This document may contain information covered by the Privacy Act 

1 

NIOSH Response to ABRWH Request for Supplemental 

Information Related to the Rocky Flats SEC Petition 

May 17, 2007 

Executive Summary: 

On May 4, 2007, the Advisory Board on Radiation Worker Health (ABRWH) requested that 

NIOSH provide the following supplemental information related to the Rocky Flats SEC petition: 

1) Thorium Issue – SC&A has concluded that the NUREG-1400 approach is not appropriate 

or bounding. NIOSH contends that they have other process specific data that could be 

used to bound worker doses. NIOSH needs to demonstrate this by documenting this new 

approach and completing example dose reconstruction(s). 

2) Building 881 – There was no Building 881 external monitoring data in the 1950’s. 

NIOSH has provided information about the processes along with the data from the early 

1960’s suggest that their coworker model may be used to bound gamma and beta doses 

for Building 881 workers. NIOSH needs to demonstrate this by documenting this new 

approach and completing example dose reconstruction(s). In addition, the possibility of 

plutonium exposures in this building needs to be addressed. 

3) Neutron Doses 1959 to 1970 – The current NIOSH approach relies on application of a 

central estimate of a building specific neutron photon ratio to estimate doses. The work 

group has remaining questions whether this approach will be bounding for all workers. 

NIOSH has additional data that may be used to estimate a bounding neutron photon ratio 

which could then be applied to bound worker doses during this time period. NIOSH 

needs to demonstrate this by documenting this new approach and completing example 

dose reconstruction(s). 

The enclosed report has been prepared to respond to these requests. Supporting reference 

documentation has been provided at: 

O:\Document Review\AB Document Review\Rocky Flats\ABRWH 5-4-07 request\ 
This document may contain information covered by the Privacy Act 

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I. ABRWH Request 1 

“Thorium Issue – SC&A has concluded that the NUREG-1400 approach is not appropriate or 

bounding. NIOSH contends that they have other process specific data that could be used to 

bound worker doses. NIOSH needs to demonstrate this by documenting this new approach and 

completing example dose reconstruction(s)”. 

I.A. Background 

Briefly, the following operations involving nontrivial quantities of thorium were conducted at 

Rocky Flats: 

1. On seven occasions between March 1, 1962 and June 30, 1966, Rocky Flats received 

thorium metal parts originally fabricated at Y-12. These parts were used in weapons 

models. NIOSH originally considered the possibility that these parts could have been 

lightly machined at Rocky Flats, however subsequent interviews with five R&D 

machinists (included in this report) indicated this did not occur. Therefore, there is no 

airborne potential from handling these parts. 

2. In 1960, three thorium ingots (approximately 80 kg each) were canned and pressed into 

desired shapes. A very detailed account of this operation has been located and provided 

to the Working Group (Calabra, 1961). Further detail was provided in the health physics 

logbooks covering this operation, including a list of personnel involved. In total, the 

project involved approximately 38 hours over a total of 8 working days. Limited 

urinalysis and air sampling was conducted during the operation, which indicated no 

internal exposures. 

3. Removal of U232 daughters (including Th228) from U233. This involved 20 kg of 

U233nitrate, which contained 47 ppm U232 as a contaminant. A very detailed account of 

this operation has been located and provided to the Working Group (Kirchner and 

Freiberg, 1965). Further detail was provided in the health physics logbooks covering this 

operation, including a list of personnel involved. This thorium strike was conducted on 

April 26-28, 1965. Air sampling was conducted during the operation. The health physics 

logbooks covering the operation, air sampling, and statements from the health physicist 

involved indicate that there were no releases during this operation. 

The Rocky Flats Working Group has conducted extensive discussions on the limited thorium 

activities at Rocky Flats, supported by detailed investigations by NIOSH. On two occasions 

(March 7, 2007 and April 19, 2007), the Working Group indicated that the thorium issue 

constituted a TBD issue, rather than a SEC issue. The requirement for proof of principle was 

also explicitly discussed at these Working Group meetings, and the Working Group declined to 

request example dose reconstructions, indicating that the extensive weight of the evidence that 

had been assembled would suffice. 

NIOSH originally proposed to use NUREG-1400 to bound the doses for these operations, as was 

proposed and accepted for small operations at Y-12. SC&A objected to the use of NUREG-1400 

at Rocky Flats unless validation was provided that this methodology provided bounding intakes. 

NIOSH provided the requested validation using air data from Simonds Saw and Steel, and the 

torch cutting operation (part of the 1960 ingot project) at Rocky Flats compared to intakes 

predicted by NUREG-1400 for these operations, as suggested by SC&A. However, SC&A 
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3 

raised objections to these comparisons, and continued to object to the application of NUREG1400. The Board appears to have adopted SC&A’s position on this issue on May 4, 2007. At 

that time, the Board requested that NIOSH provide bounding dose reconstructions using other 

process specific data. Therefore, NIOSH has produced two example dose reconstructions, one 

for each of the operations which had hypothetically nontrivial airborne exposure potential at 

Rocky Flats (the 1960 ingot operation, and 1965 thorium strike). 

I.B. Interviews on machining/trimming of thorium metal parts at Rocky Flats

Telephone interviews conducted on January 16th and 17th, 2007 with five former Rocky Flats 

personnel and all were asked the following questions: 

1. When did you start and how long did you work at Rocky Flats? 

2. What were your responsibilities? 

3. What can you tell us about the machining and or trimming of thorium parts at Rocky 

Flats? 

 

[Name Withheld]

[Name Withheld] started in 1953 as a chemical operator. He was involved with rolling, forming, 

and machining operations at Buildings 776, 881, and 883. He was [Withheld] of building 444 

from 1964 to 1986. In building 444 he was responsible for all rolling, forming and machining 

operations. During his tenure at Building 444 there was only beryllium and depleted uranium 

being machined. He was also involved with R&D and “special orders” and would have known if 

thorium was being machined. He clearly stated that to the best of his recall, there was no 

machining or trimming of thorium metal. When asked if there was hypothetically thorium 

machining done at RF how would that have been accomplished? He said all 

radioactive/pyrophoric metal with resultant turnings or chips would be been covered by a waterbased oil emulsion. This would have minimized dispersion of chips and minimized potential 

fires. 

[Name Withheld]

[Name Withheld] started in 1963 and worked [Withheld] years at Rocky Flats until [Withheld]. 

He left Rocky Flat for 2 years beginning in 1964 to attend [Withheld] and [Withheld] to get his 

masters degree. His master’s thesis was [Withheld]. He is currently a consultant to LANL on 

machine production and gauging for new pit production and still holds an active “Q” clearance. 

While at Rocky Flats he was the [Withheld] for “non-traditional” machining. His responsibilities 

were R&D on machining techniques like electrochemical and electrical discharge machining. He 

clearly stated that in his position he never saw a piece of thorium metal at Rocky Flats. He said if 

he knew that some thorium was present, he would have gone to look because of his interest. His 

primary work was in Building 881 which he stated went “cold” in 1965. When asked what would 

be meant by “light machining” he stated at Rocky Flats that would have been no more than 

removal of 10 mils or less of material. He has excellent recall of R&D machining activities at 

Rocky Flats. 

[Name Withheld]

[Name Withheld] started in 1953 worked [Withheld] years until [Withheld]. He was the liaison 

between R&D and production in the uranium facilities (883, 881, 444). He did not recall any 
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machining of thorium. When asked what would be meant by “light machining”, he clearly stated 

that “it would be the minimal removal of material 30 mils of less and gram quantities removed”. 

He also said that if there were significant quantities thorium present, he would have known about 

it and would have remembered. 

[Name Withheld]

[Name Withheld] started in1948 at Hanford and came to Rocky Flats in 1968 as a machinist. He 

worked for [Withheld] years at Rocky Flats. He was a supervisor and a General Foreman for 

production and in 1975 became the [Withheld]. In 1992 He was the [Withheld]. He had no recall 

that any machining of thorium metal took place at Rocky Flats. 

[Name Withheld]

[Name Withheld]started in June, 1956 and worked at Rocky Flats until [Withheld] as machinist 

in the R&D area. He also had no recall of any machining of thorium at Rocky Flats. 

Summary of conclusions with regard to machining of thorium parts at Rocky Flats 

The ChemRisk Task 3/4 report clearly states that “There was light production of thorium parts in 

Building 881 in the 1950s to early 1960s. The report also states the “The major use has been 

fabrication of metal parts from natural thorium” and was also used as a “stand in” for the more 

expensive U or Pu components in various phases of development programs. It is not explicitly 

stated in the report itself whether “light production of thorium parts” refers to the ingot operation 

in 1960, or to some machining on the preformed Th metal parts received from Y-12. The 

interviews above indicate the former, as none of the workers interviewed recalls any machining 

of thorium parts. Furthermore, the handling of prefabricated parts from Y-12 occurred between 

1962 and 1966, rather than early in the 1960s, which also indicates that the quote refers to the 

ingot operation. 
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I.C. Example dose reconstructions 

I.C.1 1960 ingot operation 

Step 1: cold-rolling of bare thorium ingots 

Step 2: other machining (pressing, canning, cutting, etc.) 

Step 3: decanning by flame-cutting with a torch 

Step 1 (rolling) intake calculation: 

The most directly-related data for use in the intake calculation comes from the Monthly Progress 

Report–Site Survey-July, 1960. This document reports the results of sampling taken during 

rolling, which observed 4.62 dpm/m3

. SC&A has noted that this data could be general air 

sample results, and has suggested that this is not appropriate for use in dose reconstruction. 

SC&A further suggested that data from Albert, 1960 be used instead. The Albert reference 

provides a measured breathing zone air concentration of 7.14x10-11 μCi/ml (2.64 Bq/m3) for 

“rolling billet to slab”. This value is for samples held for two weeks to allow for decay of shortlived airborne daughter species which would have negligible contribution to the dose. This value 

is approximately 35 times higher than the sample result collected during the rolling at Rocky 

Flats, therefore, NIOSH is hopeful that this can be agreed to be bounding. Using the Albert data 

as suggested by SC&A, the resulting intake for this step of the operation would be: 

I = (2.64 Bq/m3

) x (1.2 m3/hr) x (8 hr) = 25.3 Bq 

Step 2 (other machining) intake calculation: 

The most directly-related data for use in the intake calculation comes from the Monthly Progress 

Report–Site Survey-July, 1960. This document reports the results of sampling taken during 

“other operations”, which observed 1.41 dpm/m3

. SC&A has noted that this data could be 

general air sample results, and has suggested that this is not appropriate for use in dose 

reconstruction. SC&A further suggested that data from Albert, 1960 be used instead. The Albert 

reference provides a measured breathing zone air concentration of 1.70x10-11 μCi/ml (0.63 

Bq/m3

) for “lathe enclosed in hood”. Interviews with site experts indicated that the equipment 

used for the ingot operation was the same equipment used for handling EU, which consisted of 

shrouded/hooded lathes, therefore this data would be appropriate. This value is for samples held 

for two weeks to allow for decay of short-lived airborne daughter species which would have 

negligible contribution to the dose. This value is approximately 27 times higher than the sample 

results collected the machining at Rocky Flats, therefore, NIOSH is hopeful that this can be 

agreed to be bounding. Using the Albert data as suggested by SC&A, the resulting intake for 

this step of the operation would be: 

I = (0.63 Bq/m3) x (1.2 m3/hr) x (30 hr) = 22.7 Bq 

Step 3 intake calculation: 

The most relevant data for calculating an intake from the decanning operation is the air sample 

taken at 3 feet from the ingot while it was being cut. In SC&A’s April 5, 2007 report, NIOSH’s 

characterization of this sample as a breathing zone sample is questioned, and it is suggested that 

the intake be calculated instead by relying on welding data from SC&A’s Bethlehem Steel 

analysis. NIOSH considers the suggested approach inadvisable for the following reasons: 
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1. the can was being removed from the ingot by flame-cutting (of the can) with a torch. It is 

not intuitively obvious to NIOSH why SC&A believes that a worker would position his 

face closer than 3 feet to the canned ingot during this operation, as doing so would 

present the very real possibility of severe burns. NIOSH continues to conclude that this 

sample is representative of the breathing zone experienced by the torch operator. 

2. It is not intuitively obvious why welding data, which uses thoriated tungsten alloy 

electrodes rather than thorium metal used in the ingot operation, would be more 

appropriate than the samples from the actual operation itself. NIOSH investigated 

welding data and the predicted intakes were lower than for the process-specific samples. 

The intake estimate below is based on the actual air sample taken at 3 feet from the ingot while 

the can was being removed with a torch. 

I = (62 dpm/m3

) x (1.2 m3/hour) x (2 hour) x (0.0167 Bq/dpm) = 2.5 Bq 

Total intake:

The total intake from steps 1-3 above is 50.5 Bq. Since the identities of the individuals involved 

in the various steps of the ingot operation are known, and it is known that individual workers 

were not involved in all steps, an intake estimate reflecting the total of all steps would be 

bounding for any individual worker. 

Selection Criteria

• Hypothetical Thorium Ingot Worker: worked 1959 through 1970 and was exposed 

to thorium in 1960. 

• Unmonitored for thorium 

Cancer Description:

Bone (ICD-9 170.0) 12/31/2000 

Colon (ICD-9 153.9) 12/31/2000 

Lung (ICD-9 162.0) 12/31/2000 

Prostate (ICD-9 185.0) 12/31/2000 

Employment (Rocky Flats Plant)

Start: 1959 

End: 1970 

Work History

 NOCTS : DOB: 1930, 

Diagnosis Date: 12/31/2000 

Former Smoker 

White-Non-Hispanic 

Dosimetry Data: No thorium dosimetry data. 

Narrative
This document may contain information covered by the Privacy Act 

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Internal dose is caused by radioactive materials that are taken into the body. A chronic intake is 

an intake of radioactive material that occurs over an extended period of time (typically weeks or 

longer). An acute intake is an intake of radioactive material that occurs over a short period of 

time (typically minutes to hours). Regardless of the rate at which the intake occurs, the internal 

dose received from radioactive materials having long half-lives occurs over an extended period 

of time and is, therefore, considered chronic. 

The internal dose to the prostate was determined by using the dose calculated for the highest 

non-metabolic organ.3

 The organ/tissue associated with this cancer is not included in the ICRP 

modeling of internal doses; so in accordance with NIOSH documentation, the largest dose to an 

exposed organ that is not described by the ICRP metabolic models was assigned as the 

appropriate internal dose (in this case, the adrenals). 

A computer code, the Integrated Modules for Bioassay Analysis (IMBA), was used to estimate 

intakes of radioactive material and the subsequent annual organ doses. The IMBA Expert 

ORAU-Edition was used for this dose reconstruction. The ICRP 66 lung model with default 

aerosol characteristics was assumed, in conjunction with ICRP 68 metabolic models. It should 

be emphasized that intake dates, scenarios, and intake levels were based upon mathematical 

models and do not necessarily prove that such intakes occurred on the given dates. These dates 

and scenarios provide an acceptable explanation of exposure and dose based upon the bioassay 

data provided. This approach is in accordance with the provisions of the Radiation Dose Reconstruction Rule (42 CFR 82)1

 and guidance in the NIOSH Internal Dose Reconstruction Implementation Guideline.2 

The energy employee assisted in the processing of thorium ingots, according to records received 

from the Department of Labor and information provided in the interview process. Employment 

records were reviewed, and no records of thorium monitoring were found. Based on air 

concentration data from similar thorium operations4

 an acute intake of 50.5 Bq of thorium-232 

was assigned on January 1, 1960. 

Summary

The doses below only capture thorium internal dose. These doses do not include external, 

medical, environmental, or internal doses from other radionuclides. 

Bone: 6.214 rem 

Colon: 0.018 rem 

Lung: 61.079 rem 

Prostate: 0.017 rem 

These dose estimates were entered into IREP with a constant distribution. 

Probability of Causation (POC)

These probably of causations are based on the dose from thorium internal dose only. These 

probably of causations do not include external, medical, environmental, or internal doses from 

other radionuclides. 
This document may contain information covered by the Privacy Act 

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Bone: 12.85% 

Colon: 0.04% 

Lung: 0.82% 

Prostate: 0.03% 

References

1. 42 CFR 82, Methods for Radiation Dose Reconstruction Under the Energy Employees 

Occupational Illness Compensation Program Act of 2000; Final Rule, Federal 

Register/Vol.67, No. 85/Thursday, May 2, 2002, p 22314, SRDB Reference ID 19392. 

2. NIOSH, (2002) Internal Dose Reconstruction Implementation Guideline, Rev 0, OCASIG-002, National Institute for Occupational Safety and Health, Office of Compensation 

Analysis and Support, Cincinnati, Ohio, SRDB Reference ID 22402. 

3. ORAUT (Oak Ridge Associated Universities Team), ORAUT-OTIB-0005, Technical 

Information Bulletin: Internal Dosimetry Organ, External Dosimetry Organ, and IREP 

Model Selection by ICD-9 Code, Rev 02 PC-1, February 10, 2006, SRDB Reference ID 

22595. 

4. Albert, R.E., Thorium – Its Industrial Hygiene Aspects, Academic Press, New York, 

1966. 
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I.C.2 1965 thorium strike 

Intake calculation: 

The thorium strike occurred in Building 81, Room 266 on April 26-28, 1965. The source for this 

information is the health physics logbooks of the time, and personal communication with the 

health physicist involved in the operation. This individual provided a wealth of process details. 

The thorium strike was a wet chemistry process (Kirchner and Freiberg, 1965) which presented 

minimal airborne potential. It was conducted inside a reaction vessel, inside a dry box under 

negative pressure. Most importantly, he stated that due to the high external radiation fields, the 

individuals involved in this operation spent minimal time near the dry box where the chemical 

reaction was being performed. They approached the box only to perform the steps in the 

chemical process, then retreated. 

In response to the Board’s request for an example dose reconstruction based on process specific 

information, NIOSH has retrieved the air sampling data for this room on the relevant days. 

There were 10 fixed-head air samplers employed in the room during the thorium strike. The 

samplers were run for the entirety of both the day and night shifts on these days (16 hours per 

day). Interestingly, the maximum and average air concentrations measured by the sampling 

network in this room throughout the month of April indicate that the concentrations measured 

during the thorium strike are toward the low end of those routinely observed in this room. These 

sampling results, project documentation (Kirchner and Frieberg, 1965, and two HP logbooks 

covering the project) and personal communication with the project health physicist, and the fact 

that this was a wet chemistry process, all indicate that there was no release of material or intake 

potential from this project. 

Under the conditions maintained on the thorium strike project (workers spending minimal time 

near the source, reaction contained in a reaction vessel and conducted inside a dry box under 

negative pressure) NIOSH considers data from fixed-head samplers to be representative of the 

atmosphere to which the workers were exposed. For the example dose reconstruction, NIOSH 

selected the highest of the air monitoring results, and applied it to the entire time period on all 

three days. 

NIOSH considers this to be bounding for the following reasons: (1) the samples are not decay 

corrected to account for natural radon background, (2) it is assumed that the worker was exposed 

for 16 hours on each of three days, (3) the highest air sample from among the 10 samplers was 

selected. The following intakes were calculated: 

April 26, 1965: 

I = (0.44 RCG) x [70 (dpm/m3

)/RCG] x (1.2 m3/hr) x (16 hr) x (0.0167 dpm/Bq) = 10 Bq 

April 27, 1965: 

I = (0.45 RCG) x [70 (dpm/m3

)/RCG] x (1.2 m3/hr) x (16 hr) x (0.0167 dpm/Bq) = 10 Bq 

April 28, 1965: 

I = (0.31 RCG) x [70 (dpm/m3

)/RCG] x (1.2 m3/hr) x (16 hr) x (0.0167 dpm/Bq) = 6.9 Bq 

Total intake = 27 Bq 

Example dose reconstruction: 
This document may contain information covered by the Privacy Act 

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Selection Criteria

• Hypothetical process operator: worked 1959 through 1970 and was exposed to 

thorium from April 26, 1965 through April 28, 1965. 

• Unmonitored for thorium 

Cancer Description:

Bone (ICD-9 170.0) 12/31/2000 

Colon (ICD-9 153.9) 12/31/2000 

Lung (ICD-9 162.0) 12/31/2000 

Prostate (ICD-9 185.0) 12/31/2000 

Employment (Rocky Flats Plant)

Start: 1959 

End: 1970 

Work History

 NOCTS : DOB: 1930, 

Diagnosis Date: 12/31/2000 

Former Smoker 

White-Non-Hispanic 

Dosimetry Data: No thorium dosimetry data. 

Narrative

Internal dose is caused by radioactive materials that are taken into the body. A chronic intake is 

an intake of radioactive material that occurs over an extended period of time (typically weeks or 

longer). An acute intake is an intake of radioactive material that occurs over a short period of 

time (typically minutes to hours). Regardless of the rate at which the intake occurs, the internal 

dose received from radioactive materials having long half-lives occurs over an extended period 

of time and is, therefore, considered chronic. 

The internal dose to the prostate was determined by using the dose calculated for the highest 

non-metabolic organ.3 The organ/tissue associated with this cancer is not included in the ICRP 

modeling of internal doses; so in accordance with NIOSH documentation, the largest dose to an 

exposed organ that is not described by the ICRP metabolic models was assigned as the 

appropriate internal dose (in this case, the adrenals). 

A computer code, the Integrated Modules for Bioassay Analysis (IMBA), was used to estimate 

intakes of radioactive material and the subsequent annual organ doses. The IMBA Expert 

ORAU-Edition was used for this dose reconstruction. The ICRP 66 lung model with default 

aerosol characteristics was assumed, in conjunction with ICRP 68 metabolic models. It should 

be emphasized that intake dates, scenarios, and intake levels were based upon mathematical 

models and do not necessarily prove that such intakes occurred on the given dates. These dates 

and scenarios provide an acceptable explanation of exposure and dose based upon the bioassay 

data provided. This approach is in accordance with the provisions of the Radiation Dose Recon-
This document may contain information covered by the Privacy Act 

11 

struction Rule (42 CFR 82)1

 and guidance in the NIOSH Internal Dose Reconstruction Implementation Guideline.2 

The energy employee assisted in the removal of uranium-232 and its associated daughters 

(specifically thorium-228) from uranium-233 metal, according to records received from the 

Department of Labor and information provided in the interview process. The energy employee’s 

monitoring records were reviewed, and no records of thorium monitoring were found. Based on 

an analysis of the highest gross alpha job-specific air monitoring data, and the claimant favorable 

assumption that all alpha activity was thorium-228, an acute intake of 10 Bq of thorium-228 was 

assigned on April 26, 1965, 10 Bq of thorium-228 was assigned on April 27, 1965, and 6.9 Bq of 

thorium-228 was assigned on April 28, 1965. The most claimant favorable solubility type was 

used. 

Summary

The doses below only capture thorium internal dose. These doses do not include external, 

medical, environmental, or internal doses from other radionuclides. 

Bone: 0.754 rem 

Colon: 0.004 rem 

Lung: 0.559 rem 

Prostate: 0.002 rem 

These dose estimates were entered into IREP with a constant distribution. 

Probability of Causation (POC)

These probably of causations are based on the dose from thorium internal dose only. These 

probably of causations do not include external, medical, environmental, or internal doses from 

other radionuclides. 

Bone: 1.87% 

Colon: 0.01% 

Lung: 1.31% 

Prostate: 0.01% 
This document may contain information covered by the Privacy Act 

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References

1. 42 CFR 82, Methods for Radiation Dose Reconstruction Under the Energy Employees 

Occupational Illness Compensation Program Act of 2000; Final Rule, Federal 

Register/Vol.67, No. 85/Thursday, May 2, 2002, p 22314, SRDB Reference ID 19392. 

2. NIOSH, (2002) Internal Dose Reconstruction Implementation Guideline, Rev 0, OCASIG-002, National Institute for Occupational Safety and Health, Office of Compensation 

Analysis and Support, Cincinnati, Ohio, SRDB Reference ID 22402. 

3. ORAUT (Oak Ridge Associated Universities Team), ORAUT-OTIB-0005, Technical 

Information Bulletin: Internal Dosimetry Organ, External Dosimetry Organ, and IREP 

Model Selection by ICD-9 Code, Rev 02 PC-1, February 10, 2006, SRDB Reference ID 

22595. 

4. Office of Compensation Analysis and Support, Whitepaper on Thorium Exposures at the 

Rocky Flats Plant, 2007. 
This document may contain information covered by the Privacy Act 

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II. ABRWH Request 2 

“Building 881 – There was no Building 881 external monitoring data in the 1950’s. NIOSH has 

provided information about the processes along with the data from the early 1960’s suggest that 

their coworker model may be used to bound gamma and beta doses for Building 881 workers. 

NIOSH needs to demonstrate this by documenting this new approach and completing example 

dose reconstruction(s). In addition, the possibility of plutonium exposures in this building needs 

to be addressed.” 

II.A. Background: 

NIOSH originally provided the analysis below on February 27, 2006, and it was discussed at the 

March 7 and April 19 Working Group meetings. 

The unmonitored periods in the 1950s were for enriched uranium workers in Building 81, as 

discussed with SC&A by ORAUT personnel (Interview with Roger Falk, 5 December, 2006). 

Based on a review of the external dosimetry worksheets posted in the project’s Site Research 

Database, NIOSH has observed that workers in the enriched uranium (EU) operations in 

Building 81 were not monitored for external radiation exposures until the fourth quarter of 1960. 

In the fourth quarter of 1960, 328 workers were monitored, on a monthly exchange frequency, 

for penetrating and non-penetrating (skin) doses. Of the 328 monitored workers, 101 workers 

had a measured zero dose for both penetrating and skin doses for the quarter. For the 227 

workers with at least one non-zero dose measurement, the following results were observed. 

Table II.A.1: Comparison of 1960 measured and coworker doses for Building 81 workers 

Penetrating Dose (mrem) Skin dose (mrem) 

1960 

Median 

95th

Percentile Maximum Median 

95th

Percentile Maximum 

4th Quarter 15 75 135 60 430 710 

Extrapolated1

to One Year 

60 300 540 380 1860 2840 

Co-worker 

Dose2

1293 7121 1645 7728 

1

 Adjusted for partial monitoring in the quarter and multiplied by 4. 

2

 Based on values in Table 7-1 in ORAUT-OTIB-0058, Rev. 01-D. 

In 1961, the EU workers in building 81 were monitored on a quarterly frequency. Two hundred 

twenty six workers had a least one non-zero dose measurement in the year; 51 workers had a 

measured zero dose for both penetrating and skin doses for the year. For the 226 workers with at 

least one non-zero dose measurement, the following results, adjusted for partial monitoring in the 

year, were observed: 
This document may contain information covered by the Privacy Act 

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Table II.A.2: Comparison of 1961 measured and coworker doses for Building 81 workers 

Penetrating Dose (mrem) Skin dose (mrem) 

1961 

Median 

95th

Percentile Maximum Median 

95th

Percentile Maximum 

One Year 45 247 900 147 970 2320 

Co-worker 

Dose 

1527 7850 1923 8201 

 

In both 1960 and 1961, the co-worker doses, which would have been assigned had these workers 

not been monitored, are very generous compared to the monitored doses. Even limiting the 

analysis to the workers with positive measured dose (which overestimates the dose for the entire 

Building 81 population as a whole), the 95th percentile coworker penetrating dose overestimates 

the maximum observed dose by a factor of 13. Similarly, for nonpenetrating dose in 1960, the 

95th percentile coworker nonpenetrating dose overestimates the maximum observed dose by a 

factor of almost 3. The co-worker data for 1961 also overestimate the observed doses in 1961: 

the 95th percentile coworker penetrating dose / maximum penetrating dose = 9, and the 95th

percentile coworker nonpenetrating dose / maximum observed nonpenetrating dose = 4. 

The practice in the 1950s was to monitor workers who had the potential to receive more than 10 

percent of the “tolerance” limit. In the early 1950s, the tolerance limit was 3 rem in a 13-week 

rolling period for whole-body penetrating exposures, which extrapolates to 3,000 mrem/quarter 

and 12,000 mrem/year. When the EU workers were monitored in 1960 and 1961, the penetrating 

doses, even the maximum doses, were in fact less than 10 percent of these values. This 

demonstrates that the process leading to selection of low-potential-dose workers was accurate. 

The recommendation of the IRCP used in the 1950s for the skin dose, as cited in ICRP 

Publication 2 (1959), page xvi, paragraph (13), was 8 rem accumulated during any 13 

consecutive weeks. The ICRP stated, “This is derived from an average of 0.6 rem/week (the 

maximum permissible weekly dose formerly recommended for the skin of the whole body) 

which in 13 weeks amounts to 7.8 rems....”. The ICRP further extrapolates this value to a yearly 

limit of 30 rem to the skin. Ten percent of these limits are 800 mrem/quarter and 3,000 

mrem/year. When the EU workers were monitored in 1960 and 1961, the skin doses, even the 

maximum dose, are less than these values. 

The building’s machining capabilities were expanded with the construction of an additional 

machine shop in 1955 to support the hollow pit design, however this simply added to the original 

machining facilities which began operation in 1953. It is unlikely that the years between 1953 

and 1955 yield such dramatic improvements in machining processes that average dose rates 

would have been significantly affected. Furthermore, the analysis above indicates that the 

coworker doses assigned overestimate even the maximum (i.e. 100th percentile) observed dose. 

It is not intuitively obvious how the expansion of Building 81 machining capabilities would have 

resulted in a lower maximum dose, since the original equipment continued to be used. 
This document may contain information covered by the Privacy Act 

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The weight of the evidence supports that coworker doses applied in earlier years would also be 

bounding of the doses received by unmonitored uranium workers in Building 81 in earlier years. 

This expectation is based on the following factors: 

• The amount of enriched uranium processed in Building 81 steadily increased throughout 

the 1950s, and plateaued in the early 1960s, therefore the source term in the early 1960s 

was higher than the source term in the 1950s; 

• There were no major changes in the Building 81 configuration (e.g. shielding 

improvements, etc.) which would have depressed the doses the workers received in the 

early 1960s relative to the doses received by workers in the 1950s. 

• It may be true that industrial hygiene practices improved with time in this building as has 

been observed at other facilities. This could lead to decreases in exposures with time, 

however as discussed during the May 7th and April 19th Working Group meetings, it does 

not seem plausible that such improvements could have been of sufficient magnitude to 

overcome the degree of overestimation provided by NIOSH’s coworker data, which 

overestimates even the maximum observed dose by a factor of at least three. 

• The workers in Building 81 were not monitored because they were judged to have an 

exposure potential of <10% of the regulatory limit. This judgment was supported once 

the workers were monitored. 

It can also be stated that the workers in 1960 and 1961 still were not required to be monitored 

according to the criteria applied for monitoring in the 1950s. In addition, a professional Health 

Physics staff was in place to make technical decisions based on field measurements and 

production levels at that time. Therefore, NIOSH remains confident that the application of 

OTIB-58 coworker doses throughout the 1950s bounds the external doses that could reasonably 

have been received expected to be received by unmonitored uranium workers in Building 81 

prior to the fourth quarter of 1960, when these workers were unmonitored for external doses. 

Based on this analysis, it can also be stated with confidence that the decision not to monitor these 

workers was in accordance with the relevant regulatory requirements at that time. 

II.B. Plutonium in Building 81: 

The ABRWH also inquired about the possible presence of plutonium in Building 81. The 

Historical American Engineering Archive notes: 

“Beginning sometime after 1960 and continuing until 1977, Building 881 housed the 

chemical recovery operations for site returns and rejected enriched uranium weapon 

components. The first step was to remove surface plutonium contamination by bathing 

the returned parts in nitric acid. The used acid solution was collected, concentrated by 

evaporation, calcined to a dry oxide, and sent to Building 771 for recovery of plutonium. 

The cleaned parts were crushed in a press, processed, and used as feed material for the 

foundry." 

Note that this process (1) began after the initiation of external monitoring for Building 81 

workers in 1960, and (2) involved only surface Pu contamination, which would not present an 

appreciable external exposure hazard compared to the much larger mass of enriched uranium in 

the site returns. 
This document may contain information covered by the Privacy Act 

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The NIOSH/ORAU Team also conducted interviews with five site experts on this question, the 

results of which are presented below. 

[Name Wothheld]: 

[Name Withheld] indicated that they never did anything with Pu in B881 except a wash process 

to remove residual Pu from oralloy (enriched uranium) parts. This process used nitric acid to 

wash the parts and remove the small amounts of Pu that remained. The parts were then scanned 

and sent to the Y-12 plant. The resulting solution was precipitated and dried. If it had less than 

500 ppm Pu, it was sent to Idaho, if not it went to B771. It was primarily a uranium compound 

with small amounts of plutonium. This process was carried out in B881 until 1976 when it was 

moved to B771, Room 174. Again [Name Withheld] pointed out this was primarily a uranium 

process. 

[Name Withheld]: 

[Name Withheld] also recalled the washing operation as the only plutonium in B881. He 

indicated the first wash was accomplished in the plutonium areas and B881 was a second lowlevel wash. The spray washing was accomplished in B881, room 266 in the second glovebox line 

from the north end. [Name Withheld] recalled that separate dosimeters were used on special 

projects and lists were kept of personnel involved. [Name Withheld] recalls that both external 

and internal dosimetry on special projects was reviewed and he recalls no over-exposures with 

any of the projects. 

[Name Withheld]:

[Name Withheld] was the [Withheld] of chemical processing for enriched uranium in B881 from 

the time the building opened until 1962. He was quite certain that there was no plutonium in the 

building during his time there. When asked about the uranium part washing operation, he 

recalled it (after 1962) but did not recall it as a source of plutonium. 

[Name Withheld]: 

[Name Withheld] was the [Withheld] of chemical processing for uranium in B881 from 1962-

1965. He recalls decontamination of the returned units as well. He characterized the plutonium 

contamination as a "nuisance situation." Spot welds on the oralloy shells were sometimes 

plutonium contaminated. 

[Name Withheld]: 

[Name Withheld] said that there were pits stored in a room on the south side of the "general" 

laboratory. They were stored in glove boxes until they were cleaned. There was some potential 

for exposure, but he termed it to be "fairly small." He recalls no instances of any exposure 

incidents as a result of the cleaning. 

Conclusion:

The only known process involving plutonium in Building 881, the cleaning of returned EU 

weapons parts with minimal Pu surface contamination, began after the initiation of external 

monitoring in 1960. Furthermore, the quantities of Pu present were too small to present an 

external exposure hazard. These conclusions are based on interviews with four site experts as 

well as historical documentation, all of which give a consistent account of this process. 
This document may contain information covered by the Privacy Act 

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Therefore, NIOSH concludes that this process does not prevent the reconstruction of external 

doses for Building 81 workers with sufficient accuracy. 

II.C. Example dose reconstruction: 

At the Board’s request, an example dose reconstruction for externally unmonitored uranium

workers in Building 81 is included.
This document may contain information covered by the Privacy Act 

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Selection Criteria

• Hypothetical Process Helper: worked 1959 through 1970 in 881 and was exposed 

to photons and electrons. 

• Unmonitored for all years of employment 

Cancer Description:

Colon (ICD-9 153.9) 12/31/2000 

Kidney (ICD-9 189.0) 12/31/2000 

Lung (ICD-9 162.0) 12/31/2000 

Skin-BCC (ICD-9 173.0) 12/31/2000 

Employment (Rocky Flats Plant)

Start: 1959 

End: 1970 

Work History

 NOCTS : Joe Description: ‘Process Helper’ 

 DOB: 1930, 

Diagnosis Date: 12/31/2000 

Former Smoker 

White-Non-Hispanic 

Dosimetry Data: No external dosimetry data. 

Narrative

External dose is received from radiation originating outside the body and is typically measured 

by dosimetry worn on the body. Radiation dose measured on a film badge or a thermoluminescent dosimeter (TLD) may have been delivered quickly (acute exposure) or slowly over the 

period of time that the employee was exposed (chronic exposure). 

The external dose to the kidney was determined by using the dose calculated for the liver.2

The energy employee worked as a process helper, primarily in Buildings 881, according to 

records received from the Department of Labor and information provided in the interview 

process. Their primary exposure would have been to photons and electrons. However, external 

electron radiation was only considered in this dose reconstruction for the skin, because it would 

not have added dose to the other cancer sites. 

 

For the purpose of estimating probability of causation, all photon and electron doses are assumed 

to be acute.1

 

Radiation Type, Energy, and Exposure Geometry

The exposure geometry was assumed to be consistent with the specific dosimetry parameters 

applicable to the Rocky Flats Plant as described in the Technical Basis Document for the Rocky 

Flats Plant – Occupational External Dosimetry.3 For determination of organ dose and to ensure 

claimant favorability, both 30–250 keV photon doses (based on the reported deep dose 
This document may contain information covered by the Privacy Act 

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measurements) and electrons >15 keV (based on the reported shallow dose measurements) have 

been applied 

In accordance with the NIOSH External Dose Reconstruction Implementation Guideline,1

 dose 

conversion factors (DCFs) appropriate for the era were used to calculate the organ dose from 

exposure to photon radiation. This exposure assumes 100% anterior-posterior geometry. A dose 

conversion factor (DCF) of 1 was used for all doses applied to the skin in this reconstruction per 

guidance in the Technical Information Bulletin: Interpretation of Dosimetry Data for Assignment 

of Shallow Dose.4 For electrons, no additional factors which correct for organ sensitivity and 

clothing attenuation were applied. 

Table II.C.1: External dose parameters 

Building 881 – Exposure (1959 to 1970) 

Photons Energy Range 30-250keV 

Photons Energy Fraction 100% 

Photons Organ DCF (Colon) 1.060 

Photons Organ DCF (Kidney) 1.064 

Photons Organ DCF (Lung) 0.986 

Photons Organ DCF (Skin) 1.000 

No uncertainty correction factors were applied to photon or electron dose to minimize the 

estimated dose. 

Dosimeter/Unmonitored Dose

No individual dosimeter results were available to reconstruct the energy employee’s dose. 

Therefore, external dose was assigned in accordance with the Technical Information Bulletin: 

External Coworker Dosimetry Data for the Rocky Flats Plant and the Technical Information 

Bulletin: Supplementary External Dose Information for the Rocky Flats Plant.5,6 These doses 

were assigned at the 95th percentile of coworker distribution to ensure that the unmonitored 

external dose was not underestimated. 

Summary

The doses below only capture external dose. These doses do not include internal, medical, or 

environmental doses. 

Colon: (63.765 rem) 

Kidney: (64.010 rem) 

Lung: (59.330 rem) 

Skin-BCC: (66.873 rem) 

Probability of Causation (POC)

These probably of causations are based on the dose from external dose only. These probably of 

causations do not include internal, medical, or environmental doses. 

Colon (71.64%) 

Kidney (68.92%) 
This document may contain information covered by the Privacy Act 

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Lung (61.52%) 

Skin-BCC (80.12%) 

References

5. NIOSH, (2006) External Dose Reconstruction Implementation Guideline, Rev 2, OCASIG-001, National Institute for Occupational Safety and Health, Office of Compensation 

Analysis and Support, Cincinnati, Ohio, SRDB Reference ID 29929. 

6. ORAUT (Oak Ridge Associated Universities Team), ORAUT-OTIB-0005, Technical 

Information Bulletin: Internal Dosimetry Organ, External Dosimetry Organ, and IREP 

Model Selection by ICD-9 Code, Rev 02 PC-1, February 10, 2006, SRDB Reference ID 

22595. 

7. ORAUT (Oak Ridge Associated Universities Team), ORAUT-TKBS-0011-6, Technical 

Basis Document for Rocky Flats Plant – Occupational External Dosimetry, Rev 00, 

January 20, 2004, SRDB Reference ID 20175. 

8. ORAUT (Oak Ridge Associated Universities Team), ORAUT-OTIB-0017, Technical 

Information Bulletin: Interpretation of Dosimetry Data for Assignment of Shallow Dose, 

Rev 01, October 11, 2005, SRDB Reference ID 19434. 

9. ORAUT (Oak Ridge Associated Universities Team), ORAUT-OTIB-0058, Technical 

Information Bulletin: External Coworker Dosimetry Data for the Rocky Flats Plant, Rev 

01 PC-1, March 29, 2007, SRDB Reference ID 31084. 

10. ORAUT (Oak Ridge Associated Universities Team), ORAUT-OTIB-0027, Technical 

Information Bulletin: Supplementary External Dose Information for the Rocky Flats 

Plant, Rev 00, May 19, 2005, SRDB Reference ID 19449. 
This document may contain information covered by the Privacy Act 

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III. ABRWH Request 3 

“Neutron Doses 1959 to 1970 – The current NIOSH approach relies on application of a central 

estimate of a building specific neutron photon ratio to estimate doses. The work group has 

remaining questions whether this approach will be bounding for all workers. NIOSH has 

additional data that may be used to estimate a bounding neutron photon ratio which could then be 

applied to bound worker doses during this time period. NIOSH needs to demonstrate this by 

documenting this new approach and completing example dose reconstruction(s).” 

III.A. Background: 

The Neutron Dose Reconstruction Project (NDRP) was undertaken in the 1990s to address 

known limitations in Rocky Flats neutron dosimetry from the early years of operation. This 

project re-evaluated neutron doses for workers monitored for beta/gamma exposures in 

plutonium buildings from 1952-69. On May 4, the Board recommended the addition of a class to 

the SEC consisting of workers who were or should have been monitored for neutrons at Rocky 

Flats from 1952-1958. Therefore this discussion will focus on the 1959-1970 time period. 

III.B. NDRP methodology 

The NDRP recalculated neutron doses according to the following equation: 

Dneutron = Doriginal + Dre-read + Dnotional 

Dneutron = recalculated total neutron dose 

Doriginal = original doses from films which were unable to be re-read 

Dre-read = measured neutron doses from re-evaluated neutron films and track plates 

Dnotional = unmonitored neutron dose estimated using neutron/gamma ratios 

Original doses from films which could not be re-evaluated 

There were a small number of films which could not be re-evaluated by the NDRP primarily 

because the films could not be located. In this situation, the NDRP let the original result stand. 

NIOSH has proposed to adjust these films by a film reading bias factor which reflects an 

adjustment for low-energy photons not detectable by NTA film, and for more accurate reading 

during the NDRP. SC&A has questioned the values used by NIOSH for this bias factor (1.99 for 

Building 71, and 1.13 for all other buildings). The original source of these correction factors was 

a report by Dr. James Ruttenber. In response to SC&A’s concerns on this issue, NIOSH has reevaluated these correction factors using films which were re-read by the NDRP. The two tables 

below show the results of NDRP re-reads for two situations (1) original film reading equal to 

zero, and (2) original film reading greater than zero. For the cases where the original reading 

was zero, NIOSH proposes to assign the 95th percentile daily neutron dose rate from the re-read 

films. For cases where the original reading was greater than zero, NIOSH proposes to assign the 

95th percentile of the ratio of re-read to original readings. This approach is consistent with the 

co-worker methods employed by NIOSH in other situations, and is claimant-favorable. This will 

ensure that doses from original, unadjusted films will not be underestimated. 

Table III.B.1: Re-read doses for original readings equal to zero 

Years N r

2 50th percentile (mrem/cycle) 95th percentile (mrem/cycle) 

1959-1969 36037 0.992 73 183 
This document may contain information covered by the Privacy Act 

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Table III.B.2: Re-read doses for original readings greater than zero 

Years N r

2

50th percentile 

(re-read/original) 

95th percentile 

(re-read/original) 

1959-1969 34327 0.999 1.640 6.950 

Re-read doses 

From 1957-70, workers were monitored with NTA film. A total of 89,976 films were located for 

the NDRP. Of these, 87,286 were matched to workers. All available neutron doses measured 

with plates or films were re-evaluated by re-reading the films, many were re-read several times. 

Individual-specific calibration factors were calculated for each person re-reading films by 

comparison to calibration films exposed to known doses. There were two sets of calibration 

films (1) one set was exposed to a bare, unmoderated neutron source of PuF4 (similar to the 

source term at Rocky Flats), and (2) the second set was exposed to a PuF4 source moderated by 7 

cm polyethylene. These two configurations yield spectra which would bookend the neutron 

fields experienced at Rocky Flats during the entire span of the NDRP (1952-69). No significant 

differences in calibration factors were observed between the two configurations. Prior to reading 

any films on any day, readers were required to read and pass an initial qualification test using 

calibration films. A separate routine quality control program was implemented to reread at least 

10% of films that were read the previous day by each reader. NIOSH proposes to use the values 

for re-read films as determined by the NDRP in dose-reconstructions. 

Notional doses 

Notional neutron doses are neutron doses that are assigned to a worker who was potentially 

exposed to neutrons in a Pu-related building for a period of time but was not originally credited 

with a neutron dose for that time. Periods with no neutron monitoring data could result from (1) 

the worker was not monitored for neutrons, (2) the worker was monitored for neutrons, but doses 

could not be evaluated from the film, (3) the worker was not likely to have been exposed to 

neutrons during that period. Notional doses were assigned to the first two instances. If a worker 

was judged to have exposure potentials greater than 10% of the regulatory limit, they were 

monitored for beta/gamma exposure. 

The NDRP calculated notional doses using neutron/photon ratios. In light of the Board’s 

continuing concern with the application of n/p ratios, and the Board’s explicit request for a new 

approach, NIOSH has revised the dose reconstruction approach for workers in plutonium 

buildings who were unmonitored for neutrons and/or gamma exposure. Rather than relying on a 

n/p ratio, which SC&A has questioned, the revised approach relies on coworker distributions of 

measured daily neutron and gamma dose-rates. 

The badging policy in place during 1959-70 was that if a worker was judged to have an exposure 

potential greater than 10% of the regulatory limits of the time, they were required to be 

monitored. Workers judged to be at lower exposure potential were not necessarily required to be 

monitored (but sometimes were). The effect of monitoring only the workers judged to be at 

higher exposure potential would be to bias the coworker distributions higher than they would be 

if the entire population had been monitored. The application of these coworker distributions 

would therefore be claimant-favorable. 
This document may contain information covered by the Privacy Act 

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SC&A and the Working Group have noted that for the years 1960-64, some of the highest doses 

calculated by the NDRP were based entirely on notional dose. The conclusion has been drawn 

that this may indicate that the workers with the highest exposure potentials were not monitored. 

NIOSH does not concur with this conclusion. Rather, the notional doses calculated by the 

NDRP in these instances represent the results of applying a worker-favorable n/p ratio to 

individuals with high gamma doses. If the application of a n/p ratio is considered questionable, 

as the Board and SC&A have indicated, then dose estimates that rely on such a technique cannot 

form the basis of a conclusion that the highest exposed workers were not monitored. SC&A has 

noted that there is very little correlation between high gamma exposures and high neutron 

exposures (e.g. Figure 8 from SCA-SEC-TASK5-0052 Supplemental Report and associated 

text). NIOSH has also noted that individual badge readings with high gamma exposures do not 

always have high associated neutron badge readings. For example, the table below shows the 

highest 10 measured gamma doses in Building 71 in 1959, the highest 10 measured gamma 

doses in Building 91 in 1959, the associated measured neutron doses, and the observed n/p ratios 

compared to the ratios used to calculate notional dose by the NDRP. In every case, the observed 

ratio is lower than the ratio applied by the NDRP. This demonstrates that the high notional doses 

calculated by the NDRP are a result of an overestimation of the n/p ratio, rather than the workers 

with the highest exposure potential being unmonitored. 

Table III.B.3: Highest 1959 gamma doses, associated neutron doses, and n/p ratios

Building Year Neutron dose Gamma dose Measured n/p NDRP n/p 

71 1959 72.00 1,613.00 0.04 1.40 

71 1959 145.00 1,290.00 0.11 1.40 

71 1959 123.00 1,180.00 0.10 1.40 

71 1959 99.00 1,118.00 0.09 1.40 

71 1959 133.00 1,063.00 0.13 1.40 

71 1959 105.00 980.00 0.11 1.40 

71 1959 144.00 907.00 0.16 1.40 

71 1959 148.00 870.00 0.17 1.40 

71 1959 154.00 735.00 0.21 1.40 

71 1959 79.00 731.00 0.11 1.40 

91 1959 52.00 85.00 0.61 3.60 

91 1959 42.00 85.00 0.49 3.60 

91 1959 54.00 80.00 0.68 3.60 

91 1959 40.00 75.00 0.53 3.60 

91 1959 27.00 75.00 0.36 3.60 

91 1959 84.00 75.00 1.12 3.60 

91 1959 84.00 75.00 1.12 3.60 

91 1959 81.00 75.00 1.08 3.60 

91 1959 30.00 75.00 0.40 3.60 

91 1959 87.00 60.00 1.45 3.60 
This document may contain information covered by the Privacy Act 

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Due to the Board’s continuing concerns and explicit request for a new approach, NIOSH has 

developed coworker distributions of daily neutron and gamma dose rates. The gamma dose rate 

distributions are based on the exposures observed from β-γ film results, and the neutron dose rate 

distributions are based on the exposures observed from the re-read NTA film results. These 

coworker distributions do not rely on n/p ratios, which should address the Board’s concerns on 

this issue. Values for 50th and 95th percentiles have been generated, which NIOSH proposes to 

apply in accordance with the various guidance documents governing assignment of coworker 

data. 

A significant overestimating factor in this approach is that NIOSH has chosen to rely on the 

distribution of individual cycle data, rather than annual doses for individual workers. The data 

show that the highest badge readings are dispersed across numerous individual workers 

throughout a given year, as opposed to a few workers consistently showing the highest badge 

results. NIOSH’s reliance on individual cycle data has the effect of applying the 50th or 95th

percentile badge readings from the entire year for the entire monitored population during the 

entire time workers were unmonitored. If NIOSH had instead relied on annual doses from 

individual workers for generation of the coworker distributions, the assigned doses would be 

lower (data not shown) because high badge readings tend to be averaged out over the course of a 

year. This should mitigate the need for a job-type analysis, as proposed by SC&A. 
Table III.B.4: Coworker gamma and neutron dose rates 

Building Year

N

r

2

50% 

(mrem/yr) 

95% 

(mrem/yr)

N

r

2

50% 

(mrem/yr) 

95

(mrem

Gamma Neutron

71 1959 4738 0.90 1356 6779 2331 0.99 2659 8071 1960 6507 0.87 574 3546 3480 1.00 4380 9271 1961 3110 0.86 1330 4849 1669 1.00 4198 9371 1962 4715 0.81 1069 4105 2916 0.94 4041 11271 1963 11496 0.81 574 3337 10116 0.99 2008 5871 1964 14151 0.89 261 2320 12414 0.99 1734 4571 1965 6186 0.92 523 4000 5477 0.99 1030 2671 1966 6034 0.87 1192 6037 5188 0.98 937 2471 1967 6853 0.90 907 5614 6104 0.96 808 3971 1968 8212 0.88 678 3468 6493 0.97 991 6971 1969 7712 0.88 590 4278 5580 1.00 1590 4871 1970 1811 0.90 240b 4278a 2734 0.99 1675 5176 1959 6651 0.83 1616 6153 442 0.99 1637 4376 1960 5474 0.75 1981 5840 542 0.97 1753 6776 1961 4529 0.71 2451 5319

0 NA 1753a 672

76 1962 5889 0.72 1721 4484 0 NA 1753a 67276 1963 4793 0.75 1616 4563 0 NA 1753a 67276 1964 1565 0.79 1741 5079 10 0.87 704 2076 1965 2945 0.83 913 5342 1434 0.96 777 1576 1966 3206 0.77 1898 7324 1978 0.98 852 1776 1967 4808 0.89 706 6486 2048 0.99 971 2776 1968 4668 0.88 283 2759 1657 0.99 954 3476 1969 4942 0.83 65 1570 714 0.98 1217 2276 1970 2385 0.84 120b 1102 131 0.72 1703 2777 1959 3073 0.89 261 2868 64 0.98 1755 3577 1960 2861 0.87 130 3598 0 NA 1947a 37977 1961 3248 0.79 1121 5214 0 NA 1947a 379

This document may contain information covered by the Privacy Act 

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Building Year N r2

50% 

(mrem/yr) 

95% 

(mrem/yr) N r2

50% 

(mrem/yr) 

95

(mrem

 Gamma Neutron 

77 1962 4253 0.78 479 2764 0 NA 1947a 37977 1963 3576 0.78 815 4343 111 0.98 1947 3777 1964 1112 0.77 965 4958 266 0.95 2268 3877 1965 1488 0.82 754 3682 995 0.99 1013 2177 1966 1674 0.78 1278 5519 1438 0.99 999 2077 1967 1334 0.86 1454 5299 1238 0.99 1048 1977 1968 1246 0.85 586 2772 863 0.97 969 1877 1969 2439 0.86 88 1991 448 0.98 717 1377 1970 800 0.81 120b 2392 57 0.70 2060 2891 1959 1497 0.88 240b 521 574 1.00 808 1791 1960 1954 0.94 240b 261 313 0.88 617 1591 1961 2011 0.98 240b 521 102 0.91 1501 2391 1962 1982 0.99 240b 519 71 0.88 2138 3191 1963 2218 0.99 240b 587 91 0.84 1104 4891 1964 1862 0.95 240b 417 72 0.98 1186 2191 1965 1375 0.97 240b 706 97 0.99 1180 2291 1966 755 0.91 118 1662 49 0.97 1189 1691 1967 766 0.94 240b 1276 0 NA 1189a 16691 1968 895 0.94 240b 665 20 0.95 669 1091 1969 739 0.96 240b 483 38 0.98 1108 2291 1970 374 0.91 209 1322 0 NA 1108a 223

All 1959 16304 0.86 782 5788 3456 1.00 2112 70All 1960 17456 0.82 433 4745 4379 0.93 3963 88All 1961 14063 0.77 1051 5058 1821 0.97 3989 92All 1962 17532 0.75 730 3964 3032 0.94 3937 111All 1963 23448 0.80 548 3807 10646 0.99 1981 57All 1964 25823 0.86 339 3285 13033 0.99 1721 44All 1965 14215 0.87 393 3852 8617 0.99 965 24All 1966 13853 0.83 900 5958 9166 0.99 907 21

This document may contain information covered by the Privacy Act 

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This document may contain information covered by the Privacy Act 

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Building Year N r2

50% 

(mrem/yr) 

95% 

(mrem/yr) N r2

50% 

(mrem/yr) 

95

(mrem

 Gamma Neutron 

All 1967 18053 0.89 482 5137 10154 0.97 873 31All 1968 19632 0.88 281 2869 10173 0.96 965 50All 1969 19356 0.87 135 2997 7803 1.00 1348 43All 1970 6898 0.81 120b 1200 4697 0.99 1314 42

NA = not available a = data in given year and building insufficient to generate estimate. The higher of the preceding or following ye

b = 50th percentile was zero, therefore value calculated as missed dose from badge exchange frequency identified 
This document may contain information covered by the Privacy Act 

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III.C. Example dose reconstruction 

Selection Criteria

• Hypothetical Process Helper: worked 1959 through 1970 in 771 and was exposed 

to photons and neutrons. 

• Monitored for photons and partially monitored for neutrons. 

Cancer Description:

Colon (ICD-9 153.9) 12/31/2000 

Kidney (ICD-9 189.0) 12/31/2000 

Lung (ICD-9 162.0) 12/31/2000 

Skin-BCC (ICD-9 173.0) 12/31/2000 

Employment (Rocky Flats Plant)

Start: 1959 

End: 1970 

Work History

 NOCTS : Joe Description: ‘Process Helper’ 

 DOB: 1930, 

Diagnosis Date: 12/31/2000 

Former Smoker 

White-Non-Hispanic 

Dosimetry Data: Photon dosimetry from June 1, 1959 through December 31, 1970. 

Neutron Dosimetry from July 2, 1962 though December 31, 1970 

with some gaps. 

Narrative

External dose is received from radiation originating outside the body and is typically measured 

by dosimetry worn on the body. Radiation dose measured on a film badge or a thermoluminescent dosimeter (TLD) may have been delivered quickly (acute exposure) or slowly over the 

period of time that the employee was exposed (chronic exposure). 

The external dose to the kidney was determined by using the dose calculated for the liver.2

The energy employee worked as a process helper, primarily in Buildings 771, according to 

records received from the Department of Labor and information provided in the interview 

process. Their primary exposure would have been to photons and neutrons. 

For the purpose of estimating probability of causation, all photon doses are assumed to be acute 

and all neutron doses are assumed to be chronic.1

 

Radiation Type, Energy, and Exposure Geometry

The records supplied by the Department of Energy and the interview process indicate the energy 

employee worked at the 771 facilities. The energy employee’s exposure geometry was assumed 

to be consistent with the specific dosimetry parameters applicable to the Rocky Flats Plant as 
This document may contain information covered by the Privacy Act 

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described in the Technical Basis Document for the Rocky Flats Plant – Occupational External 

Dosimetry.3

 

In accordance with the NIOSH External Dose Reconstruction Implementation Guideline,1

 dose 

conversion factors (DCFs) appropriate for the era were used to calculate the organ dose from 

exposure to photon and neutron radiation. This exposure assumes 100% anterior-posterior 

geometry. Plutonium specific DCFs were applied for <30 keV photons. 

In the NIOSH External Dose Reconstruction Implementation Guideline,1 organ dose conversion 

factors are tabulated by averaging the energy specific values from ICRP 74 (1986)9

 over the 

IREP photon energy range. The lowest photon energy interval in the Interactive 

RadioEpidemiological Program (IREP) is categorized as less than 30 keV. Plutonium emits 

several X-rays in this energy range; however, a simple average as used in the Implementation 

Guideline may not result in the most accurate dose conversion factor. For plutonium work, the 

average X-ray energy is approximately 17 keV. As a result, using 20 keV as a claimantfavorable single-point estimate is most appropriate. Since the low energy photon dose to 

glovebox workers, laboratory technicians, maintenance workers, metallurgical operators, and D 

& D workers is predominantly in the anterior-posterior (AP) geometry,3

 single-point estimate 

values using anterior-posterior (AP) geometry were calculated for 16 organs listed in ICRP 74.9 

Some workers (site support personnel, chemical operators when not working with gloveboxes, 

support personnel, and radiation technicians) were estimated to have received varying amounts 

of non-AP dose.3 Since there is significant uncertainty in the individual exposure geometry and 

AP geometry is generally claimant favorable or neutral compared to other geometries for most 

cancers, an AP geometry is applied for all <30 keV photon exposures. 

A dose conversion factor (DCF) of 1 was used for all doses applied to the skin in this 

reconstruction per guidance in the Technical Information Bulletin: Interpretation of Dosimetry 

Data for Assignment of Shallow Dose.4

 For neutrons, additional correction factors (which 

incorporate the energy range fractions) and ICRP 60 correction factors were applied in 

accordance with the Technical Basis Document for the Rocky Flats Plant – Occupational 

External Dosimetry.3

 

Table III.C.1: External dose parameters 

Building 771 – Exposure (1959 to 1970) 

Photon Neutron 

Energy Range <30 keV 30-

250keV <10 keV 10–100 

keV 

100 keV – 

2 MeV 2–20 MeV 

Energy Fraction 100% 100% — — — — 

ICRP 60 CF — — 0.085 0.034 1.361 0.327 

Organ DCF (Colon) 0.0150 1.0597 2.0117 0.9612 0.5037 0.9666 

Organ DCF (Kidney) 0.0410 1.0638 2.0384 0.9972 0.6641 1.0466 

Organ DCF (Lung) 0.0300 0.9860 1.5234 0.7509 0.5791 1.0042 

Organ DCF (Skin) 1.000 1.000 1.000 1.000 1.000 1.000 

No uncertainty correction factors were applied to photon or neutron dose to minimize the 

estimated dose. 
This document may contain information covered by the Privacy Act 

30 

Dosimeter Dose

Individual dosimeter results were used to reconstruct the energy employee’s dose. Corrections to 

the reported doses were applied as described above. The photon doses based on measured and 

coworker data were estimated using algorithms from the Technical Information Bulletin: 

Supplementary External Dose Information for the Rocky Flats Plant.6

 

The recent re-evaluation of individual neutron dosimetry data provided detailed neutron 

dosimetry data and was used as reported,7

 per guidance in the Technical Information Bulletin: 

Use of Rocky Flats Neutron Dose Reconstruction Project Data in Dose Reconstructions.8

 

For non-affected original neutron dosimetry data that was not re-evaluated by the Neutron Dose 

Reconstruction Project (NDR Project),7 doses were adjusted based on an analysis of all reevaluated dosimetry at the 95th percentile. An adjustment factor of 6.95 for all non-zero data and 

183 mrem for each zero were applied. 

Coworker Dose Assignment 

During the periods that the energy employee was on site and not monitored (i.e., there was no 

reported gamma dose or NDRP Notional dose was reported), external dose was assigned in 

based on external daily dose rates derived for coworker dosimetry data.5

 These doses were 

assigned at the 95th percentile of coworker distribution to ensure that the unmonitored external 

dose to the organ was not underestimated. This results in a coworker dose assignment of 0.675 

rem of shallow dose, 2.840 rem of deep dose, and 36.302 rem of neutron dose. 

Missed Dose 

A potential missed dose was assigned to each actual or potential dosimeter cycle where a zero 

was reported to provide a claimant-favorable estimate of the potential external doses received. A 

missed dose represents the dose that could have been received but may not have been recorded 

due to the dosimeter detection limits or site reporting practices. 

The total number of gamma dosimeter cycles where a zero was assigned was 7. This number 

was chosen to ensure that a best estimate of instances of a zero badge reading was accounted for 

in this dose reconstruction. Based on limit of detection information provided in the Technical 

Basis Document for the Rocky Flats Plant,3

 this results in a potential missed dose of 0.148 rem to 

the colon, 0.149 rem to the kidney, 0.138 rem to the lung, and 0.140 rem to the skin from 30–250 

keV photon radiation. The calculated missed dose was applied as the geometric mean with 

associated geometric standard deviation for the purpose of calculating probability of causation. 

The total number of re-evaluated neutron zero dosimeter reported from the NDR Project was 

assigned per guidance in the Technical Basis Document for the Rocky Flats Plant.3

 The total 

number of re-evaluated neutron zero dosimeter cycles assigned was 20. This number was chosen 

to ensure that a best estimate of instances of a zero badge reading was accounted for in this dose 

reconstruction. Based on limit of detection information provided in the Technical Basis 

Document for the Rocky Flats Plant,3 this results in a potential missed dose for of 2.786 rem to 

the colon, 3.359 rem to the kidney, 2.939 rem to the lung, and 4.176 rem to the skin. The 
This document may contain information covered by the Privacy Act 

31 

calculated missed dose was applied as the geometric mean with associated geometric standard 

deviation for the purpose of calculating probability of causation. 

Summary

The doses below only capture external dose. These doses do not include internal, medical, or 

environmental doses. 

Colon: (79.971 rem) 

Kidney: (93.412 rem) 

Lung: (82.509 rem) 

Skin-BCC: (119.441 rem) 

Probability of Causation (POC)

These probably of causations are based on the dose from external dose only. These probably of 

causations do not include internal, medical, or environmental doses. 

Colon (56.05%) 

Kidney (56.30%) 

Lung (49.02%) 

Skin-BCC (77.03%) 

References

1. NIOSH, (2006) External Dose Reconstruction Implementation Guideline, Rev 2, OCASIG-001, National Institute for Occupational Safety and Health, Office of Compensation 

Analysis and Support, Cincinnati, Ohio, SRDB Reference ID 29929. 

2. ORAUT (Oak Ridge Associated Universities Team), ORAUT-OTIB-0005, Technical 

Information Bulletin: Internal Dosimetry Organ, External Dosimetry Organ, and IREP 

Model Selection by ICD-9 Code, Rev 02 PC-1, February 10, 2006, SRDB Reference ID 

22595. 

3. ORAUT (Oak Ridge Associated Universities Team), ORAUT-TKBS-0011-6, Technical 

Basis Document for Rocky Flats Plant – Occupational External Dosimetry, Rev 00, 

January 20, 2004, SRDB Reference ID 20175. 

4. ORAUT (Oak Ridge Associated Universities Team), ORAUT-OTIB-0017, Technical 

Information Bulletin: Interpretation of Dosimetry Data for Assignment of Shallow Dose, 

Rev 01, October 11, 2005, SRDB Reference ID 19434. 

5. ORAUT (Oak Ridge Associated Universities Team), ORAUT-OTIB-0058, Technical 

Information Bulletin: External Coworker Dosimetry Data for the Rocky Flats Plant, Rev 

01 PC-1, March 29, 2007, SRDB Reference ID 31084. 
This document may contain information covered by the Privacy Act 

32 

6. ORAUT (Oak Ridge Associated Universities Team), ORAUT-OTIB-0027, Technical 

Information Bulletin: Supplementary External Dose Information for the Rocky Flats 

Plant, Rev 00, May 19, 2005, SRDB Reference ID 19449. 

7. ORAUT (Oak Ridge Associated Universities Team), Neutron Dose Reconstruction 

Protocol, ORISE 05-0199, SRDB Reference ID 17126. 

8. ORAUT (Oak Ridge Associated Universities Team), ORAUT-OTIB-0050, Technical 

Information Bulletin: Use of Rocky Flats Neutron Dose Reconstruction Project Data in 

Dose Reconstructions, Rev 00, December 13, 2005, SRDB Reference ID 20223. 

9. ICRP (International Commission on Radiological Protection), 1996, Conversion 

Coefficients for Use in Radiological Protection Against External Radiation, ICRP 

Publication 74, Annals of the ICRP Vol 26(3/4), Pergamon Press, Oxford, England, 

SRDB Reference ID 7979. 
This document may contain information covered by the Privacy Act 

33 

Attachment 1: Energy Employee’s Monitoring Data 

Measured Dose 

(mrem) 

Year Building SKIN PEN Gamma 

Neutron 

(non-affected) 

Neutron 

(re-read) 

1959 71 931 363 363 0 0 

1960 71 1151 975 975 0 0 

1961 71 672 608 608 0 0 

1962 71 1345 1104 1104 0 168 

1963 71 2129 1840 1269 0 2572 

1964 71 1258 891 891 0 1633 

1965 71 1359 1145 1065 80 743 

1966 71 3055 2714 2263 66 606 

1967 71 2314 1577 1120 0 682 

1968 71 1846 1097 795 50 863 

1969 71 2089 1289 885 110 567 

1970 71 1498 1086 690 176 326 

Unmonitored Period 

(days) Zeros 

Year Building Gamma Neutron Gamma 

Neutron 

(non-affected) 

Neutron 

(re-read) 

1959 71 151 368 0 0 0 

1960 71 0 364 0 0 0 

1961 71 0 363 0 0 0 

1962 71 0 270 1 0 0 

1963 71 0 87 1 0 2 

1964 71 0 0 0 9 9 

1965 71 0 0 2 1 2 

1966 71 0 0 2 1 2 

1967 71 0 0 0 0 3 

1968 71 0 0 1 1 0 

1969 71 0 0 0 2 0 

1970 71 0 0 0 2 2