Response to “Summary Dose Estimates for GSI” -1- SC&A – October 6, 2013

NOTICE: This report has been reviewed for Privacy Act information and has been cleared for distribution.

However, this report is pre-decisional and has not been reviewed by the Advisory Board on Radiation and Worker 

Health for factual accuracy or applicability within the requirements of 42 CFR 82.

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

FROM: Robert Anigstein and John Mauro, SC&A

SUBJECT: Response to NIOSH White Paper: “Summary Dose Estimates for GSI” 

DATE: October 6, 2013

SC&A Response to NIOSH White Paper: “Summary Dose Estimates for GSI”

On August 21, 2013, NIOSH prepared a white paper entitled “Summary Dose Estimates for 

GSI” (Allen 2013b). The present memo is the SC&A response to the NIOSH white paper. Our 

comments are keyed to the topical headings in the white paper. We will limit our comments to 

areas of disagreement.

1 Uranium Intake Estimates

The uranium intake estimate postulates two source terms of uranium dust: 

1. Uranium dust that becomes airborne during uranium handling operations

2. Uranium that has settled to the floor and is resuspended

During the June 20, 2013 teleconference meeting of the Advisory Board’s Work Group on TBD 

6000, the work group, NIOSH, and SC&A agreed to use a fixed value of 68.7 dpm/m3as the 

airborne uranium activity concentration during uranium handling operations. The three parties 

also agreed that the annual duration of exposure to this concentration shall be equal to the 

duration of the uranium handling operations during each time period, based on the purchase 

orders from the Mallinckrodt Chemical Works (MCW). The earliest purchase order is for the 

period 3/1/1958 to 6/30/1958; there are no POs covering the period October 1, 1952 (the 

beginning of the covered period) until March 1, 1958. During the discussion, Jim Neton 

(Associate Director for Science) agreed that the maximum uranium handling hours during any 

one year, based on the POs, should be assigned to this period. This value, 437.5 h/y, for the 

period 7/1/1961–6/30/1962, should therefore be assigned for each year or fraction of a year, from 

November 1, 1952 (the beginning of the covered period) until February 28, 1958.1 Furthermore, 

since there is an MCW PO for the period 3/1/58–6/30/58, the annual hours for that period should 

be derived from that PO. However, Allen (2013b) used a value of 337.5 h/y for the entire period 

11/1/1952–6/30/1958, resulting in substantially lower intakes during this period. 

The NIOSH estimate of airborne uranium activity due to resuspension is based on the 

aforementioned airborne activity of 68.7 dpm/m3, a settling velocity of 0.00075 m/s, a settling 

time of 30 d, and a resuspension factor (RF) of 10-5 m

-1

. The settling velocity and settling time 

are based on TBD-6000 (Allen 2011). We agree that these are reasonable parameter values to 

apply to the betatron shooting room and, by extension, to the rest of the GSI facility.

Table 1 shows a comparison of the SC&A and NIOSH annual hours of uranium handling and the 

calculated inhaled activities from both uranium handling operations and resuspension during the 

entire period of AEC operation. As shown in the table, the SC&A estimate is 24% higher than 

 

1

 Dr. Neton referred to a rounded value of 400 h/y.
Response to “Summary Dose Estimates for GSI” -2- SC&A – October 6, 2013

NOTICE: This report has been reviewed for Privacy Act information and has been cleared for distribution.

However, this report is pre-decisional and has not been reviewed by the Advisory Board on Radiation and Worker 

Health for factual accuracy or applicability within the requirements of 42 CFR 82.

the NIOSH calculation during the first 51⁄2 years of the operational period and by a smaller 

amount during the next 4 months. The slightly lower value calculated by SC&A during the 

remainder of the operational period is most likely due to round-off error, since both analyses 

utilized the same values of input parameters.

Table 1. Intakes of Airborne Uranium by Inhalation During GSI Operational Period

Start Date End Date 

Hours/year Inhalation dpm/cal. day

Δ

SC&A NIOSH SC&A NIOSH

10/1/52 2/28/58 437.5 337.5 113.09 91.4 24%

3/1/58 6/30/58 375.0 337.5 98.97 91.4 8%

7/1/58 10/31/58 337.5 337.5 90.50 91.4 -1%

11/1/58 6/30/59 337.5 337.5 90.50 91.4 -1%

7/1/59 6/30/61 337.5 337.5 90.50 91.4 -1%

7/1/61 6/30/62 437.5 437.5 113.09 114.22 -1%

7/1/62 6/30/63 125.0 125.0 42.50 42.93 -1%

7/1/63 6/30/65 28.1 28.0 20.62 20.8 -1%

7/1/65 6/30/66 12.9 13.0 17.17 17.38 -1%

During the residual period, the resuspended dust is the only source of inhalation exposure. Allen 

(2013b) abruptly lowered the RF from 10-5to 10-6 m

-1

on the day that uranium operations ceased. 

As we stated in an earlier memo: 

We do not agree that an RF of 10-6 m

-1

is applicable to the residual period. As Allen 

(2013a) correctly pointed out, this RF value is applicable to aged activity in a 

quiescent setting. Although the contamination gradually aged, the setting—the 

betatron shooting room and other areas of GSI—was hardly quiescent. (Anigstein 

and Mauro 2013b)

We believe that, during the residual period, an RF of 10-5 m

-1

should be used to calculate the 

resuspended dust. This would result in a 10-fold increase in the inhaled activities during this 

period. We agree that the surficial contamination should be gradually decreased during the 

residual period, according to OTIB-0070 (Sharfi 2012, Table 4-2).

2 Summary of External Dose Estimates

2.1 Doses to Plant Personnel During the Radium Era

Allen (2013b) assigned doses from external photon radiation to nonadministrative plant 

personnel during the Radium Era. These annual doses were represented by a triangular 

distribution, with a lower limit of 6.279 rem, a mode of 9.69 rem, and a maximum of 12 or 

15 rem, depending on the year. The maximum was set to equal the AEC dose limit in effect at 

the time. However, we disagree with the assumption that the upper bound of the distribution 

should be reduced from 15 rem/y (not 15 R/y) to 12 rem/y on Jan. 1, 1959. The basis of the 

maximum dose is the statement on GSI’s AEC license application that the applicable AEC dose 
Response to “Summary Dose Estimates for GSI” -3- SC&A – October 6, 2013

NOTICE: This report has been reviewed for Privacy Act information and has been cleared for distribution.

However, this report is pre-decisional and has not been reviewed by the Advisory Board on Radiation and Worker 

Health for factual accuracy or applicability within the requirements of 42 CFR 82.

limits for each time period were never exceeded. As stated in our earlier memo (Anigstein and 

Mauro 2013a), 

Ten CFR 20 was first put into effect on February 28, 1957. At that time, the 

regulation limited doses to blood-forming organs, gonads, and the lens of the eye to 

300 mrem/week. This limit had been presented in NBS Handbook 41, dated March 

30, 1949 (NBS 1949). The handbook embodied the recommendations of the NCRP, 

which AEC had agreed to follow, first in regulating its own operations, and later in 

promulgating rules for AEC licensees. Since the AEC observed the limit of 

300 mrem/week (15 rem/y) since about 1949, this limit can assumed to have been 

followed by GSI during the early years of the operational period.

On January 1, 1961, a revised 10 CFR 20 went into effect. This rule limited doses to 

a maximum of 3 rem per calendar quarter to a worker if his prior exposure history 

was known and if his lifetime dose did not exceed 5 (N-18), where N is his age in 

years. Thus, a worker well past the age of 18 with a sufficiently low cumulative 

exposure could receive doses as high as 12 rem in any one year. Thus, the upper 

bound of the triangular distribution should be set at 15 rem per year during the period 

1953–1960, and reduced to 12 rem/year during 1961–62.

The net effect of our recommended change would be to increase the maximum dose by 25% 

during the years 1959–1960.

We also disagree with the NIOSH assumption that this distribution, based in part on the estimate 

of doses to a radiographer utilizing 226Ra sources, should be used to assign doses during 1963. 

GSI was granted an AEC license to use 60Co sources for radiography, and procured two such 

sources in May 1962, at which time the use of 226Ra was discontinued. The New Betatron 

Building was constructed in 1963 and began operations sometime during that year. We believe 

that scenarios based on exposures of plant personnel to stray radiation from this betatron should 

be the basis of external doses, beginning in 1963 through the end of the covered period.

2.2 Doses to Plant Personnel After the Radium Era

Allen (2013b) assigned doses from external photon and neutron radiation to nonadministrative 

plant personnel following the Radium Era. The doses are based on the bounding exposure 

scenario, that of a layout man working just outside the New Betatron Building, which yields a 

photon dose of 4.483 rem/y and a neutron dose of 148 mrem/y. We have a number of objections 

to Allen’s exposure analysis, which we discussed in detail in our previous memos (Anigstein and 

Mauro 2012, 2013a). In summary, “we . . . disagree with [Allen’s] arbitrarily selected set of 15 

betatron shooting scenarios, 10 of which do not represent realistic practices employed in betatron 

radiography.” Next, “we disagree with Allen’s . . . methodology of apportioning the shots 

among these 15 scenarios on the basis of the MCNPX analyses of exposure rates at the location 

of the film-badge storage rack.” As we stated earlier,

A question arose during the February 21, 2013, work group meeting regarding the 

betatron shooting scenarios. David Allen maintained that the betatron scenario 
Response to “Summary Dose Estimates for GSI” -4- SC&A – October 6, 2013

NOTICE: This report has been reviewed for Privacy Act information and has been cleared for distribution.

However, this report is pre-decisional and has not been reviewed by the Advisory Board on Radiation and Worker 

Health for factual accuracy or applicability within the requirements of 42 CFR 82.

proposed by SC&A was implausible because it would have led to exposures of 

control film badges in excess of 10 mR per week. We disagree with this assumption 

for two reasons. First, there is no information on where film badge No. 1, called 

“Betatron CTL,” was stored. Since there was only one such badge, it could have 

been kept in the Old Betatron Building, which is where Gillium Burgess, one of the

earlier betatron supervisors, had his office. The 000 control badge should have been 

stored on the film badge rack, along with all the workers’ film badges. Allen 

correctly pointed out that the reports of this badge always had readings of M. The 

manner in which the dose to this badge was evaluated is explained in the memo from 

Joseph Zlotnicki, CHP (former Landauer official, currently a member of the SC&A 

staff) (see Attachment 1). The evaluation of the badges was a multi-step process. 

Landauer retained a control film that was matched to each batch of films sent to a 

customer—this film never left its premises. When the films were returned from the 

customer, the in-house control film was developed alongside the other films. The 

base fog on this film was subtracted from the densitometer readings of all the other 

films. The remaining “dose” on the 000 customer control badge was evaluated. If 

this film read less than 50 mrem, and if the reading was lower than that of one-half 

of the badges issued to workers, the reading was subtracted from that of the other 

badges, as well as from itself. In these cases, which constituted the vast majority, the 

control badge would be reported as M. Thus, the M readings cannot be used to place 

an upper limit on the cumulative weekly exposure at the film badge location. 

Furthermore, as we pointed out earlier (Anigstein and Mauro 2012), we disagree 

with the NIOSH MCNP model of the New Betatron Building, which used incorrect 

assumptions about the thickness and density of the control room wall and the 

absence of equipment, furniture, and internal walls that would have reduced the 

exposure rate at the film badge rack. We therefore restate our opinion that the 

betatron shooting scenario described in our earlier report (Anigstein and Olsher 

2012) is more realistic and more claimant favorable than the one proposed by 

NIOSH. (Anigstein and Mauro 2013a)

Table 2 lists the exposures to penetrating external radiation to the layout man, which are 

applicable to all nonadministrative plant personnel from 1963 to June 30, 1966. In both the 

SC&A and NIOSH analyses, the radiation exposure of the layout man constitutes the bounding 

scenario for photon exposure. In the NIOSH analysis, this is also the bounding scenario for 

neutron exposure.

As shown in Table 2, our estimated photon exposure is more than twice as high as the NIOSH 

estimate. We note that the results of the SC&A analysis are presented in units of roentgens, 

while the NIOSH results are listed as rem. However, we note that in an earlier report, Allen 

(2012) listed a value of 4,483 mr for this quantity, which is equal to 4.483 R. He apparently uses 

the two sets of units interchangeably. Our estimated neutron dose is more than three times 

higher than the NIOSH estimate. We note that, according to our analysis, the limiting scenario 

for neutron exposures during the period 1952–1963 is the radiation exposure of the betatron 

operator, whereas it is that of the layout man in the NIOSH analysis. However, the differences in 

neutron doses in the two SC&A scenarios are slight. 
Response to “Summary Dose Estimates for GSI” -5- SC&A – October 6, 2013

NOTICE: This report has been reviewed for Privacy Act information and has been cleared for distribution.

However, this report is pre-decisional and has not been reviewed by the Advisory Board on Radiation and Worker 

Health for factual accuracy or applicability within the requirements of 42 CFR 82.

Table 2. Annual Doses to Layout Men

Photon exposure Neutron dose (rem)

Beta dose to skin (rads)

Hands and forearms Other skin

SC&A (R) NIOSH (rem) SC&A NIOSH SC&A NIOSH SC&A NIOSH

9.20 4.483 0.46 0.148 4.20 2.658 2.45 1.462

Note: SC&A values from Anigstein and Olsher (2012), NIOSH values from Allen (2013b)

In both the SC&A and NIOSH analyses, the radiation exposure of betatron operator constitutes 

the bounding scenario for dose to the skin from beta rays. According to Allen (2013b), the beta 

skin dose analyses were performed using MCNPX Version 2.7. We used the same version of the 

code in our analyses; as shown in Table 3, the two analyses produced significantly different 

results. Since we have not seen the NIOSH MCNP files, we cannot explain the discrepancies. 

We do observe that the differences increase steadily during the years 1961–1966. We further 

observe that, since the annual exposure duration from uranium handling decreases steadily 

during these years, the handling of irradiated steel accounts for a steadily increasing fraction of 

the skin dose. It seems likely that differences in modeling the exposure to irradiated steel might 

account for at least some of the differences in the results.

Table 3. Annual Doses to Betatron Operators

Year

Neutron dose

(rem)

Beta dose to skin (rads)

Hands and forearms Other skin

SC&A NIOSH SC&A NIOSH Δ (%) SC&A NIOSH Δ (%)

10/1/1952-1957 0.48 0.050 33.4 26.904 24% 6.27 2.755 127%

1958 0.48 0.050 32.1 26.904 19% 6.22 2.755 126%

1959-1960 0.48 0.050 30.9 26.904 15% 6.18 2.755 124%

1961 0.48 0.056 34.2 30.496 12% 6.30 2.946 114%

1962 0.48 0.043 27.2 22.863 19% 6.04 2.539 138%

1963 0.47 0.019 13.9 8.154 70% 5.56 1.755 217%

1964 0.46 0.013 10.7 4.669 129% 5.45 1.569 247%

1965 0.46 0.012 10.2 4.130 147% 5.43 1.541 252%

1966a0.23 0.006 4.8 1.796 170% 2.71 0.756 258%

Note: SC&A values from Anigstein and Olsher (2012)

a

During contract period: January 1–June 30

3 Conclusions

The work group, SC&A, and NIOSH have reached consensus on several issues related to the 

assessment of radiation exposures of GSI workers. However, there are significant areas of 

disagreement between NIOSH and SC&A that remain to be resolved. 
Response to “Summary Dose Estimates for GSI” -6- SC&A – October 6, 2013

NOTICE: This report has been reviewed for Privacy Act information and has been cleared for distribution.

However, this report is pre-decisional and has not been reviewed by the Advisory Board on Radiation and Worker 

Health for factual accuracy or applicability within the requirements of 42 CFR 82.

3.1 Uranium Intake Estimates

We are in complete agreement with the methodology to estimate uranium intakes during the 

operational period. We disagree about the duration of uranium handling operations during the 

period 10/1/52–6/30/58. The higher annual duration recommended by SC&A, especially during 

10/1/52–2/28/58, would result in a 24% increase in inhaled activities during this period.

We do not agree with the NIOSH estimate of inhaled activities during the residual period, which 

is based on an RF of 10-6 m

-1

. Such an RF applies to a quiescent environment after 

decontamination, neither of which conditions necessarily apply to the betatron buildings nor the 

rest of the GSI facilities, especially during the period 7/1/1966–12/31/1973, since steelmaking 

activities continued until near the end of this period and the betatron buildings were still being 

used to radiograph steel castings. A higher RF would lead to proportionately higher inhaled 

activities.

3.2 Doses to Administrative Personnel from External Exposures

We are in complete agreement with NIOSH about the assignment of photon exposures of 

administrative personnel, with the understanding that the resulting doses shall be assigned only if 

it can be conclusively established that the employee performed administrative functions, that the 

employee’s work station was remote from the production areas of the plant, and that the 

employee did not frequently enter the production areas. 

3.3 External Exposures of Plant Personnel During the Radium Era

We are in agreement on estimating annual doses from external exposure of nonadministrative 

personnel based a triangular distribution, with a lower limit of 6.279 rem, a mode of 9.69 rem, 

and a maximum of 12 or 15 rem, depending on the year. However, we believe that upper bound 

of the distribution should be reduced from 15 rem to 12 rem starting in 1961, not 1959, since the 

lower AEC limits did not take effect until January 1, 1961. Furthermore, we believe that the 

6.279, 9.69, 12 rem distribution should apply through 1962, not 1963, since the Radium Era 

ended with GSI’s acquisition of 60Co sources in May 1962.

3.4 External Exposures of Plant Personnel from the Operation of the New Betatron

We believe that the limiting photon exposures of nonadministrative personnel to radiation during 

the period 1963–June 30, 1966 should be based on the exposure of the layout man to stray 

photon radiation from the New Betatron—NIOSH applies doses derived from this scenario to the 

period 1964–June 30, 1966. We disagree with the NIOSH analysis of this scenario: our derived 

doses are more than twice as high as those calculated by NIOSH. Our analysis shows that the 

limiting neutron exposures are of the betatron operator, who is estimated to have received 

slightly higher neutron doses than the layout man. Our calculated neutron doses from either 

scenario are over three times as high as the maximum neutron dose estimated by NIOSH.
Response to “Summary Dose Estimates for GSI” -7- SC&A – October 6, 2013

NOTICE: This report has been reviewed for Privacy Act information and has been cleared for distribution.

However, this report is pre-decisional and has not been reviewed by the Advisory Board on Radiation and Worker 

Health for factual accuracy or applicability within the requirements of 42 CFR 82.

3.5 Skin Doses from Beta Rays

Based on our analyses, we agree with NIOSH that the exposures of the betatron operator to beta 

radiation constitute bounding scenarios for doses to the skin. However, our calculated doses 

range from 12% to 258% higher than those listed by Allen (2013b) for corresponding areas of 

skin and corresponding time periods.
Response to “Summary Dose Estimates for GSI” -8- SC&A – October 6, 2013

NOTICE: This report has been reviewed for Privacy Act information and has been cleared for distribution.

However, this report is pre-decisional and has not been reviewed by the Advisory Board on Radiation and Worker 

Health for factual accuracy or applicability within the requirements of 42 CFR 82.

References

Allen, D. 2011. “Site Profile for Atomic Worker Employers that Worked Uranium Metals,” 

Battelle-TBD-6000 Rev. 1. http://www.cdc.gov/niosh/ocas/awedocs.html#b6000.

Allen, D. 2012. “Battelle-TBD-6000 Appendix BB General Steel Industries: Addendum to Dose 

Estimates for Betatron Operations.”

Allen, D. 2013a. “GSI White Paper: Issues Raised in February 21, 2013 Work Group Meeting.” 

http://www.cdc.gov/niosh/ocas/pdfs/dps/dc-gsiwgiss-0413.pdf

Allen, D. 2013b. “White Paper: Summary Dose Estimates for GSI.” 

http://www.cdc.gov/niosh/ocas/pdfs/dps/dc-gside-0813.pdf

Anigstein, R., and J. Mauro. 2012. “Review of ‘Addendum to Dose Estimates for Betatron 

Operations White Paper’.” Memo to Advisory Board on Radiation and Worker Health Work 

Group on TBD-6000.

http://www.cdc.gov/niosh/ocas/pdfs/abrwh/scarpts/sca-comts-nbtronadd-032512.pdf

Anigstein, R., and J. Mauro. 2013a. “Review of NIOSH Estimates of External Exposures at 

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

http://www.cdc.gov/niosh/ocas/pdfs/abrwh/scarpts/sca-gsiextexp-042313.pdf

Anigstein, R., and J. Mauro. 2013b. “Review of NIOSH Estimates of Internal Exposures at GSI.” 

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

http://www.cdc.gov/niosh/ocas/pdfs/abrwh/scarpts/sca-gsiintexp-042413.pdf

Anigstein, R., and R. H. Olsher. 2012. “Response to ‘Battelle-TBD-6000 Appendix BB General 

Steel Industries: Dose Estimates for Betatron Operations.’”

Atomic Energy Commission (U.S.) (AEC). 1957. “Title 10—Atomic Energy; Chapter 1—

Atomic Energy Commission: Part 20—Standards for Protection Against Radiation. Federal 

Register, 22(19), 548–554.

Atomic Energy Commission (U.S.) (AEC). 1960. “Title 10—Atomic Energy; Chapter 1—

Atomic Energy Commission: Part 20—Standards for Protection Against Radiation. Federal 

Register, 25(224), 10914–10924.

National Bureau of Standards (NBS), Department of Commerce (U.S.) 1949. “Medical X-ray 

Protection up to Two Million Volts: Handbook 41.” 

http://www.orau.org/ptp/Library/NBS/NBS-41%20Medical%20X-Ray2.pdf

National Committee on Radiation Protection (NCRP). 1958. “Maximum Permissible Radiation 

Exposures to Man, Addendum to National Bureau of Standards Handbook 59, Permissible Dose 

from External Sources of Ionizing Radiations,” Recommendations of the National Committee on 

Radiation Protection. Washington, D.C.: U.S. Department of Commerce.
Response to “Summary Dose Estimates for GSI” -9- SC&A – October 6, 2013

NOTICE: This report has been reviewed for Privacy Act information and has been cleared for distribution.

However, this report is pre-decisional and has not been reviewed by the Advisory Board on Radiation and Worker 

Health for factual accuracy or applicability within the requirements of 42 CFR 82.

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.
Response to “Summary Dose Estimates for GSI” A-1 SC&A – October 6, 2013

NOTICE: This report has been reviewed for Privacy Act information and has been cleared for distribution.

However, this report is pre-decisional and has not been reviewed by the Advisory Board on Radiation and Worker 

Health for factual accuracy or applicability within the requirements of 42 CFR 82.

Attachment 1

To: Bob Anigstein, SC&A

From: Joe Zlotnicki 

Date: Tuesday, March 05, 2013

Subject: GSI and Landauer “Control” dosimeters

This memo serves to document recent discussions and email correspondence that I had with you 

and with Landauer personnel regarding General Steel Industries and the methodology Landauer 

used to manage control dosimeters. 

On 2/21/13 you wrote to me with the following questions that documented our prior phone call:

1. Please clarify that the dose derived for the unnumbered control badge dose is 

subtracted from all other film badges, including itself, which is why it is almost always 

reported as "M." It is the actual derived (but unreported) dose that's subtracted, not the 

M. Please explain the reason for this practice.

2. Best guess regarding the identity of the "BETATRON CTL" badge, badge #001. This 

badge disappeared in the reports from about March 1966. 

3. Historical practices 1963-1966.

I spoke with Dr. Craig Yoder at Landauer on 2/26/13 and the following description of historical 

film practices at Landauer is based on my knowledge and that discussion:

Underlying information common to film technology

Film dosimetry was complex and there were a number of industry-wide practices and Landauerspecific procedures that were used to address shortcomings in the technology. It is important to 

understand some of these issues in trying to understand specific practices with Control 

dosimeters. In addition, nomenclature varied within the industry and differs from other 

laboratory practices. For example, spike, control, process control, QC, blank and calibrate need 

to be viewed with caution as they can have special meanings in the Landauer or industry 

vernacular.

Film was purchased in batches and developed an increase in base fog (optical density, or O.D.) 

over time. This increase in base optical density is due to inherent chemical/thermal and radiation 

processes. Each batch (or emulsion) had a unique tracking number and all processes and 

procedures were based both on the emulsion and unique chemical processing batch when the 

dosimeter was developed. 
Response to “Summary Dose Estimates for GSI” A-2 SC&A – October 6, 2013

NOTICE: This report has been reviewed for Privacy Act information and has been cleared for distribution.

However, this report is pre-decisional and has not been reviewed by the Advisory Board on Radiation and Worker 

Health for factual accuracy or applicability within the requirements of 42 CFR 82.

The nature of film technology is that the sensitivity of the film plus densitometer has a minimum 

step function of about 6 mR (at Cs-137 energy). This corresponds to a change of 0.02 O.D. 

Thus the reading process is quantized in increments of 6, 12, 18 mR, etc. This is important when 

considered in tandem with Landauer’s rounding and definition of “M” practices that are 

discussed below.

The densitometer sensitivity was typically ±0.02 optical density while the base fog was about 

0.3–0.4. It is clear that statistical swings on the order of 10mR can be caused by small variations 

from film to film and in chemical processing, etc. It was only with great care and consistent 

practices that 10mR was utilized as the minimum reportable dose. As the film aged (say 4–

6 months from manufacture) it became increasingly difficult to achieve the minimum 

detectable/reportable. (It is outside the scope of this discussion, but as photon energy has a very 

large impact on film sensitivity, lower energy x-rays were easily detected at this level whereas 

high energy gamma was probably only detectable at the 10-mR level under optimum 

circumstances.)

Landauer Procedures

To correct for base fog, Landauer had several procedures and practices. First, film was always 

identified by emulsion (batch). Blanks, controls and calibration film were all selected from the 

same emulsion for a given client film dosimeter. This way, the underlying aging process would 

largely be the same for in-house Landauer process control dosimeters and client dosimeters. Of 

course, storage conditions were not identical but at least the film had the same age and 

underlying properties. Note that film had a relatively short shelf life due to base fog build up and 

could not be re-zeroed as OSL and TLD can be.

As part of the process for calculating dose, Landauer subtracted the base fog density from all 

dosimeters in a group of film by utilizing process blanks. Calibration spikes and customer 

controls had this blank value subtracted prior to use in defining the dose response curve and 

process zero for the densitometers. In a normal situation, this meant that client controls that were 

stored in a low background would report as Minimal, or effectively zero.

Control badges were issued to clients to address several concerns. Primarily, they were to 

ascertain if the dosimeters were exposed in transit going to or from the client site. For larger 

clients, several controls were sent with the outbound shipment to ensure that spare controls were 

available when late badges were returned. If everything worked normally, then fog and 

background subtraction were addressed by Landauer in-house blanks that were subtracted from 

ALL badges including client control badges. If any residual “dose” remained on the client 

control (presumably due to transit exposure or elevated natural background) then this was 

subtracted from the batch of client badges. Effectively, the Control badge was set to zero.

If there were a slight difference in dose between in-house base fog blanks and client badges, then 

it is possible that a plus or minus 6 mR would be measured on a given badge. However, negative 

doses were not recorded as such and were assigned a zero. Close oversight was maintained on 
Response to “Summary Dose Estimates for GSI” A-3 SC&A – October 6, 2013

NOTICE: This report has been reviewed for Privacy Act information and has been cleared for distribution.

However, this report is pre-decisional and has not been reviewed by the Advisory Board on Radiation and Worker 

Health for factual accuracy or applicability within the requirements of 42 CFR 82.

the statistical batch of film and the distribution of net optical densities was tracked to look for 

anomalies.

The procedure for defining a “Minimal” dose is important and potentially confusing. All 

measured client doses were rounded down at 9 mR or less. Thus a 9 mR becomes a Minimal 

whereas a 10-14 becomes a 10 mR. 15-24 becomes 20 mR, etc. So the unusual rounding 

practice only applies to the less than 10mR reporting. This process applies to all badges 

including Control badges. If the Control showed a dose in excess of 50 mR, then a message 

would have been printed on the report indicating that fact. This was a rare occurrence for any 

client.

Landauer Numbering Protocol

All Control badges in the Landauer system were given a “0000” I.D. Any other number was not 

a Control, regardless of the name on the badge. Some clients put “Control” or “CNTL” in the 

name field for their own purposes, most commonly to monitor the x-ray or accelerator control 

room as an area monitor. However, the Landauer computer system was programmed to only use 

a 0000 designation as a Control. Any other badge would have been reported normally after 

Control subtraction.

Please let me know if you have any additional questions or concerns,

Sincerely,

Joe Zlotnicki CHP

President, 3Z Consulting Inc