TO: Advisory Board on Radiation and Worker Health Work Group on TBD-6000 

FROM: Robert Anigstein, Stephen Marschke, and John Mauro, SC&A 

SUBJECT: Alternative Model for the Calculation of Uranium Intakes at GSI 

DATE: July 25, 2012 

Alternative Model for the Calculation of Uranium Intakes at GSI 

1 Background 

During Meeting 84 of the Advisory Board on Radiation and Worker Health, the Board tasked 

SC&A with reviewing NIOSH’s use of surrogate data for estimating intakes of uranium 

at General Steel Industries (GSI). Anigstein (2012) found that the use of the concentration of 

airborne uranium dust, based on the uranium slug stamping scenario as a surrogate for airborne 

dust concentrations during uranium handling operations at GSI, does not comply with the five 

criteria for the use of surrogate data adopted by the Advisory Board (ABRWH 2010). He also 

considered other sources of surrogate data and found issues with the use of these data for GSI. 

However, as stated in his report, “The contamination levels on the floor of the Old Betatron 

Building at the time of the 1993 cleanup, reported by Murray and Brown (1994), together with 

the depletion rate and the varying hours of uranium handling operations at GSI, could be used to 

calculate the average surficial uranium concentrations during each year of the operational and 

residual periods.” We believe that NIOSH can calculate bounding values of uranium intakes by 

GSI workers without the use of surrogate data. In accordance with advice from Paul Ziemer, 

Chair of the ABRWH Work Group on TBD-6000, we present an explicit example of our 

suggested approach. 

Allen and Glover (2007) described two pathways for the inhalation and concomitant inadvertent 

ingestion of uranium dust: intakes of uranium aerosols during uranium handling operations and 

the later resuspension of dust that settled to the floor during these operations. Anigstein (2012) 

raised several questions regarding their approach, including the use of a uranium aerosol 

concentration, based on surrogate data, the duration of settling of the aerosols, and the possible 

sloughing off of large particles of uranium oxide that contribute to the accumulation of uranium 

on the floor but not to the aerosol generated during handling operations. He found that such an 

approach is not suitable for calculating uranium contamination generated during the intermittent 

uranium handling operations. He also questioned the methodology of calculating the surficial 

uranium contamination levels throughout the operational and residual periods based on the 

accumulation of uranium during one year of uranium handling operations, with no year-to-year 

carryover and no removal. 

2 Methodology of Alternative Model 

Our alternative approach assumes the continuous introduction of uranium dust throughout the 

operational period. The average rate of accumulation of uranium on the floor during each time 

period specified in the purchase orders for uranium radiography issued by the Mallinckrodt 

Chemical Works (MCW) is proportional to the fractional time of uranium handling during that 

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time period. The removal rate is based on the exponential source-term depletion rate 

recommended in OTIB-0070 (Sharfi 2012). The final surficial concentration is given by the 

average contamination level on the floor of the Old Betatron Building at the time of the FUSRAP 

cleanup, reported by Murray and Brown (1994). The airborne concentration due to resuspension 

at any time during both the operational and residual periods is calculated by applying the 

appropriate resuspension factor to the time-varying surficial concentration. The airborne 

concentration during uranium handling operations is calculated from the derived accumulation 

rate and the deposition velocity of the uranium aerosol. 

2.1 Surficial Uranium Activity Concentration 

The surficial uranium activity concentration can be derived from the following differential 

equation: 

i = surficial uranium concentration due to accumulation during time period i (Bq/m2) 

t = time (d) 

μ = fractional removal rate 

= 6.7 × 10-4 d-1 (Sharfi 2012) 

fi = fraction of time during period i during which uranium handling operations occurred 

= 

Ti = duration of uranium handling operations during time period i 

ti1 = start of time period i 

ti2 = end of time period i 

R = rate of uranium accumulation during uranium handling operations (Bq m-2 d-1) 

Integrating equation 1, using the boundary condition i(ti1) = 0, we obtain the following 

expression for the surficial activity concentration at time t: 

i(t) = surficial uranium activity concentration at time t due to accumulation during time 

period i (Bq/m2

) 

Alternative Model for Uranium Intakes at GSI -2- SC&A –July 25, 2012
 

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S(t) = total surficial uranium activity concentration at time t (Bq/m2) 

n(t) = number of uranium-handling time intervals up to time t 

Table 1 lists the dates covered by the purchase orders for uranium radiography issued by MCW. 

The total hours of uranium handling were based on the monthly or annual costs for each period, 

as well as on the hourly rate, as specified by the purchase orders. These were divided by the 

length of each period (in hours) to calculate the fraction of each period during which uranium 

handling operations took place, represented by the symbol fi in equations 1 and 2. Because the 

frequency of uranium handling operations prior to March 1, 1958, is unknown, the highest 

fraction of the later periods was assigned to the period starting from January 1, 1953, the 

beginning of covered operations at GSI. 

Table 1. Uranium Handling Times, Based on MCW Purchase Orders 

Period covered by PO Uranium handling 

Dates Hoursa Total (h)

b Fractionc 

1/1/53d 2/28/58d 0.0499e 

3/1/58 6/30/58 2928 125 0.0427 

7/1/58 10/31/58 2952 112.5 0.0381 

11/1/58 6/30/59 5808 225 0.0387 

7/1/59 6/30/60 8784 337.5 0.0384 

7/1/60 6/30/61 8760 337.5 0.0385 

7/1/61 6/30/62 8760 437.5 0.0499 

7/1/62 6/30/63 8760 125 0.0143 

7/1/63 6/30/64 8784 28.12 0.0032 

7/1/64 6/30/65 8760 28.12 0.0032 

7/1/65 6/30/66 8760 12.86 0.0015 

a Duration of period 

b Total hours of uranium handling operations during specified time period 

c Fraction of time devoted to uranium handling operations (column 4 ÷ column 3) 

d No purchase orders found for this period 

e Maximum of all later periods 

To evaluate equation 2 at time t, we must solve for R. We can derive R from equation 2, 

provided we know the value of S(t) at some known time t. Such a value can be obtained from 

the results of the survey of the Old Betatron Building performed by the Oak Ridge National 

Laboratory (ORNL) on June 7, 1993, as reported by Murray and Brown (1994, Table 4). The 

authors list the -radiation levels at 31 random locations on the first floor of this building, as 

shown in Table 2 of the present memo. These random samples constitute the best available data 

for estimating the average contamination level on the floor of the shooting room of the Old 

Betatron Building. Since the authors state that the MDA = 50 dpm/100 cm2

, we set readings of 

“<MDA” to one-half that value—25 dpm/100 cm2

. The average of the 31 readings— converted 

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to units of Bq/m2—is equal to 43.6 Bq/m2. The time tf is expressed in days since January 1, 

1953, the start of the operational period at GSI, until the date of the survey. 

Figure 1 shows the locations of the measurements reported by Murray and Brown (1994), 

superimposed on a map of the first floor of the Old Betatron Building. The blue squares denote 

the random locations. In addition, the ORNL team scanned the entire floor using large-area floor 

monitors and G-M pancake probes. Red squares indicate locations that showed elevated ß/γ

activities. Since they represent biased samples, the activities at these locations were not used in 

the present calculations but are shown to indicate the localized distribution of these “hot spots.” 

At 13 of these 25 locations, the -radiation levels were <MDA. We note that a separate ORNL 

survey of the New Betatron Building “showed no residual 238U attributable to former AECsupported operations at this site.” (Murray and Uziel 1992) 

Figure 1. Locations of -Activity Measurements in Old Betatron Building 

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Table 2. Alpha Activity Concentrations on Floor of Old Betatron Building 

Sample location Alpha 

North East dpm/100 cm2 Bq/m2a
 

0 4 <MDA 42 

0 8 <MDA 42 

0 14 <MDA 42 

1 6 <MDA 42 

1 10 <MDA 42 

1 17 <MDA 42 

2 2 <MDA 42 

2 4 <MDA 42 

2 11 21 35 

3 3 <MDA 42 

3 7 35 58 

3 9 <MDA 42 

4 12 <MDA 42 

4 15 <MDA 42 

5 6 <MDA 42 

6 7 <MDA 42 

6 10 <MDA 42 

7 8 <MDA 42 

7 13 35 58 

7 17 <MDA 42 

8 1 <MDA 42 

9 5 <MDA 42 

9 14 <MDA 42 

10 17 42 70 

11 1 <MDA 42 

14 5 28 47 

18 3 <MDA 42 

20 1 <MDA 42 

21 3 <MDA 42 

21 5 <MDA 42 

22 4 <MDA 42 

Average 43.6 

Source: Murray and Brown (1994, Table 4) 

a

 Calculated assuming “<MDA” = 25 dpm/100 cm2 

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Since tf is after the end of uranium-handling operations, we can solve for R using the second and 

third lines of equation 2. Substituting tf for t, we obtain 

R	 = rate of accumulation of surficial uranium during uranium handling operations 

= 1,195 Bq m-2 d-1 

S(tf) = average -activity concentration on first floor of Old Betatron Building at time tf 

= 43.6 Bq/m2 

tf	 = time of ORNL survey 

= 14,767 d (January 1, 1953–June 7, 1993) 

The average surficial activity concentration during a given calendar year can be estimated by 

calculating S(t) on June 30, the midpoint of the year. The results are listed in Table 3. As shown 

in this table, the surficial concentration increases markedly from 1953 to 1958, then gradually 

levels off and starts to decline after 1962. To determine the accuracy of using the midyear date, 

we explicitly calculated the average concentrations during the first 5 years by integrating the first 

line of equation 2 over each year. We found that the results of this exact calculation agree with 

the concentration on June 30, with a maximum difference of 0.6% during the first year and less 

than 0.2% in the following 4 years. This indicates that using the midpoint of the year results in 

an accurate determination. 

2.2 Airborne Uranium Activity Concentration 

The airborne uranium activity concentration due to surficial contamination is calculated by 

applying a resuspension factor to the surficial concentration. 

χS(t) = airborne uranium activity concentration due to surficial contamination at time t 

Fr	 = resuspension factor 

= 1 × 10-5 m-1
 

The airborne uranium activity concentration due to uranium handling operations can be 

estimated from R, the rate of accumulation. 

Alternative Model for Uranium Intakes at GSI -6- SC&A –July 25, 2012
 

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Table 3. Inhalation of Uranium by GSI Workers 

Year 

Bq/m2a

U concentration 

 dpm/100 cm2a dpm/m3b

Intake (dpm/calendar day) 

 Resuspension U handling Total 

1953 10,119 6,071 6.07 64.87 1,591.28 1,656.15 

1954 27,245 16,347 16.35 174.67 1,591.28 1,765.95 

1955 40,657 24,394 24.39 260.65 1,591.28 1,851.93 

1956 51,209 30,725 30.73 328.30 1,591.28 1,919.58 

1957 59,402 35,641 35.64 380.82 1,591.28 1,972.10 

1958 64,775 38,865 38.86 415.27 1,327.34 1,742.61 

1959 65,559 39,336 39.34 420.30 1,229.22 1,649.52 

1960 66,123 39,674 39.67 423.91 1,229.25 1,653.16 

1961 66,682 40,009 40.01 427.49 1,410.92 1,838.41 

1962 71,491 42,895 42.89 458.32 1,018.30 1,476.62 

1963 61,489 36,893 36.89 394.20 276.89 671.09 

1964 49,322 29,593 29.59 316.20 102.44 418.64 

1965 39,887 23,932 23.93 255.72 74.30 330.02 

1966 31,800 19,080 19.08 203.87 23.19 227.06 

1967 24,902 14,941 14.94 159.64 159.64 

1968 19,473 11,684 11.68 124.84 124.84 

1969 15,259 9,155 9.16 97.82 97.82 

1970 11,949 7,169 7.17 76.60 76.60 

1971 9,356 5,614 5.61 59.98 59.98 

1972 7,317 4,390 4.39 46.91 46.91 

1973 5,733 3,440 3.44 36.76 36.76 

1974 4,489 2,694 2.69 28.78 28.78 

1975 3,516 2,109 2.11 22.54 22.54 

1976 2,749 1,650 1.65 17.62 17.62 

1977 2,154 1,293 1.29 13.81 13.81 

1978 1,687 1,012 1.01 10.81 10.81 

1979 1,321 793 0.79 8.47 8.47 

1980 1,033 620 0.62 6.62 6.62 

1981 809 486 0.49 5.19 5.19 

1982 634 380 0.38 4.06 4.06 

1983 496 298 0.30 3.18 3.18 

1984 388 233 0.23 2.49 2.49 

1985 304 182 0.18 1.95 1.95 

1986 238 143 0.14 1.53 1.53 

1987 186 112 0.11 1.20 1.20 

1988 146 87 0.09 0.93 0.93 

1989 114 69 0.07 0.73 0.73 

1990 89 54 0.05 0.57 0.57 

1991 70 42 0.04 0.45 0.45 

1992 55 33 0.03 0.35 0.35 

1993 46 28 0.03 0.29 0.29 

a Surficial activity concentration 

b Airborne activity due to resuspension from contaminated surface 

c Intakes continued until July 10, 1993, the date remediation under FUSRAP was completed. 

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χh	 = airborne uranium activity concentration due to uranium handling activities 

= 18.44 Bq/m3 

= 1,106 dpm/m3 

c1 = conversion factor 

= 86,400 s/d 

vd	 = deposition velocity of 5 μm AMAD particles 

= 7.5 × 10-4 m/s (Allen and Glover 2007) 

2.3 Resuspension Factor 

We selected a value of 1 × 10-5 m-1 for the resuspension factor from the range of values cited by 

Sharfi (2012). Anigstein and Mauro (2012) had suggested a value of 5 × 10-5 m-1 as a plausible 

upper bound in the model described by Allen and Glover (2007). Use of such a value would 

have made the results of their model more claimant favorable. However, as we shall 

demonstrate, the present value is more appropriate for our proposed alternative model. 

Reviewing some of the resuspension factors cited by Sharfi (2012), we note that a factor of 

1 × 10-6 m

-1 is appropriate for a decommissioned facility in which surfaces would have been 

cleaned or washed, and no fresh radioactive material would have accumulated (Abu-Eid et al. 

2002, p 4). A facility with fresh, recently deposited surficial contamination that is subject to 

pedestrian and vehicular traffic may have a factor of 5 × 10-5 m

-1 or higher. In the proposed 

model, the contamination includes both recently deposited activity and material that may have 

accumulated over the entire 13.5-y period of AEC operations, whereas Allen and Glover (2007) 

postulated an accumulation over a period of one year. We believe that the intermediate value of 

1 × 10-5 m

-1 is a plausible upper bound for use with the proposed model. 

The resuspension factor should also be in reasonable agreement with the fractional removal rate 

recommended by Sharfi (2012). Removal can involve various unspecified mechanisms; for the 

purpose of the present comparison, we will assume that some unknown fraction of the removal is 

through resuspension, with the resuspended material being vented to the outside air. The 

relationship among the removal rate, the resuspension factor, and the building ventilation rate 

can be expressed by the following equation: 

μr = removal rate due to resuspension (d-1) 

≤ μ 

c2 = conversion factor 

= 24 h/d 

r = air exchange rate (h-1) 

V = volume of affected region (m3) 

A = area of contaminated surface (m2

) 

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H = height of shooting room in Old Betatron Building 

= 

= 35 ft = 10.7 m 

Solving equation 6 for the air exchange rate, r, we obtain 

r = 0.26 h-1 

μr = μ 

This is a plausible value of the effective air exchange rate for the shooting room in the Old 

Betatron Building. Although an often-quoted nominal value of the ventilation rate for an 

industrial building is about 1 h-1, there are several factors that would have reduced the effective 

air exchange rate in this instance. First, we note that the ventilators (exhaust fans) were located 

on the roof. Air entered the building primarily through the outside entrance. The entrance to the 

exposure room of the New Betatron Building was approximately 17 ft high, as scaled from the 

elevation drawing shown in Figure 2. The entrance to the Old Betatron Building (shown in 

Figure 3) appears to have had a similar height. Some of the air entering through the door would 

have been exhausted directly through the roof, without mixing with the air in the workers’ 

breathing zone that is a few feet above the floor. Thus, the effective air exchange rate while the 

outside door was open would have been less than the exhaust rate. Furthermore, the room had 

no openings other than the outside door and the door to the control room. The outside door 

would normally have been opened only to permit the passage of castings to be radiographed and 

probably, in summer, to allow additional ventilation, but would be closed while radiography was 

in progress. The door to the control was always locked while betatron radiography was in 

progress, which was approximately 41% of the time (SC&A 2008). 

Figure 2. Elevation Drawing of New Betatron Building—Detail (NRC 2009) 

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Figure 3. Photo of Old Betatron Building ( 2007) 

The purpose of this calculation is to bound the selected resuspension factor. In reality, there are 

means of removing the surficial activity other than by resuspension and subsequent exhaust. 

Thus, the removal rate due to resuspension, μr, is most likely substantially less than μ, which, 

according to equation 7, would imply a lower air exchange rate. A higher value of Fr would lead 

to a still smaller air exchange rate, one that would no longer be realistic. This calculation 

confirms that Fr = 1 × 10-5 m

-1 constitutes a plausible upper bound. 

2.4 Intakes of Uranium by Inhalation 

We calculated the inhalation of uranium by GSI workers by assuming that they were exposed to 

the airborne concentration χS(t), evaluated for each calendar year, during each work shift, 

including the periods of uranium handling operations. We assumed that they were also exposed 

to the airborne concentration, χh, during the entire time of uranium handling operations, as listed 

in Table 1. The intakes are based on a breathing rate of 1.2 m3

/h; they are listed in Table 3 for 

each calendar year from the start of AEC operations until the time of the cleanup under 

FUSRAP. 

3 Discussion of Model 

Our alternative model utilizes all available GSI-specific data, and does not employ any 

measurements from other sites. The model assumes two sources of airborne uranium 

concentrations: disturbance of the uranium metal during handling operations and resuspension 

of uranium dust on the floor. Since these concentrations have two separate source terms, it is 

appropriate that the intakes should be additive. 

The model does assume that the uranium oxide released during the uranium handling operation 

forms an aerosol with a particle size distribution equal to 5 μm AMAD. Anigstein (2012) 

questioned that assumption as employed by Allen and Glover (2007), because the uranium oxide 

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could have included a wide range of particle sizes, including large particles that were not 

respirable and would not have been included in the measurements reported by Harris and 

Kingsley (1959) that were the basis of the uranium source term used by Allen and Glover. In the 

present analysis, however, that calculation is performed in reverse. The uranium source term 

(i.e., the deposition rate), is calculated from the site-specific surficial contamination levels. 

Since the derived concentration during uranium handling operations is inversely proportional to 

the deposition velocity, which increases with particle size, our assumption regarding the particle 

size distribution leads to a higher concentration and to a higher respirable fraction than would 

result from larger particles. Both of these are claimant-favorable results. 

4 Results 

One result of the analysis is an evaluation of the airborne uranium activity concentration due to 

uranium handling activities equal to 1,106 dpm/m3

. This value is approximately twice the 

breathing-zone activity concentration of 590 dpm/m3 reported by Harris and Kingsley (1959) for 

the uranium slug-stamping operation. However, it is about 30% of the value of 3,926 dpm/m3 

presented by Anigstein (2012) that was derived from Adley (1952) for the operation “Unloading 

rods from truck with fork lift.” On the one hand, the activity concentration derived for GSI is 

conservative in that it is based on the assumption that the deposition is entirely from 5 μm 

AMAD particles, as discussed in section 3 of the present memo. However, it could increase by a 

factor of 2 if we were to assume, as did Allen and Glover (2007), that the uranium handling 

operations took place during only one-half of the hours authorized by the MCW POs, the 

remainder of the time being spent on radiographing the uranium metal. Since the higher 

concentration would be offset by a shorter exposure duration, such an assumption would not 

change the uranium intakes and resulting internal doses, but it would affect the comparison of the 

derived concentration with data from other sites. 

Next, we compare the calculated uranium surficial activity concentrations with those used by 

Allen and Glover (2007). These authors derived a maximum value of 1,170 dpm/100 cm2

, which 

they assigned to all time periods following July 1, 1961. By contrast, we derived values ranging 

from 27 dpm/100 cm2

 during 1993, prior to the FUSRAP cleanup, to a high of about 43,000 

dpm/100 cm2

 during 1962. Finally, we note that Allen and Glover listed intakes ranging from 

0.932 dpm per calendar day during the entire residual period to about 128 dpm per calendar day 

in 1962. Our model results in intakes ranging from 0.29 dpm per calendar day at the end of the 

residual period to a high of 971 dpm per calendar day in 1957. For the entire 40.5 years from 

1953 to mid 1993, Allen and Glover’s average inhaled intake is about 29 dpm per calendar day, 

vs. an average of 261 dpm per calendar day calculated with our alternate model. 

5 Conclusion 

We have presented an alternate model for the assessment of uranium intakes at GSI that does not 

employ surrogate data derived from measurements at other sites. The model produces results 

that are plausible and claimant favorable, and are in a form that is suitable for use by NIOSH in 

performing dose reconstructions. 

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References 

ABRWH Work Group on Use of Surrogate Data. 2010. “Criteria for the Use of Surrogate Data,” 

Final Draft (May 14, 2010). 

http://www.cdc.gov/niosh/ocas/pdfs/abrwh/proc/abrwh-proc-sd-r0.pdf. 

Abu-Eid, R. M., et al. 2002. “Re-evaluation of the Indoor Resuspension Factor for the Screening 

Analysis of the Building Occupancy Scenario for NRC’s License Termination Rule: Draft 

Report for Comment,” NUREG-1720. 

http://www.nrc.gov/reading-rm/doc-collections/nuregs/staff/sr1720/ 

Adley, F. E., W. E. Gill, and R. H. Scott. 1952. “Study of Atmospheric Contamination in the 

Metal Melt Building,” HW-23352 (Rev.); SRDB Ref. ID: 27354. 

Allen, D., and S. Glover. 2007. “Site Profiles for Atomic Weapons Employers That Worked 

Uranium and Thorium Metals - Appendix BB: General Steel Industries,” Battelle-TBD-6000, 

Appendix BB, Rev. 0. http://www.cdc.gov/niosh/ocas/pdfs/tbd/b-6000-apbb-r0.pdf. 

Anigstein, R. 2012. “Review of the Use of Surrogate Data for Estimating Intakes of Uranium at 

General Steel Industries.” http://www.cdc.gov/niosh/ocas/pdfs/abrwh/scarpts/sca-gsisd-r0.pdf 

Anigstein, R., and J. Mauro. 2012. “Reply to NIOSH Response to SC&A Memo Dated May 30, 

2012.” Memo to Advisory Board on Radiation and Worker Health Work Group on TBD-6000 

(June 11, 2012). http://www.cdc.gov/niosh/ocas/pdfs/abrwh/scarpts/sca-gsiid-061112.pdf 

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Archives of Industrial Health, 19, 540–565. 

Murray, M. E., and K. S. Brown. 1994. “Results of the Independent Verification Survey at the 

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Oak Ridge National Laboratory. 

Murray, M. E., and M. S. Uziel. 1992. “Results of the Radiological Survey at the New Betatron 

Building, Granite City Steel Facility, Granite City, Illinois (GSG002),” ORNL/RASA-91/8. Oak 

Ridge, TN: Oak Ridge National Laboratory. 

Nuclear Regulatory Commission (U.S.) (NRC). 2009. “General Steel Industries Byproduct 

Materials License for Cobalt-60, Budd Company Model 41706 Sealed Source not to exceed 1 

curie in a single source,” ADAMS acquisition number ML093480290. 

 2007. < > “Fwd: Overhead door / Old Betatron” 

September 27, 2007, personal e-mail to John Mauro and Robert Anigstein, SC&A, Inc., and 

Larry J. Elliott, NIOSH, CC: 

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S. Cohen & Associates (SC&A). 2008. “Review of ‘Site Profiles for Atomic Weapons 

Employers That Worked Uranium and Thorium Metals - Appendix BB: General Steel 

Industries,’ Battelle-TBD-6000, Appendix BB, Rev. 0.” 

http://www.cdc.gov/niosh/ocas/pdfs/abrwh/scarpts/sca-t1-24-r1.pdf 

Sharfi, M. M. 2012. “Dose Reconstruction During Residual Radioactivity Periods at Atomic 

Weapons Employer Facilities,” ORAUT-OTIB-0070, Rev. 01. 

http://www.cdc.gov/niosh/ocas/pdfs/tibs/or-t70-r1.pdf. 

Alternative Model for Uranium Intakes at GSI -13- SC&A –July 25, 2012
 

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