uranium produced was limited to not more than 50,000 pounds per year. In 

March 1962, the uranium recovery work was discontinued with the expiration of the contract (US DOE 

1985). 
Effective Date: 06/29/2004 Revision No. 01 Document No. ORAUT-TKBS-0002 Page 5 of 19 

According to the contract signed in October of 1951, Blockson, and later Olin Mathieson, was 

responsible for the health and safety of the employees at the site and for conforming to AEC health 

and safety regulations and requirements. In Amendment 3, effective January 1, 1958, this statement 

was deleted. Personnel with the Formerly Utilized Sites Remedial Action Program (FUSRAP) 

conducted records searches for information regarding the uranium recovery activities at Blockson. No 

records of health and safety inspections by the AEC were found as a result of their search, although 

there was evidence of periodic visits by AEC personnel to review and audit process operations (US 

DOE 1985). 

The recovery plant was put into operation on August 15, 1952, approximately 17 months after 

research on the process was begun. The process was patented and the patent, USP 2743156, was 

assigned to the AEC (Stolz, Jr. 1958). A one-story, 100-by-175-foot building was built specifically to 

house the uranium recovery process (US DOE 1983, US DOE 1985). The recovery plant was 

designed to be capable of recovering uranium from 1500 tons of phosphate daily (Stolz, Jr. 1958). 

Figure 1 shows the schematic flowchart of the Blockson process for the recovery of uranium from wet 

phosphoric acid. 

Figure 1. The Blockson process for the recovery of uranium from 

phosphoric acid. [Reproduced from Clegg and Foley 1958]. 

The Blockson Chemical Company manufactured wet-process phosphoric acid from Florida phosphate 

rock (Barr et al. 1955, Clegg and Foley 1958). The Blockson plant produced technical phosphates 

rather than fertilizers from wet phosphoric acid (Wilkinson 1976). In the process, the phosphate rock 
Effective Date: 06/29/2004 Revision No. 01 Document No. ORAUT-TKBS-0002 Page 6 of 19 

is calcined and then digested with sulfuric acid resulting in phosphogypsum and phosphoric acid. The 

phosphogypsum partitions most of the calcium and radium, and the phosphoric acid partitions around 

90% of the uranium. Very little uranium is lost to the phosphogypsum. The phosphoric acid is then 

converted into monosodium phosphate and other phosphorus derivatives. The uranium by-product is 

precipitated from the monosodium phosphate stream. The monosodium phosphate liquor is heated 

and clarified. Sodium hydrosulfite (Na2S2O4) is added to precipitate the uranium. The liquor is filtered 

and the filtrate is returned to the phosphate-processing plant. The precipitate, containing about 5% 

U3O8 is slurried in water in an upgrading step in which the uranium is redissolved. The uranium is 

then reprecipitated as sodium uranous phosphate. The slurry is filtered and the precipitate, known as 

yellowcake and containing 40 to 60% U3O8, is dried for shipping (Clegg and Foley 1958, McGinley 

2002, Wimpfen 2002). The uranium content of the phosphate rock consumed in these processes 

averaged about 0.014% U3O8 (Stolz, Jr. 1958). 

3.0 Estimation of Internal Exposure

The greatest potential for internal exposure associated with the uranium recovery process arises in 

the final packing areas. Here the essentially pure uranium compound is dried and barreled for 

shipping resulting in a potentially dusty operation (Eidson and Damon 1984, US NRC 2002b, Wimpfen 

2002). In all other areas of the plant, wet processes are used and the surface contamination and dust 

exposures are minimal (Clegg and Foley 1958, US NRC 2002b). 

A study was done (Eidson and Damon 1984) of uranium aerosols generated during yellowcake 

packaging operations at four uranium mills. The study described a sequence of steps common to all 

four uranium mills: 

1. No activity. This is when the plant is shut down for maintenance or all available yellowcake 

was packaged during a previous shift. Workers are generally not present during this step. 

2. Drum loading. This occurs when a drum is placed under a hopper containing the dried 

yellowcake. The yellowcake is allowed to fall into the drum. The amount of time workers 

spend in this area varies as it depends on the size of the yellowcake inventory in the hopper. 

(It is not clear whether or not a hopper was used at the Blockson Chemical facility.) 

3. Drum uncovering. This step occurs when a filled drum is removed from beneath the hopper. 

In some cases, the drum may be vibrated to compact the yellowcake before uncovering. 

4. Powder sampling. This occurs when a worker takes a sample of yellowcake to analyze for 

moisture content. 

5. Lid sealing. This occurs when a worker places a lid on the drum and seals it. 

6. Other activities. This step includes maintenance and hosing area and equipment with water to 

clean. Hosing the packaging area to clean is a routine operation at uranium mills. 

During the study, aerosol samples were taken in yellowcake packaging areas before, during, and after 

drums of yellowcake were filled and sealed. Median aerosol concentrations during the study ranged 

from 40 to 340 μg U/m3

. Results from analysis of the air samples showed that appreciable amounts 

of airborne uranium would be expected to deposit in the nasopharyngeal region of the respiratory tract 

if inhaled by a worker not wearing respiratory protection (Eidson and Damon 1984). 
Effective Date: 06/29/2004 Revision No. 01 Document No. ORAUT-TKBS-0002 Page 7 of 19 

In order to estimate worker exposure at the Blockson uranium recovery facility, the total amount of 

U3O8 produced from 1952 to 1962 was estimated. A report showed that by the end of 1955, Blockson 

produced 1,221,470 pounds of uranium concentrate (US DOE 1985), which is roughly 600 pounds of 

U3O8 per day assuming a U3O8 concentration of 60%. However, the Blockson process was designed 

to process only 1500 tons of phosphate daily, which is approximately 400 pounds of U3O8 per day. 

Another document indicates that production was limited by contract to not more than 50,000 pounds 

of U3O8 per year (US DOE 1983). Production was stopped in March 1962 (US DOE 1985). To 

estimate the source term, it was assumed that production was limited by contract to 50,000 pounds of 

U3O8 per year. Assuming the same rate of production in 1952 and 1962 as in 1953 through 1961, the 

amount of U3O8 produced is estimated at 18,900 and 12,300 pounds in 1952 and 1962, respectively. 

These annual estimated production values were used to calculate the amount of U3O8 produced per 

day shown in Table 1. 

Thus, the estimated total amount of U3O8 produced at Blockson from 1952 to 1962 was 480,000 

pounds, which is approximately 800,000 to1,200,000 pounds of uranium concentrate (for a U3O8

concentration of 60 to 40%, respectively). 

Table 1. Calculation of the quantity of U3O8 aerosolized per day from the estimated amount of 

yellowcake produced each year. 

Work year 

Number of 

days operated 

per year 

Pounds of U3O8

produced annually 

(lbs) 

Pounds of U3O8

produced per day 

(lbs/day) 

Quantity of U3O8

aerosolized per day 

(lbs/day) 

8/15/1952-12/31/1952 138 18,900 137 1.32E-04 

1953-1961 365 50,000 137 1.32E-04 

1/1/1962-3/31/1962 90 12,300 137 1.32E-04 

Table 2 shows the quantities used to calculate the daily intake by inhalation of natural uranium based 

on the estimated source term. The daily concentration of uranium in the air at the Blockson plant was 

calculated by assuming that 9.6E-07 (US NRC 2002c) of the U3O8 produced per day was aerosolized. 

This value was divided by the estimated interior volume of Building 55. The volume of the building 

was estimated by using the reported dimensions of 100-by-175 feet (US DOE 1983). Although 

Building 55 was reported to be a one-story building (US DOE 1985), drawings of the building that 

were used to show radiological survey locations indicate that there were four levels. The first level 

contained a loading dock and storage room, the second level contained change rooms and a lab, the 

third where there was a soundproof booth, and the fourth level which consisted of catwalks that 

allowed access to various parts of the process equipment (US DOE 1983). The interior height of the 

building was assumed to be 30 feet: 8 feet for each of the first two levels and 7 feet each for the third 

and fourth levels. Based on the radiological survey report (US DOE 1983) and the photograph shown 

below in Figure 2, this is a reasonable assumption. The estimated volume of the building was 

reduced by 10% to account for equipment displacement. This gives an estimate of 13,400 m3

 for the 

interior volume of Building 55. 

Table 2. Estimation of the daily intake of natural uranium based on source term. 

Quantity of natural 

U in air per day 

(μg)

Natural U air 

concentration per day

(μg/m3

)

Activity of air 

concentration 

(pCi/m3)

Breathing 

rate 

(m3/h)

Daily 

intake 

(pCi/d)

5.1E+04 3.8 2. 6 1.2 25 
Effective Date: 06/29/2004 Revision No. 01 Document No. ORAUT-TKBS-0002 Page 8 of 19 

Figure 2. Photograph of Building 55 of the Blockson Chemical 

Company. [Reproduced from Barr et al. 1955]. 

The daily air concentration of uranium activity was multiplied by the breathing rate for adult light 

workers to obtain an estimated intake of 25 pCi per day of natural uranium due to inhalation. The 

breathing rate was calculated from the volume of air breathed by an adult light worker shown in Table 

6 on pg. 23 of ICRP Publication 66 (ICRP 1994). The light worker category assumes an activity of 1/3 

sitting and 2/3 light exercise. 

Although no air monitoring data were found for the Blockson facility, urinalysis data for 25 workers 

were found for the period between 4/20/1954 and 2/20/1958. Urine samples were received by the 

AEC on 10 different dates. The number of samples that were analyzed for each worker varied from 1 

to 10 with values ranging between 0.000 to 0.017 mg/L. The Health and Safety Division of the AEC 

New York Operation Office performed the analyses. The method of analysis was the fluorometric 

method, which had a detection limit during that time period of 0.0038 mg/L (Wilson 1958). These data 

were used to fit intakes for each of the workers assuming a relative error of 30%. A chronic inhalation 

intake from 08/15/1952 to 03/31/1962 was assumed for each worker. The material was assumed to 

be Absorption Type M (ICRP 1995) and ICRP 66 default parameters were used to calculate intakes. 

The resulting calculated chronic intake rates were lognormally distributed with a median of 24 pCi/d 

and a geometric standard deviation of 1.6. 

The calculated chronic intake rate of 24 pCi/d is used to estimate internal organ dose for workers with 

no monitoring records. The annual dose for the organ of interest should be calculated assuming 

exposure to a natural uranium mixture. The start and end dates for the chronic intake should 

encompass the period of time the employee worked during the potential exposure period. This period 

would normally be considered to begin with the recovery plant startup in August 1952, and end when 

uranium recovery operations ceased. However, to be claimant favorable, the exposure period is 

assumed to begin when research on the process began, or about 17 months prior to the recovery 

plant startup in August 1952. Thus, the covered period is assumed to be from March 1, 1951 through 

March 31, 1962. The annual organ doses can then be entered into the NIOSH IREP program as the 

annual dose due to chronic exposure to alpha radiation using a lognormal distribution with a 

geometric standard deviation (GSD) of 1.6. 
Effective Date: 06/29/2004 Revision No. 01 Document No. ORAUT-TKBS-0002 Page 9 of 19 

Considerable variation in the behavior of U3O8 has been observed with some studies indicating 

Absorption Type M and other studies indicating Absorption Type S. The ICRP in Publication 71 

recommends the use of Absorption Type M in the absence of specific information (ICRP 1995). The 

application of Absorption Type M in intake calculations using urinalysis data resulted in a daily intake 

value consistent with the value estimated from the source term. The ICRP Publication 66 (ICRP 

1994) default values should be used for the deposition parameters. 

While uranium milling is specifically excluded from the statutory definition of “atomic weapons 

employer facility,” it is interesting to compare the measured uranium air concentrations obtained in the 

Eidson and Damon (1984) study with the air concentrations calculated for Building 55 from urinalysis 

data and source term estimates. Note that the U3O8 production rate at Blockson was significantly 

lower than the production rates at the uranium mills in the study. Production rates at uranium mills 

average around 4000 pounds of U3O8 per day (Eisenbud 1987, US DOE 1997). The Blockson 

process was designed to produce only about 400 pounds of U3O8 per day (Stolz, Jr. 1958), which is 

only about 10% of the average production capacity of mills. In the uranium mill study, aerosol 

samples were taken in yellowcake packaging areas before, during, and after drums of yellowcake 

were filled and sealed. Four drums (containing approximately 1000 pounds of yellowcake) were 

loaded in succession and sealed. Powder samples were taken from each before they were sealed 

and aerosols generated by yellowcake sampling were sampled. An additional set of aerosol samples 

were taken during the drum-sealing step. To obtain aerosol samples during the drum-loading step, 

two of the mills loaded more than one drum simultaneously but at different rates. Thus, assuming 

there is a direct relationship between U3O8 production rate and uranium air concentration, it is 

possible to make a rough comparison. Table 3 compares the uranium air concentration in Building 55 

to uranium air concentrations measured during the uranium mills study. 

Table 3. Comparison of Blockson Building 55 uranium air concentrations calculated from urinalysis 

results and source term estimates to air concentrations measured in uranium mills. 

Estimated 

intake rate 

(pCi/day) 

Uranium air 

concentration 

(pCi/m3

) 

Uranium air 

concentration 

(μg U/m3) 

From urinalysis results 24 2.5 3.6 

From source term estimates 25 2.6 3.8 

Uranium mills study (Eidson and Damon 1984) 40-340 

Uranium mills study (reduced by a factor of 10 to compare 

to Blockson design rates) 

 4-34 

Uranium mills study (reduced by a factor of 25 to compare 

to estimated Blockson production rates) 

 1.6-14 

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 0.49 pCi/d, resulting in an annual ingestion intake of 123 pCi. The annual ingestion of 123 pCi 

per year from March 1951 to March 1962 results in an annual dose to the most exposed organ (bone 

surfaces) of less than 1 mrem and is therefore not included in this dose reconstruction. 

4.0 Estimation of Radon Exposures

Reserved 
Effective Date: 06/29/2004 Revision No. 01 Document No. ORAUT-TKBS-0002 Page 10 of 19 

5.0 Estimation of External Exposure

The primary radionuclides of interest for potential external exposure in Building 55 are U-238 and 

daughter radionuclides Th-234 and Pa-234m. The uranium recovery process at Blockson was a byproduct process that was designed to fit into the already existent phosphate process (Stolz, Jr. 1958). 

At the Blockson facility, a side-stream of the phosphoric acid was diverted to Building 55 where the 

uranium was separated (Wimpfen 2002). This phosphoric acid was an intermediate product in 

Blockson’s normal commercial production of technical phosphates (US DOE 1983). In the 

manufacture of this phosphoric acid, phosphate rock 

is digested with sulfuric acid resulting in phosphoric 

acid and phosphogypsum. The uranium remains 

with the phosphoric acid and the radium 

preferentially follows the phosphogypsum (Roessler 

et al. 1979, Laiche and Scott 1991). Therefore, the 

potential radium exposure was due to the 

commercial operation already in progress at 

Blockson and not due to the AEC-related work. A 

radiological survey of Building 55 that was done in 

1978 showed that contamination within the building 

was primarily uranium (US DOE 1983). 

For the purpose of dose reconstruction, it is 

assumed that there was a potential for external 

exposure from four sources: submersion in air 

contaminated with yellowcake dust, exposure from 

contaminated surfaces, exposure from contaminated 

skin, and exposure to drums of yellowcake. 

For estimating external exposure due to submersion 

in air contaminated with yellowcake dust, the 

estimated air concentration values (Table 3, from 

urinalysis results) and an assumed 2000-hr work 

year were combined with dose coefficients for U-238 

and daughter radionuclides Th-234 and Pa-234m 

from Federal Guidance Report No. 12 (US EPA 

1993). Table 4 shows external annual organ dose estimates due to submersion of workers in air 

contaminated with yellowcake dust. The cumulative dose from 1951 through 1962 is less than 1 

mrem and is therefore not included in the dose estimation. 

Clegg and Foley (1958) state that freshly separated yellowcake has a very low gamma emission rate; 

therefore, external radiation is of no particular concern at this stage of the process. However, due to 

ingrowth of daughter radionuclides in the yellowcake, the radiation levels increase for several months 

following production (US NRC 2002b). 

For accumulations of processed yellowcake dust, the surface beta dose rate from U-238 daughters is 

negligible just after separation, but rises steadily until Pa-234m and Th-234 reach equilibrium 

concentrations. After a few months, the beta surface dose rate is about 150 mrem/hr (US NRC 

2002b). Figure 3 shows the rise in beta dose rate during 100 days after separation from ore. 

Table 4. Annual organ doses due to 

submersion in air contaminated with 

yellowcake dust. 

Organ 

Annual organ dose 

(rem) 

Adrenal 5.17E-08 

U bladder 5.32E-08 

Bone surface 1.55E-07 

Brain 6.73E-08 

Breast 7.97E-08 

Esophagus 5.06E-08 

Stomach wall 5.71E-08 

Small intestine 4.98E-08 

Upper large intestine wall 5.21E-08 

Lower large intestine wall 5.07E-08 

Kidney 5.83E-08 

Liver 5.80E-08 

Lung 6.53E-08 

Muscle 6.46E-08 

Ovaries 4.95E-08 

Pancreas 4.88E-08 

Red bone marrow 6.03E-08 

Skin 3.64E-06 

Spleen 5.81E-08 

Testes 6.86E-08 

Thymus 6.15E-08 

Thyroid 6.85E-08 

Uterus 4.84E-08 
Effective Date: 06/29/2004 Revision No. 01 Document No. ORAUT-TKBS-0002 Page 11 of 19 

Figure 3. Beta dose rate on the surface of yellowcake. 

[Reproduced from US NRC 2002b] 

Figure 4. Beta dose rate from yellowcake separated from 

ore for more than 100 days as a function of distance from 

the surface. [Reproduced from US NRC 2002b] 

Figure 4 above shows that the beta dose rate from the surface of yellowcake decreases rapidly as a 

function of distance from the surface. The rapid decrease in the beta dose rate with distance, and the 

shielding afforded by shoes and clothing, reduces dose from electron exposure, particularly from 

yellowcake deposited on floors. 
Effective Date: 06/29/2004 Revision No. 01 Document No. ORAUT-TKBS-0002 Page 12 of 19 

The most likely possibility of external exposure from surface contamination was assumed to occur in 

the yellowcake packaging area. Because the AEC had strict material accountability procedures, the 

accumulation of process material was likely controlled. However, to be claimant favorable, it was 

assumed that a certain amount of yellowcake was allowed to build up between cleanings. 

To estimate the quantity of yellowcake contamination on surfaces, the air concentration determined 

from urinalysis results 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 the 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 building. This value is within the range of deposition velocities 

(2.7E-06 to 2.7E-03 m/s) measured in various studies (US NRC 2002a) and is considered claimantfavorable. Also, room air exchange rates, ventilation, and plant 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 surface from Federal Guidance Report 

No. 12 (US EPA 1993). Table 5 shows maximum external dose estimates due to exposure to ground 

surface contamination. 

With the exception of dose to the skin, the annual organ 

dose for each of the organs is less than 1 mrem. These 

values are significantly lower than some of the exposure 

rates measured in a 1978 survey (US DOE 1983). 

According to this survey, the median external exposure 

rate at 1 meter was 0.03 mR/h with a maximum of 0.3 

mR/h. Therefore, to estimate potential external exposure 

to contaminated surfaces in the plant, the median 

exposure rate was multiplied by the Exposure (R) to 

Organ Dose (rem) photon dose conversion factors from 

Appendix B of the NIOSH External Dose Reconstruction 

Implementation Guideline (NIOSH 2002). The exposure 

geometry was assumed to be isotropic and the exposure 

rate was divided evenly between the conversion factors 

for photons with energy between 30 and 250 keV and 

photons with energy greater than 250 keV. Table 6 shows 

the calculated annual organ doses from exposure to 

contaminated surfaces during plant operations. 

The organ doses in the second and third columns of Table 

6 are entered into the NIOSH IREP program assuming a 

chronic exposure and a lognormal distribution with a GSD 

of 4.0. 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. 

Table 5. Annual organ doses due to 

exposure to ground surface 

contamination. Bold italics indicate 

annual dose greater than 1 mrem. 

Organ 

Annual dose 

(rem) 

Adrenal 1.2E-05 

U bladder 1.3E-05 

Bone surface 3.3E-05 

Brain 1.2E-05 

Breast 1.6E-05 

Esophagus 1.1E-05 

Stomach wall 1.3E-05 

Small intestine 1.2E-05 

Upper large intestine wall 1.2E-05 

Lower large intestine wall 1.2E-05 

Kidney 1.3E-05 

Liver 1.3E-05 

Lung 1.3E-05 

Muscle 1.6E-05 

Ovaries 1.2E-05 

Pancreas 1.1E-05 

Red bone marrow 1.3E-05 

Skin 5.9E-03 

Spleen 1.3E-05 

Testes 1.6E-05 

Thymus 1.3E-05 

Thyroid 1.4E-05 

Uterus 1.2E-05 
Effective Date: 06/29/2004 Revision No. 01 Document No. ORAUT-TKBS-0002 Page 13 of 19 

Table 6. Annual organ doses due to exposure to 

contaminated surfaces. 

Annual organ dose (rem) 

Organ 

Photons 

E=30-250 keV 

Photons 

E>250 keV Total 

Bladder 1.61E-02 1.94E-02 3.55E-02 

Red bone marrow 1.67E-02 2.00E-02 3.67E-02 

Bone surface 2.81E-02 2.04E-02 4.86E-02 

Breast 2.12E-02 2.22E-02 4.35E-02 

Colon 1.55E-02 1.90E-02 3.45E-02 

Esophagus 1.50E-02 1.96E-02 3.46E-02 

Eye 2.23E-02 2.28E-02 4.50E-02 

Ovaries 1.48E-02 1.88E-02 3.36E-02 

Testes 1.90E-02 2.08E-02 3.98E-02 

Liver 1.70E-02 2.00E-02 3.70E-02 

Lung 1.88E-02 2.12E-02 3.99E-02 

Remainder organs 1.67E-02 1.99E-02 3.65E-02 

Skin 2.19E-02 2.28E-02 4.47E-02 

Stomach 1.70E-02 1.99E-02 3.69E-02 

Thymus 1.84E-02 2.05E-02 3.89E-02 

Thyroid 1.92E-02 2.14E-02 4.06E-02 

Uterus 1.46E-02 1.81E-02 3.27E-02 

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

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

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

al. 1999). 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. Several claimant interviews indicated that workers 

took showers as part of their contamination control program, so it was assumed that 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 and daughter radionuclides 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 to the skin due to electron exposure from skin contaminated with yellowcake is estimated 

to be 0.0018 rem. However, this skin dose is negligible compared to the shallow dose estimated from 

exposure to a drum of aged yellowcake. This scenario is described next. 

There was also the potential for exposure to drums of yellowcake during drum loading, sealing and 

sampling, and moving the drums to storage. It was assumed that 50 drums of yellowcake were 

loaded and packed each year (1000 pounds per drum, 50,000 pounds per year). MicroShield© 

(Grove Engineering 2003) and MCNP (LANL 2003) calculations were done to estimate the exposure 

to a drum of yellowcake at the surface of the drum, at 30 cm (1 ft), and at 1 m. Also, NIOSH/OCAS 

provided results of survey measurements of partially filled drums of UF4 at the DOE facility at Fernald. 

Measurements were taken at the sides of the drum at the center and bottom. The mean 

measurements for the center and the bottom of the drum were averaged together to get a dose rate of 

1.3 mrem/h at the surface. To get an estimate of the dose rate at 1 foot from the UF4 drums, the 

surface dose rate was divided by the average ratio of the surface to 1 foot calculated dose rates 

obtained with MicroShield and MCNP. Table 7 shows the results of the calculations for the 

yellowcake drums and the UF4 drums. 
Effective Date: 06/29/2004 Revision No. 01 Document No. ORAUT-TKBS-0002 Page 14 of 19 

Table 7. Results of calculations of the exposure rate from a drum 

of yellowcake and a drum of UF4. 

 Surface (side) 

Exposure rate 

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 

The UF4 values were larger and, to be conservative, were used to estimate the annual dose. During 

an interview, a claimant stated that he spent 8 hours per day, 1 or 2 days per week loading drums 

onto trucks and boxcars. Thus, to be claimant favorable in estimating the most likely annual dose, it 

was assumed the worker was 1 foot from the drum of UF4 for 8 hours per day, 1 day per week, and 50 

weeks per year. It was assumed that the amount of time spent loading barrels was lognormally 

distributed, and the assumption that the worker was exposed 40 hours per week for 50 weeks per 

year was considered to be the upper 95th percentile. Thus, the annual dose due to exposure to drums 

of UF4 (analog for yellowcake) was calculated to be 0.096 rem. 

The organ doses were calculated by multiplying the estimated annual dose of 0.096 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 2002). 

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 8 below shows the 

annual organ doses due to the potential exposure to drums of yellowcake. 

Table 8. Annual organ doses due to exposure to drums of 

yellowcake. 

Annual organ dose (rem) 

Organ

Photons 

E=30-250 keV 

Photons 

E>250 keV Total 

Bladder 4.51E-02 4.37E-02 8.88E-02 

Red bone marrow 2.30E-02 3.58E-02 5.88E-02 

Bone surface 4.39E-02 3.80E-02 8.19E-02 

Breast 4.61E-02 4.64E-02 9.24E-02 

Colon 3.85E-02 4.19E-02 8.04E-02 

Esophagus 2.51E-02 3.70E-02 6.21E-02 

Eye 4.54E-02 4.36E-02 8.89E-02 

Ovaries 3.48E-02 4.07E-02 7.56E-02 

Testes 5.23E-02 4.68E-02 9.91E-02 

Liver 3.86E-02 4.24E-02 8.11E-02 

Lung 3.60E-02 4.16E-02 7.75E-02 

Remainder organs 3.21E-02 3.91E-02 7.11E-02 

Skin 3.25E-02 4.14E-02 7.39E-02 

Stomach 4.56E-02 4.40E-02 8.96E-02 

Thymus 5.11E-02 4.43E-02 9.54E-02 

Thyroid 5.24E-02 4.82E-02 1.01E-01 

Uterus 3.67E-02 3.89E-02 7.56E-02 

The organ doses in the second and third columns of Table 8 are entered into the NIOSH IREP 

program assuming a chronic exposure and a lognormal distribution with a GSD of 2.7. 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. 
Effective Date: 06/29/2004 Revision No. 01 Document No. ORAUT-TKBS-0002 Page 15 of 19 

It was assumed that there was a potential to receive a shallow dose from exposure to open drums 

during drum loading and sealing. According to Figure 4, the dose rate at 1 foot from the surface of 

aged yellowcake is between 1 and 2 mrem/h. Therefore, to be claimant-favorable, it was assumed 

that the claimant spent 8 hours per week, 50 weeks per year at 1 foot from the surface of aged 

yellowcake at a dose rate of 2 mrem/h. Again, the time of exposure was assumed to be lognormally 

distributed and an exposure time of 40 hours per week, 50 weeks per year was assumed to be the 

95th percentile. Thus, the annual shallow dose from exposure to open drums of yellowcake is 

assumed to be 0.8 rem per year with a GSD of 2.7. 

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 2002). Table 9 

below shows the annual organ doses due to the assumed 

annual diagnostic chest x-ray (Kathren et al. 2003). The 

values in Table 9 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%. 

6.0 Estimation of Exposure to Residual Activity

After conclusion of the AEC activities in Building 55 in 

March of 1962, the building continued to be used for 

chemical processing and production of phosphate products 

from phosphate rock (US DOE 1983). Prior to the 1978 

survey by Argonne National Laboratory, there were no 

records of any radiological surveys or decontamination 

activities at the site. The results of the 1978 survey showed 

that thirty-three localized areas and three larger general 

areas exceeded allowable limits for uranium and radium226. In 15 of those locations, contamination was 

determined to be removable and available for transfer to 

other areas. Thus, dose due to exposure to residual activity 

is estimated for the purpose of dose reconstruction. 

According to this survey, the median external exposure rate at 1 meter was 0.03 mR/h with a 

maximum of 0.3 mR/h. Therefore, to estimate potential external exposure to contaminated surfaces in 

the plant, the median exposure rate was multiplied by the Exposure (R) to Organ Dose (rem) photon 

dose conversion factors from Appendix B of the NIOSH External Dose Reconstruction Implementation 

Guideline (NIOSH 2002). The exposure geometry was assumed to be isotropic and the exposure rate 

was divided evenly between the conversion factors for photons with energy between 30 and 250 keV 

and photons with energy greater than 250 keV. Table 10 shows the calculated annual organ doses 

from external exposure to residual radioactivity after the end of AEC operations at the site. 

The organ doses in the second and third columns of Table 10 are entered into the NIOSH IREP 

program assuming a chronic exposure and a lognormal distribution with a GSD of 4.0. 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 residual contamination exposure period is 

assumed to begin on April 1, 1962 and end on the employee’s last day of work in Building 55. 

The maximum internal exposure from residual radioactivity was estimated by assuming that the facility 

was uniformly contaminated at the level of maximum smear result (considered removable 

Table 9. 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 (contents) 2.5E-02 

Bone marrow (male) 1.84E-02 

Bone marrow (female) 1.72E-02 

Skin 2.70E-01 
Effective Date: 06/29/2004 Revision No. 01 Document No. ORAUT-TKBS-0002 Page 16 of 19 

Table 10. Annual organ doses due to external exposure to 

residual radioactivity. 

Annual organ dose (rem) 

Organ

Photons 

E=30-250 keV 

Photons 

E>250 keV Total 

Bladder 1.61E-02 1.94E-02 3.55E-02 

Red bone marrow 1.67E-02 2.00E-02 3.67E-02 

Bone surface 2.81E-02 2.04E-02 4.86E-02 

Breast 2.12E-02 2.22E-02 4.35E-02 

Colon 1.55E-02 1.90E-02 3.45E-02 

Esophagus 1.50E-02 1.96E-02 3.46E-02 

Eye 2.23E-02 2.28E-02 4.50E-02 

Ovaries 1.48E-02 1.88E-02 3.36E-02 

Testes 1.90E-02 2.08E-02 3.98E-02 

Liver 1.70E-02 2.00E-02 3.70E-02 

Lung 1.88E-02 2.12E-02 3.99E-02 

Remainder organs 1.67E-02 1.99E-02 3.65E-02 

Skin 2.19E-02 2.28E-02 4.47E-02 

Stomach 1.70E-02 1.99E-02 3.69E-02 

Thymus 1.84E-02 2.05E-02 3.89E-02 

Thyroid 1.92E-02 2.14E-02 4.06E-02 

Uterus 1.46E-02 1.81E-02 3.27E-02 

contamination) of 640 dpm/100 cm2. This value was multiplied by a resuspension factor of 1E-06 m-1

(US NRC 2002c). This resulted in an estimated maximum residual air concentration of 0.03 pCi/m3

. 

Assuming a breathing rate of 1.2 m3

/h and a 2000-h work year results in a possible annual inhalation 

intake of 71 pCi. This value is considered negligible as it results in an annual dose of less than 10 

mrem to the maximally exposed organ and is not included in the dose reconstruction. 
Effective Date: 06/29/2004 Revision No. 01 Document No. ORAUT-TKBS-0002 Page 17 of 19 

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