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This is a working document prepared by NIOSH’s Division of Compensation Analysis and Support (DCAS) or its contractor for use in discussions 

with the ABRWH or its Working Groups or Subcommittees. Draft, preliminary, interim, and White Paper documents are not final NIOSH or 

ABRWH (or their technical support and review contractors) positions unless specifically marked as such. This document represents preliminary 

positions taken on technical issues prepared by NIOSH or its contractor. NOTICE: This report has been reviewed to identify and redact any 

information that is protected by the Privacy Act 5 USC §552a and has been cleared for distribution. 

Reassessment of Internal Radiation Dose 

from Sources at the Rocky Flats Plant 

Critical Mass Laboratory

White Paper 

Rev. 0 

National Institute for Occupational 

Safety and Health 

November 28, 2016

James S. Bogard, Daniel Stempfley, Bob Burns

Oak Ridge Associated Universities Team

Reviewed by LaVon Rutherford 

Division of Compensation Analysis and Support 
White Paper: Reassessment of Internal Dose from Sources at the RFP CML

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with the ABRWH or its Working Groups or Subcommittees. Draft, preliminary, interim, and White Paper documents are not final NIOSH or 

ABRWH (or their technical support and review contractors) positions unless specifically marked as such. This document represents preliminary 

positions taken on technical issues prepared by NIOSH or its contractor. NOTICE: This report has been reviewed to identify and redact any 

information that is protected by the Privacy Act 5 USC §552a and has been cleared for distribution. 

PURPOSE

This report reevaluates prior assumptions used to assess upper bounds on personnel doses from 

mixed fission and activation products (MFAP) at the Rocky Flats Plant (RFP) Critical Mass 

Laboratory (CML). A publicly released document citing a maximum reactor operating power of 

10 milliwatts (mW) and a typical experiment duration of one hour were previously used to 

estimate the MFAP inventory built up over time in high-enriched uranium (HEU) solution fuel 

used in CML experiments. External radiation dose was monitored by radiation dosimeters 

assigned to each individual and has not been considered herein. 

In the earlier assessment, internal doses from the estimated MFAP inventory were bounded by 

assuming removable surface contamination from dried solution spills to be at the limit for a 

posted Contamination Area. The average air concentration of respirable particles was then 

estimated by applying a resuspension factor. Recently captured documents contain air 

monitoring results, surface contamination measurements, more-accurate power estimates based 

on carefully measured gamma photon emissions from the irradiated fuel, records of experiment 

duration, and neutron flux profiles. This information provides data against which the prior 

assumptions can be re-evaluated. 

MFAP PERSONNEL EXPOSURE CONCERNS

Fission products in irradiated fuels, and activation products in both the fuel and containment 

materials, are sources of external radiation dose to personnel using or working around the fuels, 

and they present an internal dose potential for personnel who might ingest or inhale them. 

Personnel dosimeters assigned to RFP radiation workers document the external exposures. 

Internal exposures might result: (1) during operations, from resuspension of contamination on 

surfaces, or (2) during facility demolition from airborne dust. CML staff was provided routine 

bioassay (urinalysis and whole-body counts) to detect intakes of plutonium, uranium, or 

americium, but MFAP were not routinely monitored. 

Surface contamination in CML experimental areas was extensive, predominantly due to 

high-enriched uranyl nitrate (HEU) solution spills1

1

 Eleven spills involved solutions or dried salts from solution experiments, compared with two contamination 

events involving other solid fuels (one spill each of low-enriched uranium and plutonium powders).

 over the course of the facility’s history 

(Rothe, 2005, PDF pp. 447, 449, 452, 454-458, 464, 467-471, 479, 486, 498, 500-502). Thus, 

MFAP important from either an acute or chronic personnel internal exposure perspective are 

those generated by CML criticality experiments involving HEU solution. MFAP atom ratios on 

CML contaminated surfaces and resuspended in air from these surfaces are presumed to be the 

same as those in solution. 

 
White Paper: Reassessment of Internal Dose from Sources at the RFP CML

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with the ABRWH or its Working Groups or Subcommittees. Draft, preliminary, interim, and White Paper documents are not final NIOSH or 

ABRWH (or their technical support and review contractors) positions unless specifically marked as such. This document represents preliminary 

positions taken on technical issues prepared by NIOSH or its contractor. NOTICE: This report has been reviewed to identify and redact any 

information that is protected by the Privacy Act 5 USC §552a and has been cleared for distribution. 

HISTORY OF UNH CRITICALITY EXPERIMENTS

Criticality experiments involving uranyl nitrate solution were conducted from the beginning of 

CML operations until the last experiment in 1987. These experiments are documented in some 

detail in a published history of the facility (Rothe, 2005, PDF pp. 376-394). Figure 1 shows the 

distribution of these experiments over time. Heights of the bars represent the average number of 

experiments per day during the Program period discussed in the published history; numbers 

above the bars are the total numbers of experiments. The dotted line shows the overall average 

number of HEU solution critical experiments over the history of CML operations. 

Date

1/1/19681/1/19701/1/19721/1/19741/1/19761/1/19781/1/19801/1/19821/1/19841/1/1986

Experiment Rate (d-1) 0.0

0.2

0.4

0.6

0.8

1.0

38

110

54

206 52

20

186

61

32

19

Average HEU Solution

Experimental Rate

(0.104 d-1)

Source: Rothe, 2005, PDF pp. 376-394

Figure 1: Time Distribution of CML Criticality Experiments Involving HEU Solution
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This is a working document prepared by NIOSH’s Division of Compensation Analysis and Support (DCAS) or its contractor for use in discussions 

with the ABRWH or its Working Groups or Subcommittees. Draft, preliminary, interim, and White Paper documents are not final NIOSH or 

ABRWH (or their technical support and review contractors) positions unless specifically marked as such. This document represents preliminary 

positions taken on technical issues prepared by NIOSH or its contractor. NOTICE: This report has been reviewed to identify and redact any 

information that is protected by the Privacy Act 5 USC §552a and has been cleared for distribution. 

The 778 experiments performed at the CML with HEU solution or systems of HEU solution plus 

HEU metal took place in 10 campaigns between May 1967 and October 1987. Table 1 

summarizes the 10 campaigns, where the “Downtime (d)” for each campaign equals the number 

of days between the last experiment of the specified campaign and the first experiment of the 

subsequent campaign. 

Table 1: CML Experimental Campaigns Involving HEU Solution

Campaign No. of Experiments Downtime (d)

1 38 32

2 110 23

3 54 398

4 206 111

5 52 387

6 20 1,377

7 186 2,069

8 61 274

9 32 182

10 19 N/A

REEVALUATED FISSION AND ACTIVATION PRODUCT LEVEL ESTIMATES

NIOSH calculated fission and activation product build-up in uranium solution experiments at the 

CML using ORIGEN-S, a computer code system for calculating time-dependent concentrations 

of radionuclides that are simultaneously generated or depleted by processes such as fission, 

neutron absorption/transmutation, and radioactive decay. Initial assumptions about power levels 

and experiment durations were taken from a U.S. Department of Energy (DOE) public document 

that stated that the experiments conducted in Building 886 (housing the CML) generally 

involved power levels of no more than 10 mW for no more than one hour. It also stated that 

approximately half of the experiments conducted in Building 886 actually achieved criticality, 

and only rarely were the radiation levels such that it was not possible to directly touch the fissile 

material and testing apparatus immediately after the experiments (Building 886, 2011, PDF p. 3). 

Some of these assumptions were challenged during an oral interview with a former CML 

Associate Research Scientist, who asserted: (1) that the neutron flux for a CML experiment 

could not be bounded; and (2) that the best one could say is that the power level was probably 

less than 50 kW [the DOE Category B reactor limit] (Personal Communication, 2015a, PDF 

p. 9).

Documents subsequently captured by NIOSH contain neutron flux and thermal power estimates 

by CML staff using analytical measurements made during or after criticality experiments. These 

calculations were performed for at least three different enriched uranium fuel configurations: (1) 

bare HEU metal (Oy) spheres; (2) Oy spheres immersed in uranyl nitrate solution (UNH); and 

(3) UNH only.
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This is a working document prepared by NIOSH’s Division of Compensation Analysis and Support (DCAS) or its contractor for use in discussions 

with the ABRWH or its Working Groups or Subcommittees. Draft, preliminary, interim, and White Paper documents are not final NIOSH or 

ABRWH (or their technical support and review contractors) positions unless specifically marked as such. This document represents preliminary 

positions taken on technical issues prepared by NIOSH or its contractor. NOTICE: This report has been reviewed to identify and redact any 

information that is protected by the Privacy Act 5 USC §552a and has been cleared for distribution. 

Thermal Power and Fission Rate Estimates in Bare Metal Experiments

Measurements of surface gamma photon fluence rates from a 7-cm-radius Oy sphere used in a 

1967 one-hour criticality experiment led to an estimate of 1.9508×10-9 megawatts (MW) 

(1.9508 mW)2

2 The conversion from MW to mW in the reference document incorporated a factor of ten error, incorrectly 

concluding that the power level was 19.508 mW; the error is corrected here.

thermal power, corresponding with 2.18×1011 total fissions3

3 Using 1 W = 3.1×1010 fissions per second.

 (Log, 1967, PDF 

p. 6). 

Thermal Power and Fission Rate Estimates in Metal-Plus-Solution Experiments

CML staff made two estimates of total fissions in experiments involving Oy immersed in UNH. 

One 1976 estimate used the residual gamma photon flux from the Oy + UNH to estimate an 

upper bound of 1.8×1018 total fissions over the 12 years of CML operation from 1965-19764

4 The estimate of total fissions is based on 27 days of continuous run time occurring in the middle of the 12–

year period and corresponds with 7.7×1011 fissions per second.

 and 

an average power of 25 W (Log, 1976, PDF pp. 2-4). However, the senior scientist who 

performed the measurements and derived the estimate noted that the gamma photon background 

from a new unirradiated Oy part was the same as the gamma flux measured from irradiated parts 

used for the analysis. This means that measurements, made six years after the last criticality 

experiment, were unable to detect additional gammas from prior experiments because of 

radioactive decay of the fission and activation products. The original estimate did not take into 

account the intrinsic gamma photon emissions from unirradiated enriched uranium and was, 

therefore, invalid5

5 The laboratory notebook entry reads: “All γ due to fission 6 years ago. (No believable background available.) 

Note: Part no. 80 never used and yet had same background/kg as others. [Therefore] Feel fission too low to detect 

by this method.” 

 (Log, 1976, PDF p. 2). 

A second estimate of 3.44×1010 fissions in an average Oy + UNH experiment with a 20-minute 

run time was derived by estimating the number of neutrons produced (8.4×1010) using the current 

generated by a neutron detector associated with an April 1977 experiment (Experiments at 

RFNSF, 1977b, PDF p. 3). This fission rate corresponds with a thermal power level of 0.92 

mW.

Thermal Power and Fission Rate Estimates in Solution-Only Experiments

The CML staff’s most rigorous estimates of fission rates were derived for UNH (solution-only) 

fuel by measuring the rate of gamma decay in the fuel of Experiment 2-8-170. This experiment 

was configured as a 2×2 array of 8-in-diameter tanks and was performed on May 4, 1977 

(Experiments at RFNSF, 1977a). The gamma photon count rate of a 2-mL aliquot withdrawn 

from the fuel after a 70.5-minute run was determined at 2- to 3-minute intervals using a “well 

crystal” calibrated to Cs-137 photons. The total number of fissions in the reactor was calculated 

 
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ABRWH (or their technical support and review contractors) positions unless specifically marked as such. This document represents preliminary 

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information that is protected by the Privacy Act 5 USC §552a and has been cleared for distribution. 

from the measured photon emission rate (Analysis Report, 1977, PDF p. 7) using Equation 1 

below: 

Where: 

Φ̇

γγ = Reactor gamma photon emission rate at time τ 

τ = Days after shutdown 

An initial evaluation6

6

 Reactor power was evaluated using the integral of Φ̇

γγ over the operating time of the reactor, giving the gamma 

energy emission rate as a function of reactor power and time after shutdown.

of the data (Introducing Nuclear Reactor Theory, 1977) arrived at a power 

estimate of 38 mW during the last 10 minutes of the experiment (when most of the 

corresponding 7.1×1011 fissions occurred), giving an average power of 5.4 mW for the 

70.5-minute duration of the experiment. 

 
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with the ABRWH or its Working Groups or Subcommittees. Draft, preliminary, interim, and White Paper documents are not final NIOSH or 

ABRWH (or their technical support and review contractors) positions unless specifically marked as such. This document represents preliminary 

positions taken on technical issues prepared by NIOSH or its contractor. NOTICE: This report has been reviewed to identify and redact any 

information that is protected by the Privacy Act 5 USC §552a and has been cleared for distribution. 

A subsequent June 7, 1977 estimate from Experiment 2-8-170 employed more precise values for 

the aliquot and reactor volumes (Analysis Report, 1977, PDF p. 4). The estimate gave 25 mW at 

the end of the experiment, corresponding with 4.73×1011 fissions (or 3.6 mW, average power). 

These results were within 2% of those obtained by the method of source multiplication 

(Summary of All Nuclear Safety Experiments, 1977; Experiment Run 2-8-170, Undated), using 

Equation 2 below: 

Where: 

f = Total number of fissions over time t 

t = Time

μμ̅= Average multiplication from the neutron detector response during the experiment 

s = Beginning neutron flux from the Cf-252 seed source 

On June 3, 1977, before the more precise June 7 calculation was completed, CML staff 

communicated officially to the Energy Research and Development Administration (ERDA). 

They provided an estimate of 8.8×1011 fissions and an average power of 6.7 mW for a typical 

(70.5 minute) experiment (Tuck, 1977). There is no indication why the values reported to ERDA 

were almost 25% higher than results of the May 4, 1977 calculation.
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with the ABRWH or its Working Groups or Subcommittees. Draft, preliminary, interim, and White Paper documents are not final NIOSH or 

ABRWH (or their technical support and review contractors) positions unless specifically marked as such. This document represents preliminary 

positions taken on technical issues prepared by NIOSH or its contractor. NOTICE: This report has been reviewed to identify and redact any 

information that is protected by the Privacy Act 5 USC §552a and has been cleared for distribution.

Conclusions about Reactor Power and Fission Rates in CML Experiments

Table 2 provides a summary of the different CML criticality experiments for which thermal 

power and fission rate were estimated from measurements.

Table 2: Average Reactor Power and Fission Rates Derived from Measurements at CML

Configuration Duration Average 

Power (mW)

Total 

Fissions

Fission Rate 

(s-1) Comment

Oy 60 min 1.9508 2.18E+11 6.06E+07 Based on exposure rate at the surface of an Oy 

sphere converted to γ photon flux.

Oy + UNH (a) 27 d 25,000 7.7E+11 3.30E+05 Estimate is invalid; measurement results were 

indistinguishable from background.

Oy + UNH (b) 20 min 0.92 3.44E+10 2.83E+07 From neutron detector current.

UNH (a) 70.5 min 5.4 7.1×1011 1.68E+08* Initial calculation on May 4, 1977, for Experiment 

2-8-170

UNH (b) 70.5 min 6.7 8.8E+11 2.08E+08* Values reported to ERDA on June 3, 1977, based 

on results of Experiment 2-8-170.

UNH (c) 70.5 min 3.6 4.73E+11 1.12E+08

Recalculated values from Experiment 2-8-170 on 

June 7, 1977. Result is within 2% of that obtained 

from evaluation of source multiplication.

Source: Log, 1967, PDF p. 6 (Oy); Log, 1976 [Oy + UNH (a)]; Experiments at RFNSF, 1977b [Oy + UNH (b)];

Analysis Report, 1977, PDF p. 7 [UNH (a)]; Tuck, 1977 [UNH (b)]; Analysis Report, 1977, PDF p. 4 [UNH (c)].

*These values are not reported in the original reference, but are calculated from total fissions and duration.

CML staff described Experiment 2-8-170 to ERDA as having a higher-than-normal power level

(Tuck, 1977, PDF p. 2). A thermal power estimate of 3.6 mW averaged over 70.5 minutes

(corresponding with 4.73×1011 total fissions) is based on careful evaluation of measurements 

made on Experiment 2-8-170. NIOSH considers this the most precise estimate available for this 

relatively high-power experiment. CML staff, however, reported to ERDA an average thermal 

power of 6.7 mW over 70.5 minutes (8.8×1011 total fissions) based on the same experiment. 

NIOSH has concluded that the value reported to ERDA is the most appropriate value to represent 

the typical CML criticality experiments for the purpose of estimating doses from fission and 

activation products. This value is less than the earlier assumption that power averaged 10 mW 

for experiments lasting an hour (1.1×1012 total fissions). However, the value is still claimantfavorable because it exceeds the most careful estimate of power for the experiment, and because

not all CML experiments achieved criticality (Building 886, 2011, PDF p. 3).

CML SURFACE CONTAMINATION

Radiological survey results

Internal doses from the estimated MFAP inventory were bounded in the earlier estimate by 

assuming that removable surface contamination from dried solution spills was evenly dispersed 

over CML surfaces at the DOE limit (1,000 dpm/100 cm2

) for a posted Contamination Area. 

The average air concentration of respirable particles was then estimated by applying a 
White Paper: Reassessment of Internal Dose from Sources at the RFP CML

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with the ABRWH or its Working Groups or Subcommittees. Draft, preliminary, interim, and White Paper documents are not final NIOSH or 

ABRWH (or their technical support and review contractors) positions unless specifically marked as such. This document represents preliminary 

positions taken on technical issues prepared by NIOSH or its contractor. NOTICE: This report has been reviewed to identify and redact any 

information that is protected by the Privacy Act 5 USC §552a and has been cleared for distribution. 

resuspension factor. Recently captured documents contain both air monitoring and surface 

contamination measurement results against which the prior assumptions can be reevaluated.

Radiological survey results for Building 886 and 875 over the period January 1981 – December 

1990 are captured in the following 105 files: 

Radiological Surveys, 1980a-1980l; 1981a-1981m; 1982a-1982l; 1983a-1983k; 1984a-1984m; 

1985a-1985m; 1986a-1986l; 1987a-1987e; 1988a-1988e; 1989a-1989g; 1990a, 1990b. 

Information in these files includes results of daily removable contamination surveys in the office

and experimental areas, and on equipment, as well as weekly penetrating-dose measurements in 

the nuclear material storage and experimental areas of Building 886. 

Maximum Reported Values

Based on the data in the files listed in the subsection above, Table 3 shows the maximum 

reported values of removable alpha surface contamination for each year.

Table 3: Maximum Removable Alpha Contamination (dpm/100 cm2)

Year

Bldg. 886 

Offices*

Bldg. 886 

Experimental 

Area**

Bldg. 886

Equipment

Bldg. 875

Outside 

Plenum

Bldg. 875

Inside 

Plenum

1981 30 366 156 12 11196

1982 18 174 (3930) -- 12 14346

1983*** <20 576 -- -- --

1984 <20 54000 -- <20 15660

1985 <20 174 -- <20 4788

1986 18 186 (2784) 3954 12 5394

1987 18 (24) 564 -- -- --

1988 12 1683 -- 12 15240

1989  18 18707 -- -- --

1990** ** 18 16212 -- 6 --

Source: See the list of 105 files in the preceding subsection.

NOTE: Values in parentheses are singular outliers and not representative.

* “Offices” includes control points (corridors and common areas) typically surveyed daily and individual 

offices surveyed weekly.

**The experimental area (Rooms 101, 102 and 103) were surveyed weekly in 1981, but the frequency 

gradually diminished, apparently in response to decreased activities in these areas. Surveys were made 

approximately monthly by 1990.

***Data in 1983 are available only for November and December.

****Elevated values of this order were observed in weekly surveys of the Room 101 walk-in hood from 

December 1989 through April 1990, as discussed in the text.
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with the ABRWH or its Working Groups or Subcommittees. Draft, preliminary, interim, and White Paper documents are not final NIOSH or 

ABRWH (or their technical support and review contractors) positions unless specifically marked as such. This document represents preliminary 

positions taken on technical issues prepared by NIOSH or its contractor. NOTICE: This report has been reviewed to identify and redact any 

information that is protected by the Privacy Act 5 USC §552a and has been cleared for distribution. 

Based on the data in the files listed in the subsection above, Table 4 shows maximum gamma 

and neutron penetrating dose rates. 

Table 4: Maximum Dose Rates in CML Experimental and Material Storage Areas

Year

Gamma* 

(mR/h): 

UNH Tank 

Room

Gamma* 

(mR/h): 

Solid Fuel 

Storage

Gamma* 

(mR/h): 

Experiment 

Room

Neutron** 

(mrem/h): 

UNH Tank 

Room

Neutron** 

(mrem/h): 

Solid Fuel 

Storage

Neutron** 

(mrem/h): 

Experiment 

Room

1981 2.0 2.9 -- 0.3 12.9 --

1982 2.0 2.5 3.7 0.2 14.7 5.4

1983 -- -- -- -- -- --

1984 5.0 3.2 -- 0.2 0.6 --

1985 2.0 2.8 -- 0.1 0.5 --

1986 2.0 6.0 0.4 1.0 2.6 0.2

1987 3.0 2.0 3.0 1.7 2.0 3.4

1988 2.4 1.5 1.8 2.0 0.8 0.6

1989*** 2.8 4.2 (250) 0.0 1.0 1.8 (6.0) 0.6

1990 2.53 3.0 -- 0.3 0.5 --

Source: See the list of 105 files in the preceding subsection.

NOTE: Values in parentheses are singular outliers and not representative.

*Gamma dose rate surveys were made at least weekly during the entire 10-year period.

** Neutron dose rate surveys were made weekly, along with gamma surveys, until 1990 when the frequency was 

reduced to monthly.

***Atypically-elevated dose-rate readings reported for July 6, 1989 are associated with a survey location within an 

area marked off inside the solid storage area.

Evaluation of Removable Contamination Survey Results

Contamination surveys were conducted daily at control points (hallways, doorways to offices), 

and weekly in offices within the unrestricted-access portion of Building 886. Surveys were 

conducted weekly in the experimental areas. Results of periodic contamination surveys in 

Building 875 and its air filtration plenums are also documented. Survey results for the office 

area are captured through December 20, 1990 (Radiological Surveys, 1990a, PDF p. 778), and 

for the experimental and material storage areas through April 1990 (Radiological Surveys, 

1990a, PDF p. 330). No average values were found, except in 1989 documents, which contained 

a few graphs showing average contamination values in different areas for limited time periods 

(Radiological Surveys, 1989b, PDF pp. 3-6). These graphs are shown in Figures 2 through 5 

below. 
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ABRWH (or their technical support and review contractors) positions unless specifically marked as such. This document represents preliminary 

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information that is protected by the Privacy Act 5 USC §552a and has been cleared for distribution. 

Table 3 above, showing maximum values for the removable alpha survey results in both the 

office and experimental areas, demonstrates that radiological containment of the contaminated 

areas was quite effective. Excursions of removable contamination above the 20 dpm\100cm2

DOE limit in “cold” (office) areas is seen to occur only a few times (and only at low levels) in 

1981 and once in 1984. Survey results in offices and hallways were typically well below 

20 dpm/100 cm2

, particularly toward the end of the period from 1981-1990 (e.g., see Figures 4 

and 5). 

Source: Radiological Surveys, 1989b, PDF p. 3

Figure 2: Weekly Averages of Removable Alpha Contamination in the Building 886 

Experimental and Material Storage Areas - Q1 1989
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ABRWH (or their technical support and review contractors) positions unless specifically marked as such. This document represents preliminary 

positions taken on technical issues prepared by NIOSH or its contractor. NOTICE: This report has been reviewed to identify and redact any 

information that is protected by the Privacy Act 5 USC §552a and has been cleared for distribution. 

Source: Radiological Surveys, 1989b, PDF p. 5 

Note: The solid line represents removable alpha contamination; the dashed line represents the DOE limit for 

uranium.

Figure 3: Average Removable Alpha Contamination in the Building 886 Experimental and Material 

Storage Areas – Q2 1989
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with the ABRWH or its Working Groups or Subcommittees. Draft, preliminary, interim, and White Paper documents are not final NIOSH or 

ABRWH (or their technical support and review contractors) positions unless specifically marked as such. This document represents preliminary 

positions taken on technical issues prepared by NIOSH or its contractor. NOTICE: This report has been reviewed to identify and redact any 

information that is protected by the Privacy Act 5 USC §552a and has been cleared for distribution. 

Source: Radiological Surveys, 1989b, PDF p. 6

Figure 4: Weekly Averages of Removable Alpha Contamination in the Building 886 Offices and Control 

Points (Hallways and Conference Rooms) – Q1 1989
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with the ABRWH or its Working Groups or Subcommittees. Draft, preliminary, interim, and White Paper documents are not final NIOSH or 

ABRWH (or their technical support and review contractors) positions unless specifically marked as such. This document represents preliminary 

positions taken on technical issues prepared by NIOSH or its contractor. NOTICE: This report has been reviewed to identify and redact any 

information that is protected by the Privacy Act 5 USC §552a and has been cleared for distribution. 

Source: Radiological Surveys, 1989b, PDF p. 4

Figure 5: Average Removable Alpha Contamination in the Building 886 Office Area for January-June 1989

Review of the survey documents indicates that excursions of removable alpha contamination 

above the Contamination Area limit in the experimental area (Room 101) and the material 

storage area (Rooms 102, and 103) were usually reduced below 1,000 dpm/100 cm2 within a few 

days, as noted below for entries in Table 3. 

• The survey sheet showing 3,930 dpm/100 cm2

 for removable contamination in Room 103 

(the Mixing Room where uranium solution was housed) on September 10, 1982, for instance, has 

notation showing that the area was decontaminated to 12 dpm/100 cm2 on the same day 

(Radiological Surveys, 1982c, PDF p. 23). 

• A survey result of 54,000 dpm/100 cm2

 on February 27, 1984 (the highest found in any of 

the captured documents), was located in the walk-in hood of Room 101 (Radiological Surveys, 

1984c, PDF p. 13). A second survey sheet with the same date shows <20 dpm/100 cm2 for all 
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survey locations in Rooms 101, 102 and 103, and bears he hand-written note, “Retake of survey 

after [name redacted] and [name redacted] cleaned all of Rm 101.” 

• The high value of 2,784 dpm/100 cm2 in 1986 is reported on a survey sheet dated June 17, 

along with the handwritten notation, “Table Top 3954 d/m/100 cm2; Rest of Table 

25.8 d/m/100 cm2

; deconned 6/18/86” (Radiological Surveys, 1986c, PDF p. 8). The survey 

sheet for June 18, 1986 indicates that all results in the controlled area were ≤ 24 dpm/100 cm2

(Radiological Surveys, 1986c, PDF p. 9).

• The 1988 high result of 1,683 dpm/100 cm2 was found just outside the walk-in hood of 

Room 101 on October 7, 1988 (Radiological Surveys, 1988d, PDF p. 37). A similar elevated 

result (1,422 dpm/100 cm2) was found inside the hood. Contamination at the location outside the 

hood was reduced to 60 dpm/100 cm2 a week later, on October 14 (Radiological Surveys, 1988d, 

PDF p. 38), but the elevated value inside the walk-in hood persisted. It was reduced to 

126 dpm/100 cm2

 in the October 21 survey report (Radiological Surveys, 1988d, PDF p. 39). 

Prompt decontamination practices were evident during most of the period from 1981-1990; 

however, elevated contamination levels in the Room 101 walk-in hood were recorded in 

December 1989 and persisted through April 1990. The persistent contamination is reflected in 

the maximum-recorded values of 1989 and 1990 shown in Table 3 above, and is detailed in the 

weekly contamination survey results presented in Table 5 below. The FBI raid of Rocky Flats in 

June 1989 resulted in the curtailment of plutonium operations. Subsequently, the plant’s 

resources were redirected toward remediation of the issues resulting from the raid. According to 

the document, A Technically Useful History of the Critical Mass Laboratory at Rocky Flats, 

“Manpower was not available to decontaminate Room 101 in the late 1980s because 

Maintenance personnel had all been dedicated to solving the plant’s larger problems” (Rothe, 

2005, PDF p. 395). Handwritten notes on the survey sheets sometime indicate that full-face 

respirator protection was required in Room 101 when contamination levels exceeded the DOE 

limit (e.g., Radiological Surveys, 1989b, PDF p. 109). Figure 6 shows the floor plan for 

Room 101, where criticality experiments were performed; the walk-in hood is the enclosure left 

of center.
White Paper: Reassessment of Internal Dose from Sources at the RFP CML

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This is a working document prepared by NIOSH’s Division of Compensation Analysis and Support (DCAS) or its contractor for use in discussions 

with the ABRWH or its Working Groups or Subcommittees. Draft, preliminary, interim, and White Paper documents are not final NIOSH or 

ABRWH (or their technical support and review contractors) positions unless specifically marked as such. This document represents preliminary 

positions taken on technical issues prepared by NIOSH or its contractor. NOTICE: This report has been reviewed to identify and redact any 

information that is protected by the Privacy Act 5 USC §552a and has been cleared for distribution. 

Source: Rothe, 2005, PDF p. 120

Figure 6: Floor Plan of the Assembly Room (Room 101), Showing the Walk-in Hood Containing the 

Vertical Split Table (V) and Solution Base (S) Experimental Locations
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This is a working document prepared by NIOSH’s Division of Compensation Analysis and Support (DCAS) or its contractor for use in discussions 

with the ABRWH or its Working Groups or Subcommittees. Draft, preliminary, interim, and White Paper documents are not final NIOSH or 

ABRWH (or their technical support and review contractors) positions unless specifically marked as such. This document represents preliminary 

positions taken on technical issues prepared by NIOSH or its contractor. NOTICE: This report has been reviewed to identify and redact any 

information that is protected by the Privacy Act 5 USC §552a and has been cleared for distribution. 

Table 5: Room 101 Walk-In Hood Contamination Results, November 1989 – April 1990

Date

Location 20* 

(dpm/100 cm2)

Location 21* 

(dpm/100 cm2)

Location 22* 

(dpm/100 cm2) Reference PDF p.

11/21/1989 27 642 --

Radiological 

Surveys, 1989b 92

11/30/1989 239 -- --

Radiological 

Surveys, 1989b 116

12/7/1989 897 12579 18707

Radiological 

Surveys, 1989b 160

12/14/1989 9 100 90

Radiological 

Surveys, 1989b 183

12/21/1989 72 51 5235

Radiological 

Surveys, 1989b 196

1/18/1990 168 693 6990

Radiological 

Surveys, 1990a 60

1/25/1990 84 672 6963

Radiological 

Surveys, 1990a 81

2/1/1990 93 1230 12219

Radiological 

Surveys, 1990a 99

2/8/1990 105 1290 13452

Radiological 

Surveys, 1990a 117

2/15/1990 255 15099 10839

Radiological 

Surveys, 1990a 135

2/22/1990 717 12369 9978

Radiological 

Surveys, 1990a 156

3/1/1990 639 14163 6933

Radiological 

Surveys, 1990a 171

3/8/1990 616 14193 6987

Radiological 

Surveys, 1990a 183

3/15/1990 468 12423 7923

Radiological 

Surveys, 1990a 207

3/22/1990 1035 16212 5403

Radiological 

Surveys, 1990a 234

3/29/1990 1893 15708 7842

Radiological 

Surveys, 1990a 252

4/5/1990 2748 12951 8967

Radiological 

Surveys, 1990a 273

4/19/1990 2556 14967 13683

Radiological 

Surveys, 1990a 291

4/12/1990 1572 13593 10353

Radiological 

Surveys, 1990a 300

4/26/1990 2037 14463 11973

Radiological 

Surveys, 1990a 330

*Locations 20, 21, and 22 correspond with the entry portal, Solution Base (S), and Vertical Split Table (V), 

respectively, as shown on the floor plan in Figure 6 above.
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with the ABRWH or its Working Groups or Subcommittees. Draft, preliminary, interim, and White Paper documents are not final NIOSH or 

ABRWH (or their technical support and review contractors) positions unless specifically marked as such. This document represents preliminary 

positions taken on technical issues prepared by NIOSH or its contractor. NOTICE: This report has been reviewed to identify and redact any 

information that is protected by the Privacy Act 5 USC §552a and has been cleared for distribution. 

Air flow through Room 101 came from outside through the walk-in hood to the exhaust plenum 

in Building 875, and through HEPA filters before being exhausted to the outside air. The 

walk-in hood was, therefore, at a lower pressure than Room 101, drawing resuspended 

contamination to the exhaust plenum and not into the rest of Room 101. This engineered feature

reduced exposure to Room 101 personnel and, along with workforce priorities redirected as a 

result of the FBI raid, probably explains why the hood was not decontaminated in late 1989 and 

1990. The last CML criticality experiment concluded in October 1987 (Rothe, 2005, PDF 

p. 393); no routine work was performed in Room 101 after this date, further reducing the 

potential for personnel exposure as the result of contamination in the walk-in hood in 1989 and 

1990. 

Additional CML Personnel Identified for 1989 

The 1989 files were different from those in other years because the forms (Radiological Surveys, 

1989b) were signed by the RCTs who were making the measurements. Based on the additional 

research performed for this review, NIOSH has identified no NOCTS claims associated with the 

seven individuals signing the survey forms. 

Discovery of Sealed Sr-90 Sources 

The file of survey results for June 1982 contains a copy of a note detailing radiological 

characteristics of three sealed Sr-90 sources (Radiological Surveys, 1982k, PDF p. 3). The note 

is reproduced as Figure 7 below. It is the only documentation in hand indicating the presence of 

Sr-90 at CML. There is no information about why the sources were at CML, how long they 

might have been there, or their disposition. Integrity of the source encapsulations was confirmed 

by removable contamination surveys of the sources and the cabinet in which they were found. 
White Paper: Reassessment of Internal Dose from Sources at the RFP CML

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with the ABRWH or its Working Groups or Subcommittees. Draft, preliminary, interim, and White Paper documents are not final NIOSH or 

ABRWH (or their technical support and review contractors) positions unless specifically marked as such. This document represents preliminary 

positions taken on technical issues prepared by NIOSH or its contractor. NOTICE: This report has been reviewed to identify and redact any 

information that is protected by the Privacy Act 5 USC §552a and has been cleared for distribution. 

Source: Radiological Surveys, 1982k, PDF p. 3 

Figure 7: Memo on β-γ Penetrating and Removable Surface Contamination Measurements Made upon 

Discovering Three Sealed Sr-90 Sources in the Building 886 Material Storage Area
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This is a working document prepared by NIOSH’s Division of Compensation Analysis and Support (DCAS) or its contractor for use in discussions 

with the ABRWH or its Working Groups or Subcommittees. Draft, preliminary, interim, and White Paper documents are not final NIOSH or 

ABRWH (or their technical support and review contractors) positions unless specifically marked as such. This document represents preliminary 

positions taken on technical issues prepared by NIOSH or its contractor. NOTICE: This report has been reviewed to identify and redact any 

information that is protected by the Privacy Act 5 USC §552a and has been cleared for distribution. 

Conclusions About Surface Contamination at CML

Surveys for removable alpha contamination at CML were conducted regularly. Excursions 

above the applicable DOE limits (20 dpm/100 cm2

 in uncontrolled office areas and 

1,000 dpm/100 cm2 in controlled-access experimental areas) were confined to discrete areas and 

were quickly decontaminated below the limits. Values above the limit in uncontrolled areas 

were only rarely observed. The largest measured value for removable contamination was 

54,000 dpm/100 cm2

, found at sampling location 21 in the walk-in hood (Radiological Surveys, 

1984c, PDF pp. 13, 14). This amount of contamination, if spread uniformly over the entire 

surface area7

7 Dimensions of the walk-in hood (Rothe, 2005, PDF p. 129) in Room 101 were 3.0 m × 4.9 m. Sampling 

location 21 was located in a part of the hood comprising about half this surface area, or 7.35 m2

. Estimated floor 

areas for the individual rooms are 120 m2 for Room 101 (Rothe, 2005, PDF p. 120); 40 m2 for Room 103 (Rothe, 

2005, PDF p. 167); and 60 m2 for Rooms 102 and 108 (connecting hallway) combined, assuming this area to be 

about half that of Room 101 (Rothe, 2005, PDF p. 110). The total estimated floor area is the sum of these values, or 

220 m2

. Distributing the contamination at sampling location 21 over the entire surface area would result in an 

average contamination level reduced by the ratio 7.35/220.

 of the controlled area, would be 1,800 dpm/100 cm2; it was decontaminated on the 

same day it was found. All other values for removable contamination captured by NIOSH 

correspond to <1,000 dpm/100 cm2, if distributed uniformly. All were quickly decontaminated, 

except for that in the walk-in hood from November 1989 – April 1990, which was contained and 

ventilated during a time of little or no personnel activity in Room 101. 

In light of the above information, the assumption that average removable contamination 

available for resuspension in the experimental and material storage areas (Rooms 101, 102, and 

103) was equal to or less than the Contamination Area limit of 1,000 dpm/100 cm2 is claimantfavorable.

CML WORKPLACE AIR MONITORING

Bounds on internal dose from MFAP were previously based on airborne concentrations 

calculated by applying a resuspension factor to surface contamination limits posted for the 

facility. This approach was based on interview comments from a former Radiation Safety 

Supervisor, who said that no routine air monitoring was performed at the CML prior to 1990, 

when this individual was assigned responsibility for Building 886 (Personal Communication, 

2015b, PDF pp. 4, 7). NIOSH has since captured formal plant-wide procedures describing a 

particulate air monitoring program during the period from 1980-1989 for alpha-particle 

emissions from uranium, plutonium, and americium at sampling locations selected by process 

knowledge or professional judgment. Additional captured documents indicate that these 

procedures appear to have been followed and that routine alpha air monitoring was performed at 

the CML during the period 1980-1989. 

 
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This is a working document prepared by NIOSH’s Division of Compensation Analysis and Support (DCAS) or its contractor for use in discussions 

with the ABRWH or its Working Groups or Subcommittees. Draft, preliminary, interim, and White Paper documents are not final NIOSH or 

ABRWH (or their technical support and review contractors) positions unless specifically marked as such. This document represents preliminary 

positions taken on technical issues prepared by NIOSH or its contractor. NOTICE: This report has been reviewed to identify and redact any 

information that is protected by the Privacy Act 5 USC §552a and has been cleared for distribution. 

Rocky Flats Plant Workplace Air Monitoring Requirements 

Formal procedures describe monitoring requirements and practices for routinely evaluating 

concentrations of alpha-emitting particulates (U, Pu, and Am) in Rocky Flats workplace air over 

the period from 1979 through 1990 (Radiation Monitoring Technical Support Programs, 1976, 

PDF pp. 19-136). The earliest captured procedure (issued November 1979) and its subsequent 

revisions through December 1990 are listed below:

• Routine Air Sampling, HS-RM-4.1, November 13, 1979 (Radiation Monitoring Technical 

Support Programs, 1976, PDF pp. 19-29) 

• Routine Air Sampling, RMPM 4.1, August 2, 1982 (Radiation Monitoring Technical 

Support Programs, 1976, PDF pp. 30-43) 

• Routine Air Sampling, RMPM 4.1, January 13, 1989, Replaces: August 2, 1982 

(Radiation Monitoring Technical Support Programs, 1976, PDF pp. 65-81) 

• Routine Air Sampling, RMPM 4.1, June 1989, Replaces: January 13, 1982 (Radiation 

Monitoring Technical Support Programs, 1976, PDF pp. 44-64) 

• Routine Air Sampling, ROI 4.1, December 18, 1989, Replaces: June 15, 1989 (Radiation 

Monitoring Technical Support Programs, 1976, PDF pp. 102-118) 

• Routine Air Sampling, ROI 4.1, Rev. 3, December 20, 1990, Replaces: December 19, 

1989 (Radiation Monitoring Technical Support Programs, 1976, PDF pp. 119-136) 
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with the ABRWH or its Working Groups or Subcommittees. Draft, preliminary, interim, and White Paper documents are not final NIOSH or 

ABRWH (or their technical support and review contractors) positions unless specifically marked as such. This document represents preliminary 

positions taken on technical issues prepared by NIOSH or its contractor. NOTICE: This report has been reviewed to identify and redact any 

information that is protected by the Privacy Act 5 USC §552a and has been cleared for distribution. 

Review of the procedures shows consistent requirements for calibrating monthly both Gross 

Alpha (Continuous) Air Monitors (CAMs) and Selective Alpha Air Monitors (SAAMs). CAM 

filters were collected and sent for analysis each weekday (Monday – Friday), except holidays, 

until at least January 1989, when RMPM 4.1 specified that the exchange frequency was to be 

determined by Operational Health Physics personnel (Radiation Monitoring Technical Support 

Programs, 1976, PDF p. 67). Subsequent revisions assigned the responsibility for establishing 

sampling frequencies to Building 123 Count Room personnel (Radiation Monitoring Technical 

Support Programs, 1976, PDF p. 49), Operational Health Physics (Radiation Monitoring 

Technical Support Programs, 1976, PDF p. 104), or to Radiological Engineering personnel 

(Radiation Monitoring Technical Support Programs, 1976, PDF p. 121). When collected, filters 

were monitored with a hand-held alpha survey instrument. Filters with excessive alpha activity 

were held for specified periods before analysis to allow decay of the short-lived radon and thoron 

progeny; a time-dependent Koval factor was applied to the analytical results to account for 

undecayed progeny (Radiation Monitoring Technical Support Programs, 1976, PDF pp. 22-23, 

36-37, 52, 72). Final results were reported as percentages of the radioactive concentration guide 

(RCG) airborne limit for the material in question, or 70 dpm/m3 for uranium8

8 The RCG used by Rocky Flats in its air-sampling program corresponds with the Maximum Allowable (air) 

Concentration (MAC) used by AEC contractors in referring to the AEC’s “preferred level” of 50 μg/m3 for all 

uranium compounds on the basis of chemical toxicity. The level was also stated as 70 dpm/m3 for natural uranium. 

(ORAUT-OTIB-0004, 2006, PDF p. 8)

 (Radiation 

Monitoring Technical Support Programs, 1976, PDF pp. 25, 39, 73). In June 1989, the airborne 

limit for uranium (in any form) was changed to 44 dpm/m3, corresponding to the DOE Order 

5480.11 Derived Air Concentration (Radiation Monitoring Technical Support Programs, 1976, 

PDF pp. 54, 104, 121). 

Air Sampling Locations 

Air particulate samplers were located in Buildings 886 and 875, as marked on monthly air head 

calibration sheet maps (Calibrations, 1987; Calibrations, 1988; Calibrations, 1989a; Calibrations, 

1989b); Figure 8 shows an example. The locations appear to be unchanged during the period 

from 1980-1989. Building 875 housed the air exhaust plenums from the Building 886 

experimental (Room 101) and material storage (Rooms 102 and 103) areas. The air head 

samplers were numbered RR-1 through RR-11 and fixed in the locations listed below: 

Building 886 

• RR-1 through RR-4, Room 103 (material storage) 

• RR-5, Room 102 (material storage) 

• RR-6 and RR-7, Room 101 (experimental area)

 
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with the ABRWH or its Working Groups or Subcommittees. Draft, preliminary, interim, and White Paper documents are not final NIOSH or 

ABRWH (or their technical support and review contractors) positions unless specifically marked as such. This document represents preliminary 

positions taken on technical issues prepared by NIOSH or its contractor. NOTICE: This report has been reviewed to identify and redact any 

information that is protected by the Privacy Act 5 USC §552a and has been cleared for distribution. 

Building 875 

• RR-8 and RR-9, outside the plenum 

• RR-10, inside the first plenum 

• RR-11, inside the tunnel from Building 886 

SAAM samplers also were located in Rooms 101-103 of Building 886 and in the second plenum 

of Building 775. The detectors on these samplers were equipped with high-voltage 

discriminators that registered only alpha energies above a certain threshold, providing a means to 

discriminate against lower-energy alpha emitters that would otherwise give false alarms. The 

filter media from SAAMs were normally discarded without analysis unless the instrument had 

alarmed or the filter failed a hand-held radiation monitor check (Radiation Monitoring Technical 

Support Programs, 1976, PDF pp. 26, 35, 51, 70, 88, 105-106, 123). 

Other air sampling results from 1981-1988 for samples labeled “875-A”, “875-B” and “875-C” 

have also been captured but their locations in Building 875 have not been identified (Air 

Samples, 1981a, 1981b; 1982a-1982i; 1983a, 1983b; 1984a-1984d; 1985a-1985e; 1986a-1986m; 

1987a; 1988a-1988g).

CML personnel did not routinely access Building 875, and consideration of air monitoring 

results for this building is limited to those in the “RR” series, with known locations and direct 

correlation with daily samples from Building 886. 
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ABRWH (or their technical support and review contractors) positions unless specifically marked as such. This document represents preliminary 

positions taken on technical issues prepared by NIOSH or its contractor. NOTICE: This report has been reviewed to identify and redact any 

information that is protected by the Privacy Act 5 USC §552a and has been cleared for distribution. 

Source: Calibrations, 1987, PDF p. 6

Figure 8: Air Sampler Locations in the Building 886 Cluster

Air Sampling Results

NIOSH has captured air sample results in 41 of 120 months during the period 1980-1989. 

Results from samplers RR-1 through RR-7 (Building 886) and RR-8 through RR-11 (Building 

875) are available for the period December 1980 to June 1989, although all but two of the results 

fall in the period from May 1983 to November 1988. The data, summarized in Table 6, indicate 

that procedural requirements for daily air monitoring appear to have been met at the CML. The 

number of analytical results per month for a particular sampler location is expected to be 

between 19 and 22, depending on the number of work days. In the captured data, daily sample 

collection is demonstrated with certainty (results available for 19 or more days) in 15 of the 41 

months (37% of the time for which records are available, but only 1% of the period from 
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ABRWH (or their technical support and review contractors) positions unless specifically marked as such. This document represents preliminary 

positions taken on technical issues prepared by NIOSH or its contractor. NOTICE: This report has been reviewed to identify and redact any 

information that is protected by the Privacy Act 5 USC §552a and has been cleared for distribution. 

1980-1989). Sample collection on at least alternate days, on average (results available for 10-18 

days), is demonstrated with certainty during an additional 10 months. The availability of air 

sampling data does not seem to correlate with known spills or criticality experiments. Only one 

sampling day’s results were captured during operations prior to 1983, but 207 sampling days 

correspond with three operational periods after 1983. Four contamination incidents occurred 

from 1980-1989, but air monitoring results were only captured for the time period in which two 

incidents occurred closely together in 1987. Experiment periods, contamination incident dates, 

and associated records of airborne contamination results are summarized in Table 7. 
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This is a working document prepared by NIOSH’s Division of Compensation Analysis and Support (DCAS) or its contractor for use in discussions 

with the ABRWH or its Working Groups or Subcommittees. Draft, preliminary, interim, and White Paper documents are not final NIOSH or 

ABRWH (or their technical support and review contractors) positions unless specifically marked as such. This document represents preliminary 

positions taken on technical issues prepared by NIOSH or its contractor. NOTICE: This report has been reviewed to identify and redact any 

information that is protected by the Privacy Act 5 USC §552a and has been cleared for distribution.

Table 6: Air Sample Results Summary for the CML Building 886 Nuclear Materials 

Storage and Experimental Areas

Year Month References Sample Days* No. of Days 

10%-99% RCG**

No. of Days ≥100% 

RCG**

1980 Dec Air Samples, 

1980 1 -- --

1983 May Air Samples, 

1983c 2 -- --

1983 Sep Air Samples, 

1983d 1 -- --

1983 Oct Air Samples, 

1983e 17 2 --

1984 Jul Air Samples, 

1984e 2 -- --

1984 Aug Air Samples,

1984f 22 1 1

1984 Sep Air Samples, 

1984g 19 2 --

1984 Oct Air Samples, 

1984h 20 6 --

1984 Nov Air Samples, 

1984i 22 8 3***

1984 Dec Air Samples, 

1984j 15 5 --

1985 Jan Air Samples, 

1985f 19 4 --

1985 Feb Air Samples, 

1985g 19 8 2

1985 Mar Air Samples, 

1985h 22 1 --

1985 Apr Air Samples, 

1985i 21 -- --

1985 May Air Samples, 

1985j 22 3 --

1985 Jun Air Samples, 

1985k 22 -- --

1985 Jul Air Samples, 

1985l 23 2 --

1985 Aug Air Samples, 

1985m 20 2 --

1985 Sep Air Samples, 

1985n 15 -- --

1985 Oct Air Samples, 

1985o 9 -- --

1985 Nov Air Samples, 

1985p 19 1 --

1985 Dec Air Samples, 

1985q 15 -- --

1986 Jan Air Samples, 

1986n 19 -- --
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ABRWH (or their technical support and review contractors) positions unless specifically marked as such. This document represents preliminary 

positions taken on technical issues prepared by NIOSH or its contractor. NOTICE: This report has been reviewed to identify and redact any 

information that is protected by the Privacy Act 5 USC §552a and has been cleared for distribution. 

Year Month References Sample Days* No. of Days 

10%-99% RCG**

No. of Days ≥100% 

RCG**

1986 Feb Air Samples, 

1986o 1 -- --

1986 Jun Air Samples, 

1986p 4 -- --

1986 Jul Air Samples, 

1986q 14 -- --

1986 Nov Air Samples, 

1986r 3 -- --

1986 Dec Air Samples, 

1986s 9 -- --

1987 Jan Air Samples, 

1987b 15 1 --

1987 Feb Air Samples, 

1987c 22 -- --

1987 Mar Air Samples, 

1987d 8 -- --

1987 May Air Samples, 

1987e 6 -- --

1987 Jul Air Samples, 

1987f 3 -- --

1987 Aug Air Samples, 

1987g 13 -- --

1987 Sep Air Samples, 

1987h 7 -- --

1987 Dec Air Samples, 

1987i 11 -- --

1988 Jan Air Samples, 

1988h 18 -- --

1988 Feb Air Samples, 

1988i 15 -- --

1988 Apr Air Samples, 

1988j 5 -- --

1988 Nov Air Samples, 

1988k; 1988l 5 -- --

1989 Jun Air Samples, 

1989 1 -- --

TOTAL N/A N/A 526 46 6

NOTE: Documents captured by NIOSH include results from only portions of years 1980 and 1983-1989, as 

shown in the table.

* The number of days in the specified month for which captured air-monitoring data indicate that samples were 

collected. Procedures required that sample filters be changed and analyzed daily, except over weekends and 

holidays.

** The number of days for which the specified fraction of the uranium Radiological Control Guide (RCG) of 70 

dpm/m3 alpha was met or exceeded.

*** Notations on the air results indicate that respirators were worn.
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with the ABRWH or its Working Groups or Subcommittees. Draft, preliminary, interim, and White Paper documents are not final NIOSH or 

ABRWH (or their technical support and review contractors) positions unless specifically marked as such. This document represents preliminary 

positions taken on technical issues prepared by NIOSH or its contractor. NOTICE: This report has been reviewed to identify and redact any 

information that is protected by the Privacy Act 5 USC §552a and has been cleared for distribution. 

Table 7: Experiment Periods, Contamination Incident Dates, and Associated Records of Airborne 

Contamination Results

Experimental Campaigns* Contamination Incident Dates** Airborne Monitoring Days Captured

None 11/25/1980 0

02/1978 – 09/1981 None 1

Summer 1982 – 12/20/1982 None 0

05/1983 – 09/1984 None 64

None 07/07/1984 – 07/20/1984 0

07/1985 – 08/1986 None 139

None 02/14/ 1987 & 03/13/1987 30

04/1986 – 10/1987 None 104

*Rothe, 2005, PDF pp. 389-393

**Rothe, 2005, PDF pp. 486-487, 498, 500-501

Building 886/875 results appear in the same reports as those for Building 865, which housed the 

Metal Research and Development Laboratory9

9 Each day’s results for Building 886 are labeled RR-1 through RR-7; those for Building 875 are labeled RR-8 

through RR-11; Building 865 results begin with a “UU-” designator.

 and which processed non-plutonium metals, 

including depleted uranium (DOE, 2011 , PDF p. 2). Reports are reviewed and initialed, and 

instructions on the report sheet specify that a copy is to be sent “TO RADIATION MONT 881” 

(to Radiation Monitoring, Building 881). 

Individual sample results are occasionally lined out in the reports, with or without explanation. 

There are multiple instances when results are lined out with the notation “W/C” or “Wrong 

Color.” The meaning of this notation is not clear, and these samples are not included in 

evaluating the results.

A total of 526 days of results were evaluated against the Radiological Control Guide (RCG) of 

70 dpm/m3

. One or more results from the seven samplers in Building 886 exceeded the RCG on 

six days; results between 10% and 100% of the RCG were recorded on 46 additional days. All 

results were below 10% of the RCG for the remaining 473 days. Respirators were worn on three

of the six days in which the RCG was exceeded (Air Samples, 1984e, PDF pp. 33, 36, 52). 

Results on the other three days in which the RCG was exceeded were: 522.02% RCG (Air 

Samples, 1984f, PDF p. 5); 111.72% RCG (Air Samples, 1985f, PDF p. 37); and 117.76% RCG

(Air Samples, 1985f, PDF p. 41). 

Results from Building 875, which housed the effluent air exhaust plenums from the Building 886 

experiment room, exceeded the RCG on only one occasion10

10 The sample designation was “RR-12” for the single result exceeding the RCG, without explanation of the 

location or purpose for the sample. A location in Building 875 is assumed.

 during the same 526 days; results 

between 10% and 100% of the RCG were recorded on two days.

 
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with the ABRWH or its Working Groups or Subcommittees. Draft, preliminary, interim, and White Paper documents are not final NIOSH or 

ABRWH (or their technical support and review contractors) positions unless specifically marked as such. This document represents preliminary 

positions taken on technical issues prepared by NIOSH or its contractor. NOTICE: This report has been reviewed to identify and redact any 

information that is protected by the Privacy Act 5 USC §552a and has been cleared for distribution. 

Conclusions About Workplace Air Monitoring at the CML

NIOSH concludes that a robust and well-defined workplace air-monitoring program for 

alpha-emitting radioisotopes was required by Rocky Flats plant procedures on continuing basis 

during the period 1980-1989. In all air-monitoring records captured by NIOSH, air particulate 

samples from the CML, Building 886, and its air exhaust filtration plenums in Building 875, 

were routinely analyzed and reported along with those from the Metal Research and 

Development Laboratory, Building 865. Sampling results were evaluated for uranium alpha 

emissions and reported as a percentage of the radioactive concentration guide (RCG) airborne 

limit for uranium, 70 dpm/m3

. The analytical reports were reviewed and initialed, and were to be 

sent to Radiation Monitoring in Building 881. Excursions in excess of the RCG were 

uncommon, occurring in Building 886 on six of 526 days for which monitoring results were 

captured, and in Building 875 only once. 

A bounding value for activity concentrations in breathing air can be calculated as the weighted 

average CC̅of air results using recorded values for three results in excess of the RCG with no 

indication that respirators were worn, and by making the claimant-favorable assumptions that 

results were 70 dpm/m3 for 46 results between 10% and 100% of RCG and 7 dpm/m3 for the 

remaining 477 recorded samples, as shown in Equation 3 below: 

Results for the 526 days of monitoring data captured for the period 1980-1989 are assumed to be 

representative of uncaptured data for that period for the following reasons: 

• The same plant health physics procedures requiring air monitoring were in effect 

continuously (with revisions) for the entire period; 

• NIOSH assumes that daily samples were collected and analyzed in accordance with plant 

procedures over the entire period; and 

• The results are probably similar because CML operations were similar over the period. 

These data were available for routine review by health physics personnel, who also had access to 

information about operations for making personnel monitoring decisions. It is therefore unlikely 

that an unrecorded intake of alpha contaminated airborne particulates occurred during this period 

because of a lack of relevant air monitoring data.
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ABRWH (or their technical support and review contractors) positions unless specifically marked as such. This document represents preliminary 

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information that is protected by the Privacy Act 5 USC §552a and has been cleared for distribution.

ASSESSMENT OF UNMONITORED RADIATION DOSE AT THE CML

Radiation dose from intake of MFAP at CML could have occurred during clean up of numerous 

fuel spills, predominantly from enriched uranyl nitrate solution (UNH), or from inhalation of 

dried, resuspended contamination deposited on surfaces as the result of these spills (Rothe, 2005, 

PDF pp. 447-449, 452-462, 464-473, 486-487, 498, 500-501). Rocky Flats workers (including 

those assigned to the CML) with the potential for receiving intakes of plutonium, americium, or 

uranium were monitored by periodic urinalysis and body counts (NIOSH, 2006, PDF p. 30). 

However, NIOSH has found no indication that confirmatory bioassays were performed for 

employees involved in clean up of any of the accidental UNH spills. Fission and activation 

products, which decay primarily by beta/gamma emission, are not likely in any case to have been 

detected by bioassay intended to detect alpha particles emitted by uranium or transuranic 

radionuclides.

Maximum MFAP internal doses to CML workers were estimated by modeling a representative 

UNH experiment and calculating the MFAP inventory based on the historical record of CML 

experiments with UNH, and on the average thermal power and duration of CML UNH criticality 

experiments reported by CML researchers. Intakes of resuspended UNH contamination with the 

same MFAP-to-uranium atom ratio as the fuel were estimated from the weighted average of air 

monitoring results in the experimental and materials storage areas of the CML. Doses were 

calculated by applying ICRP 68 (ICRP, 1995) dose conversion factors for three solubility 

categories of dosimetrically significant radionuclides, using the method described in ORAUTOTIB-0054 (2015).

Calculation of Fission and Activation Product Content of UNH Solution

A series of experiments was performed at the CML during the mid-1970s to determine the 

critical height of UNH in suspended, cylindrical tanks. Experiments were performed using 

different uranium concentrations, different tanks, and with and without neutron reflectors. The 

same uranium enrichment was used in all experiments; only the concentration was varied.

NIOSH chose one of the unreflected suspended tank experiments from the mid-1970s to 

represent UNH experiments performed over the CML’s history. There were ten unreflected 

suspended tank experiments in all. Two of them used a stainless-steel tank and the remainder 

used an aluminum tank.

NIOSH selected one of the experiments that used the stainless-steel tank so that the calculated 

fission and activation product content would include iron activation products. The tank had an 

inside diameter of 27.92 cm and an inside height of 41.6 cm. For the selected experiment, the 

tank was filled with UNH with a uranium enrichment of 93.172 weight percent U-235 at a

concentration of 145.68 g/L. The other unreflected stainless steel tank experiment used UNH 

with the same enrichment, but at a concentration of 346.73 g/L. The lower concentration was 

selected because it represents the middle of the range of concentrations used across the ten 

unreflected tank experiments. The critical solution height for the selected experiment was found 
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ABRWH (or their technical support and review contractors) positions unless specifically marked as such. This document represents preliminary 

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information that is protected by the Privacy Act 5 USC §552a and has been cleared for distribution. 

to be 31.20 cm. The selected experiment is documented as Case Number 1 in “Minimally 

Reflected Cylinders of Highly Enriched Solutions of Uranyl Nitrate,” HEU-SOL-THERM-001, 

from the International Handbook of Evaluated Criticality Safety Benchmark Experiments

(Palmer, 2004). 

The fission and activation product composition for the selected experiment was calculated using 

SCALE. SCALE is a modular system of computer codes for nuclear- and radiological 

engineering-related analyses from the Oak Ridge National Laboratory. SCALE’s TRITON 

module was used to develop a case-specific cross-section library for the selected suspended tank 

experiment. The case-specific library was then used by the ORIGEN-S code to determine the 

time-dependent fission and activation product content of the UNH and the stainless-steel tank 

over the CML’s history. 

The TRITON module performs neutron transport and depletion calculations for irradiated 

nuclear fuels. Neutron transport is performed using either discrete ordinates methods (via the 

NEWT code) or via Monte Carlo calculations (using the KENO –V or KENO-VI codes). 

TRITON’s capabilities include creating case-specific cross-section library files that can be used 

by other elements of the SCALE code system, notably the ORIGEN-S code. Among the 

numerous capabilities of ORIGEN-S is calculation of the inventory of activation products, 

actinides, and fission products in a composition as a function of time and burnup (as applicable). 

TRITON was used to model the stainless-steel tank and UNH solution geometry of the selected 

suspended tank experiment as a system of cylinders representing the tank bottom, tank walls, and 

the UNH volume. The tank and solution heights were truncated to the critical solution height of 

31.20 cm. Neutron transport was performed via the KENO-VI Monte Carlo code. 

The case-specific cross-section library created by TRITON was subsequently used in an 

ORIGEN-S calculation to determine the fission and activation product content of the solution, 

and the tank volume at the end of the tenth campaign and thereafter. This reflects a modeling 

assumption that the same tank and solution were used for all 778 experiments involving HEU 

solutions performed at the CML over its operating history. The composition used for the 

ORIGEN-S calculations was a homogenized mixture of the HEU solution, the 304 stainless steel 

tank walls and bottom, and the associated impurities. The composition was irradiated and 

decayed in the ORIGEN-S case using the history provided in Table 7 above. The first campaign 

was represented by a 38-hour irradiation followed by 32 days of decay, and so on, through the 

final 19-hour irradiation representing the tenth campaign. 
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ABRWH (or their technical support and review contractors) positions unless specifically marked as such. This document represents preliminary 

positions taken on technical issues prepared by NIOSH or its contractor. NOTICE: This report has been reviewed to identify and redact any 

information that is protected by the Privacy Act 5 USC §552a and has been cleared for distribution. 

Each experiment was assumed to have lasted for 70.5 minutes and to have produced an average 

thermal power11

11 The power in this context is an average value representing a given number of fissions over a given interval of 

time because nuclear criticality experiments are not steady-state and are not typically considered in terms of a power 

level. An average power was used for modeling purposes to account for any change in the composition of the fissile 

solution (i.e., for depletion effects) and to compute the ingrowth of fission products as the solution was used. In 

reality, most of the fissions would have occurred near the end of a given experiment when the system was

alternately placed in slightly subcritical and slightly supercritical states.

 of 6.7 mW (the average power and duration reported to ERDA in 1977) (Tuck, 

1977, PDF p. 6).12

12 The 6.7 mW value is used, rather than the more precise 3.6 mW for 70.5 minutes estimated by CML staff 

after the communication with ERDA. There is no indication why 6.7 mW was reported instead of a lower value of 

5.4 mW initially calculated by CML staff before refining their estimate. The higher value is adopted as more 

favorable for claimants.

The ORIGEN-S calculation produced time-dependent fission and activation product inventories, 

with radioactive decay between each of the ten campaigns, for the HEU solution defined in the 

selected benchmark case, and using the timeline for all uranium solution experiments conducted 

over the CML’s history. Radioactive decay corrections were also applied at the end of the final 

experiment, using a decay period of 180 days from those specified in ORAUT-OTIB 0054 

(2015). 

Calculation of Inhalation Intakes and Committed Organ Doses 

Solution spills resulting in surface (floor) contamination occurred throughout CML’s operating 

history, as shown in Table 8. Most spills occurred in the late 1960s, but there were also several 

spills during the 1980s. NIOSH performed organ dose calculations after applying a decay 

interval of only 180 days to the ending fission and activation product inventory for the HEU 

solution. This is favorable to the claimant, given that the calculation of the fission and activation 

product inventory represents the entire operating history of the facility (i.e., includes the period 

after the last major spill).

 
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ABRWH (or their technical support and review contractors) positions unless specifically marked as such. This document represents preliminary 

positions taken on technical issues prepared by NIOSH or its contractor. NOTICE: This report has been reviewed to identify and redact any 

information that is protected by the Privacy Act 5 USC §552a and has been cleared for distribution. 

Table 8: High-enriched Uranium Solution Spills Over the CML’s History

Date Volume 

(L)

Uranium 

Mass (kg) Contaminated Area PDF Page(s)

7/2/1965 0.5 0.225 22 m2 (Floor) 447

7/14/1965 --- --- 10 m2 (Floor) 448-449

7/22/1965 --- --- Small amount (Floor) 449

11/30/1967 --- 9 (Inside large duct, filter housing, vent line)* 452-462

2/16/1968 --- 1.14** (Floor and cable trenches) 464-467

5/11/1968 0.06 --- (Workman’s knee) 467

5/9/1969 150. 1 16.1 20 m2(Mixing Room floor)*** 467-473

11/25/1980 7 2.66 (Assembly Room hood) 486-487

7/7-20/1984 --- --- (Walk-in hood)  498

2/14/1987 --- --- (Personnel, facility and fixtures)** *** 500-501

3/13/1987 --- --- (Personnel, facility and fixtures)***** 501

Source: Rothe, 2005

*Although contamination was confined to ducts and a filter housing, clean up of this incident resulted in a 

blowback of dried salts, resulting in facial contamination of a staff member. 

**Two conflicting accounts refer to this value as either the solution mass or uranium mass.

***Standing HEU solution covering the floor was cleaned up by a staff member using a critically safe vacuum 

and wearing plastic booties and a half-face respirator. 

 Potentially-contaminated workmen repairing a leaking flange were required to evacuate when a criticality 

alarm was triggered by an electrician.

** ***Two essentially identical events resulted in personnel, fixture, and facility contamination by resuspended 

high-enriched uranyl nitrate salts accidentally knocked from the surface of a large reactivity shim.

ORAUT-OTIB-0054 (2015) describes a method used to reduce the large number of fission and 

activation product isotopes in an ORIGEN-S result to a set of 36 dosimetrically significant 

nuclides. The same 36 nuclides were considered in the inhalation intake and committed organ 

dose calculations for the CML. Intakes were computed using the 180-day activity values for the 

HEU solution shown in Table 9, corrected for the average airborne concentration level, and 

assuming an intake period of 4,000 hours (two working years) at a breathing rate of 1.2 m3/h. 

The two-year intake period was selected to be consistent with ORAUT-OTIB-0054 (2015), 

which provides a basis for assigning internal dose from unknown inhalation of fission and 

activation product mixtures. Radioactive decay of the isotopic mixture over the two-year intake 

period was not considered, which is favorable to the claimant. 
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ABRWH (or their technical support and review contractors) positions unless specifically marked as such. This document represents preliminary 

positions taken on technical issues prepared by NIOSH or its contractor. NOTICE: This report has been reviewed to identify and redact any 

information that is protected by the Privacy Act 5 USC §552a and has been cleared for distribution. 

Table 9: Accumulated Activities of Dosimetrically Significant Fission and 

Activation Products in UNH Fuel at the CML

Isotope Activity (Bq)

Mn-54 1.800E+02

Fe-55 1.664E+03

Co-58 1.359E+02

Co-60 3.077E+00

Sr-89 1.287E+04

Sr-90 2.506E+04

Y-90 2.507E+04

Y-91 1.984E+04

Zr-95 2.734E+04

Nb-95 5.025E+04

Mo-99 5.206E-14

Ru-103 4.732E+03

Ru-106 3.889E+03

Cd-113m 1.658E+00

Cd-115m 9.955E-01

Sb-125 3.236E+02

Te-129m 1.051E+02

Te-132 3.937E-11

I-131 8.936E-02

I-132 4.055E-11

Cs-134 7.004E-02

Cs-136 4.318E-02

Cs-137 2.567E+04

Ba-140 3.719E+01

La-140 4.281E+01

Ce-141 5.509E+03

Ce-144 5.417E+04

Pr-143 6.690E+01

Pr-144 5.417E+04

Nd-147 3.204E+00

Pm-147 2.171E+04

Pm-148m 0.000E+00

Sm-151 6.867E+02
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ABRWH (or their technical support and review contractors) positions unless specifically marked as such. This document represents preliminary 

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information that is protected by the Privacy Act 5 USC §552a and has been cleared for distribution. 

Isotope Activity (Bq)

Eu-154 1.691E-03

Eu-155 3.225E+02

Ta-182 0.000E+00

NOTE: Inventory in the UNH fuel was accumulated without radioactive decay until the final 

experiment, after which the activities of individual isotopes were decay-corrected for 180 days to 

yield these values.

Evaluation of air monitoring data in the earlier section, CML Workplace Air Monitoring,

determined that a weighted-average concentration of 13.5 dpm/m3 for airborne alpha activity was 

favorable to the claimant. Inhalation intakes were computed by assuming that the airborne alpha 

activity consisted entirely of HEU having a specific activity of 70 μCi/g. Therefore, the airborne 

mass concentration was 8.7 × 10-8 g/m3

 and the total intake over the 4,000-hour period was 

4.2×10-4 g. 

The UNH fuel, as modeled, contained 2,782.8 g HEU. The ratio of total inhalation intake over 

the two-year intake period and HEU solution mass gives an intake fraction f = 1.5 × 10-7. The 

inhalation intake I for the 36 dosimetrically-significant nuclides for the two-year period was then 

determined as the product of the ORIGEN-S result for each nuclide after 180 days of decay 

(activity A) and the intake fraction, i.e., I = f A. 

Committed organ doses were computed (see Equation 4 below) as the sum of products of the 

180-day intakes, computed for each of the 36 dosimetrically significant nuclides, and their

corresponding inhalation dose conversion factors (DCF), einh(50), from ICRP 68 (ICRP, 1995). 

The DCFs were partitioned into three solubility categories to account for different absorption 

types, using the method described in ORAUT-OTIB-0054 (2015). Three committed dose values 

were therefore computed for each of the 25 individual organs defined in ICRP 68 (ICRP, 1995), 

corresponding to soluble, moderately soluble, and insoluble materials. Table 10 shows the 

maximum committed organ dose values for the three solubility categories. 

Table 10: Maximum Committed Organ Doses from Inhalation of Airborne 

Fission and Activation Products at CML

Solubility H(50) (Sv) Tissue

Soluble (Type F) 2.5 × 10-9 Bone surface

Moderately Soluble (Type M) 2.3 × 10-9 Bone surface

Insoluble (Type S) 2.4 × 10-9 Lung
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ABRWH (or their technical support and review contractors) positions unless specifically marked as such. This document represents preliminary 

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information that is protected by the Privacy Act 5 USC §552a and has been cleared for distribution. 

These values are greatly reduced over previously calculated committed doses of 3.7×10-7 Sv 

(soluble, bone surface), 4.0×10-7 Sv (moderately soluble, lung), and 6.1×10-7 Sv (insoluble, 

lung). Earlier committed doses were calculated using airborne concentrations derived by 

applying a resuspension factor to the claimant-favorable DOE limit on removable surface 

contamination. The orders-of-magnitude difference in the two sets of dose values is due to a 

prior miscalculation in converting from 100 cm2 to m2. The corrected calculation uses a 

resuspension factor of 1.5×10-4 m-1 and 2,000 dpm/100 cm2 removable alpha. This calculation is 

applied over the entire 220 m2

 footprint of the CML experimental and material storage areas, 

giving an estimated air concentration of 30 dpm/m3. The claimant-favorable 13.5 dpm/m3

weighted average alpha air concentration from routine monitoring results, used in calculations 

for this analysis, further reduces the dose estimate from that obtained previously. 

Previous estimates of fission and activation product inventories also assumed a thermal power of 

10 mW for 60 minutes, compared with the lesser (but still claimant-favorable) value of 6.7 mW 

for 70.5 minutes used in this analysis to describe conditions for the typical CML criticality 

experiment.

SUMMARY AND CONCLUSIONS

Re-evaluation of unmonitored personnel dose using recently captured air monitoring data and 

reactor performance estimates for the Rocky Flats Plant Critical Mass Laboratory has resulted in 

estimates on the order of a few nanosieverts for maximum organ doses due to inhalation of 

resuspended contamination containing mixed fission and activation products. These estimates 

are over two orders of magnitude lower than previous estimates. The greatest contributor to the 

large reduction in estimated doses is a correction in the calculation of a conversion factor. 

Lesser contributors to the reduction are lower estimates of reactor power in a typical criticality 

experiment at CML, and a lower value for respirable alpha air concentrations based on routine 

air monitoring results.

NIOSH computer-modeled criticality experiments using estimates of typical reactor power and 

corresponding neutron flux documented by CML research staff and alpha air concentrations from 

routine monitoring results. Based on its modeling, NIOSH concludes that no significant 

personnel dose to Rocky Flats workers or contractors resulted from the generation of fission or 

activation products in the uranyl nitrate fuel or resuspended contamination from fuel spills as a 

result of criticality experiments conducted at CML over its lifetime.
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information that is protected by the Privacy Act 5 USC §552a and has been cleared for distribution. 

REFERENCES

Air Samples, 1980, Air Samples Buildings 883 and 886; Rocky Flats Plant; December 1980; 

SRDB Ref ID: 159122 

Air Samples, 1981a, Air Samples Building 875; Rocky Flats Plant; January 1981; SRDB Ref ID: 

159124 

Air Samples, 1981b, Air Samples Building 875; Rocky Flats Plant; December 1981; SRDB Ref 

ID: 159125 

Air Samples, 1982a, Air Samples Building 875; Rocky Flats Plant; January 1982; SRDB Ref ID: 

159126 

Air Samples, 1982b, Air Samples Building 875; Rocky Flats Plant; February 1982; SRDB Ref 

ID: 159127 

Air Samples, 1982c, Air Samples Building 875; Rocky Flats Plant; March 1982; SRDB Ref ID: 

159128 

Air Samples, 1982d, Air Samples Building 875; Rocky Flats Plant; April 1982; SRDB Ref ID: 

159129 

Air Samples, 1982e, Air Samples Building 875; Rocky Flats Plant; May 1982; SRDB Ref ID: 

159130 

Air Samples, 1982f, Air Samples Building 875; Rocky Flats Plant; June 1982; SRDB Ref ID: 

159131 

Air Samples, 1982g, Air Samples Building 875; Rocky Flats Plant; July 1982; SRDB Ref ID: 

159132 

Air Samples, 1982h, Air Samples Building 875; Rocky Flats Plant; August 1982; SRDB Ref ID: 

159133 

Air Samples, 1982i, Air Samples Building 875; Rocky Flats Plant; November 1982; SRDB Ref 

ID: 159135 

Air Samples, 1983a, Air Samples Building 875; Rocky Flats Plant; May 1983; SRDB Ref ID: 

159137 

Air Samples, 1983b, Air Samples Building 875; Rocky Flats Plant; June 1983; SRDB Ref ID: 

159138 
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ABRWH (or their technical support and review contractors) positions unless specifically marked as such. This document represents preliminary 

positions taken on technical issues prepared by NIOSH or its contractor. NOTICE: This report has been reviewed to identify and redact any 

information that is protected by the Privacy Act 5 USC §552a and has been cleared for distribution. 

Air Samples, 1983c, Air Samples Buildings 865 and 886; Rocky Flats Plant; May 1983; SRDB 

Ref ID: 159136 

Air Samples, 1983d, Air Samples Buildings 865 and 886; Rocky Flats Plant; September 1983; 

SRDB Ref ID: 159154 

Air Samples, 1983e, Air Filter Direct Alpha Counting System Buildings 865 and 886; Rocky 

Flats Plant; October 1983; SRDB Ref ID: 156834 

Air Samples, 1984a, Air Samples Building 875; Rocky Flats Plant; January 1984; SRDB Ref ID: 

159155 

Air Samples, 1984b, Air Samples Building 875; Rocky Flats Plant; April 1984; SRDB Ref ID: 

159158 

Air Samples, 1984c, Air Samples Building 875; Rocky Flats Plant; May 1984; SRDB Ref ID: 

159160 

Air Samples, 1984d, Air Samples Building 875; Rocky Flats Plant; June 1984; SRDB Ref ID: 

159162 

Air Samples, 1984e, Air Sample Direct Alpha Counting System Buildings 865 and 886; Rocky 

Flats Plant; July 1984; SRDB Ref ID: 156857 

Air Samples, 1984f, Air Sample Direct Alpha Counting System Buildings 865 and 886; Rocky 

Flats Plant; August 1984; SRDB Ref ID: 156858 

Air Samples, 1984g, Air Sample Direct Alpha Counting System Buildings 865 and 886; Rocky 

Flats Plant; September 1984; SRDB Ref ID: 156859 

Air Samples, 1984h, Air Sample Direct Alpha Counting System Buildings 865 and 886; Rocky 

Flats Plant; October 1984; SRDB Ref ID: 156860 

Air Samples, 1984i, Air Sample Direct Alpha Counting System Buildings 865 and 886; Rocky 

Flats Plant; November 1984; SRDB Ref ID: 156861 

Air Samples, 1984j, Air Sample Direct Alpha Counting System Buildings 865 and 886; Rocky 

Flats Plant; December 1984; SRDB Ref ID: 156862 

Air Samples, 1985a, Air Filter Direct Alpha Counting System Building 875; Rocky Flats Plant; 

January 1985; SRDB Ref ID: 156864 

Air Samples, 1985b, Air Filter Direct Alpha Counting System Building 875; Rocky Flats Plant; 

February 1985; SRDB Ref ID: 156875 
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ABRWH (or their technical support and review contractors) positions unless specifically marked as such. This document represents preliminary 

positions taken on technical issues prepared by NIOSH or its contractor. NOTICE: This report has been reviewed to identify and redact any 

information that is protected by the Privacy Act 5 USC §552a and has been cleared for distribution. 

Air Samples, 1985c, Air Filter Direct Alpha Counting System Building 875; Rocky Flats Plant; 

March 1985; SRDB Ref ID: 156877 

Air Samples, 1985d, Air Filter Direct Alpha Counting System Building 875; Rocky Flats Plant; 

April 1985; SRDB Ref ID: 156879 

Air Samples, 1985e, Air Samples Building 875; Rocky Flats Plant; December 1985; SRDB Ref 

ID: 159139 

Air Samples, 1985f, Air Sample Direct Alpha Counting System Buildings 865 and 886; Rocky 

Flats Plant; January 1985; SRDB Ref ID: 156863 

Air Samples, 1985g, Air Sample Direct Alpha Counting System Buildings 865 and 886; Rocky 

Flats Plant; February 1985; SRDB Ref ID: 156867 

Air Samples, 1985h, Air Sample Direct Alpha Counting System Buildings 865 and 886; Rocky 

Flats Plant; Mar 1985; SRDB Ref ID: 156876 

Air Samples, 1985i, Air Sample Direct Alpha Counting System Buildings 865 and 886; Rocky 

Flats Plant; April 1985; SRDB Ref ID: 156878 

Air Samples, 1985j, Air Samples Direct Alpha Counting System Building 865 and 886; Rocky 

Flats Plant; May 1985; SRDB Ref ID: 160323 

Air Samples, 1985k, Air Samples Buildings 865 and 886; Rocky Flats Plant; June 1985; SRDB 

Ref ID: 159142 

Air Samples, 1985l, Air Samples Buildings 865 and 886; Rocky Flats Plant; July 1985; SRDB 

Ref ID: 159143 

Air Samples, 1985m, Air Samples Buildings 865 and 886; Rocky Flats Plant; August 1985; 

SRDB Ref ID: 159144 

Air Samples, 1985n, Air Samples Buildings 865 and 886; Rocky Flats Plant; September 1985; 

SRDB Ref ID: 159145 

Air Samples, 1985o, Air Samples Buildings 865 and 886; Rocky Flats Plant; October 1985; 

SRDB Ref ID: 159146 

Air Samples, 1985p, Air Samples Buildings 865 and 886; Rocky Flats Plant; November 1985; 

SRDB Ref ID: 159147 

Air Samples, 1985q, Air Samples Buildings 865 and 886; Rocky Flats Plant; December 1985; 

SRDB Ref ID: 159148 
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ABRWH (or their technical support and review contractors) positions unless specifically marked as such. This document represents preliminary 

positions taken on technical issues prepared by NIOSH or its contractor. NOTICE: This report has been reviewed to identify and redact any 

information that is protected by the Privacy Act 5 USC §552a and has been cleared for distribution. 

Air Samples, 1986a, Air Filter Direct Alpha Counting System Building 875; Rocky Flats Plant; 

January 1986; SRDB Ref ID: 156881 

Air Samples, 1986b, Air Filter Direct Alpha Counting System Building 875; Rocky Flats Plant; 

February 1986; SRDB Ref ID: 156883 

Air Samples, 1986c, Air Filter Direct Alpha Counting System Building 875; Rocky Flats Plant; 

March 1986; SRDB Ref ID: 156884 

Air Samples, 1986d, Air Filter Direct Alpha Counting System Building 875; Rocky Flats Plant; 

April 1986; SRDB Ref ID: 156885 

Air Samples, 1986e, Air Filter Direct Alpha Counting System Building 875; Rocky Flats Plant; 

May 1986; SRDB Ref ID: 156886; 

Air Samples, 1986f, Air Filter Direct Alpha Counting System Building 875; Rocky Flats Plant; 

June 1986; SRDB Ref ID: 156888 

Air Samples, 1986g, Air Filter Direct Alpha Counting System Building 875; Rocky Flats Plant; 

July 1986; SRDB Ref ID: 156890 

Air Samples, 1986h, Air Filter Direct Alpha Counting System Building 875; Rocky Flats Plant; 

August 1986; SRDB Ref ID: 156891 

Air Samples, 1986i, Air Filter Direct Alpha Counting System Building 875; Rocky Flats Plant; 

September 1986; SRDB Ref ID: 156892 

Air Samples, 1986j, Air Sample Travel Logs Building 886 Using Ludium 12; Rocky Flats Plant; 

September 24-26, 1986; SRDB Ref ID: 156893 

Air Samples, 1986k, Air Filter Direct Alpha Counting System Building 875; Rocky Flats Plant; 

October 1986; SRDB Ref ID: 156894 

Air Samples, 1986l, Air Filter Direct Alpha Counting System Building 875; Rocky Flats Plant; 

November 1986; SRDB Ref ID: 156895 

Air Samples, 1986m, Air Filter Direct Alpha Counting System Building 875; Rocky Flats Plant; 

December 1986; SRDB Ref ID: 156897 

Air Samples, 1986n, Air Sample Direct Alpha Counting System Building 865 and 886; Rocky 

Flats Plant; January 1986; SRDB Ref ID: 156880 

Air Samples, 1986o, Air Sample Direct Alpha Counting System Building 865 and 886; Rocky 

Flats Plant; February 1986; SRDB Ref ID: 156882 
White Paper: Reassessment of Internal Dose from Sources at the RFP CML

Page 41 of 51

This is a working document prepared by NIOSH’s Division of Compensation Analysis and Support (DCAS) or its contractor for use in discussions 

with the ABRWH or its Working Groups or Subcommittees. Draft, preliminary, interim, and White Paper documents are not final NIOSH or 

ABRWH (or their technical support and review contractors) positions unless specifically marked as such. This document represents preliminary 

positions taken on technical issues prepared by NIOSH or its contractor. NOTICE: This report has been reviewed to identify and redact any 

information that is protected by the Privacy Act 5 USC §552a and has been cleared for distribution. 

Air Samples, 1986p, Air Sample Direct Alpha Counting System Building 865 and 886; Rocky 

Flats Plant; June 1986; SRDB Ref ID: 156887 

Air Samples, 1986q, Air Sample Direct Alpha Counting System Building 865 and 886; Rocky 

Flats Plant; July 1986; SRDB Ref ID: 156889 

Air Samples, 1986r, Air Samples Direct Alpha Counting System Building 865 and 886; Rocky 

Flats Plant; November 1986; SRDB Ref ID: 160325 

Air Samples, 1986s, Air Sample Direct Alpha Counting System Building 865 and 886; Rocky 

Flats Plant; December 1986; SRDB Ref ID: 156896 

Air Samples, 1987a, Air Samples Building 875; Rocky Flats Plant; January 1987; SRDB Ref ID: 

159140 

Air Samples, 1987b, Air Samples Buildings 865 and 886; Rocky Flats Plant; January 1987; 

SRDB Ref ID: 159149 

Air Samples, 1987c, Air Samples Building 886; Rocky Flats Plant; February 1987; SRDB Ref 

ID: 159123 

Air Samples, 1987d, Air Samples Building 886; Rocky Flats Plant; Mar 1987; SRDB Ref ID: 

159150 

Air Samples, 1987e, Air Samples Building 886; Rocky Flats Plant; May 1987; SRDB Ref ID: 

159151 

Air Samples, 1987f, Air Samples Building 886; Rocky Flats Plant; July 1987; SRDB Ref ID: 

159152 

Air Samples, 1987g, Air Samples Building 886; Rocky Flats Plant; August 1987; SRDB Ref ID: 

159153 

Air Samples, 1987h, Air Samples Building 886; Rocky Flats Plant; September 1987; SRDB Ref 

ID: 159156 

Air Samples, 1987i, Air Samples Building 886; Rocky Flats Plant; December 1987; SRDB Ref 

ID: 159157 

Air Samples, 1988a, Effluent Air Survey Building 875; Rocky Flats Plant; May 1988; SRDB Ref 

ID: 156898 

Air Samples, 1988b, Air Filter Direct Alpha Counting System Building 875; Rocky Flats Plant; 

January 1988; SRDB Ref ID: 156899 
White Paper: Reassessment of Internal Dose from Sources at the RFP CML

Page 42 of 51

This is a working document prepared by NIOSH’s Division of Compensation Analysis and Support (DCAS) or its contractor for use in discussions 

with the ABRWH or its Working Groups or Subcommittees. Draft, preliminary, interim, and White Paper documents are not final NIOSH or 

ABRWH (or their technical support and review contractors) positions unless specifically marked as such. This document represents preliminary 

positions taken on technical issues prepared by NIOSH or its contractor. NOTICE: This report has been reviewed to identify and redact any 

information that is protected by the Privacy Act 5 USC §552a and has been cleared for distribution. 

Air Samples, 1988c, Air Filter Direct Alpha Counting System Building 875; Rocky Flats Plant; 

February 1988; SRDB Ref ID: 156901 

Air Samples, 1988d, Air Filter Direct Alpha Counting System Building 875; Rocky Flats Plant; 

March 1988; SRDB Ref ID: 156903 

Air Samples, 1988e, Air Filter Direct Alpha Counting System Building 875; Rocky Flats Plant; 

April 1988; SRDB Ref ID: 156905 

Air Samples, 1988f, Air Filter Direct Alpha Counting System Building 875; Rocky Flats Plant; 

May 1988; SRDB Ref ID: 156906 

Air Samples, 1988g, Air Samples Building 875; Rocky Flats Plant; September 1988; SRDB Ref 

ID: 159141 

Air Samples, 1988h, Air Sample Direct Alpha Counting System Building 886; Rocky Flats Plant; 

January 1988; SRDB Ref ID: 156900 

Air Samples, 1988i, Air Sample Direct Alpha Counting System Building 886; Rocky Flats Plant; 

February 1988; SRDB Ref ID: 156902 

Air Samples, 1988j, Air Sample Direct Alpha Counting System Building 886; Rocky Flats Plant; 

April 1988; SRDB Ref ID: 156907 

Air Samples, 1988k, Air Samples Building 886; Rocky Flats Plant; November 1988; SRDB Ref 

ID: 159159 

Air Samples, 1988l, Air Samples Building 886; Rocky Flats Plant; November 1988; SRDB Ref 

ID: 159161 

Air Samples, 1989, Air Samples Building 886; Rocky Flats Plant; June 1989; SRDB Ref ID: 

159163 

Analysis Report; August 7, 1977; SRDB Ref ID: 157754 

Building 886, 2011, Rocky Flats Plant, Critical Mass Laboratory, HAER No. CO-83-A (Rocky 

Flats Plant, Building 886); U.S. Department of Energy, Legacy Management, Historic American 

Engineering Record, Rocky Flats Site, Building 886; November 3, 2011; SRDB Ref ID: 104452 

Calibrations, 1987, Building 886 and 875 Airhead Calibrations; Rocky Flats Plant; JanuaryApril, July-August, and October 1987; SRDB Ref ID: 152336 

Calibrations, 1988, Building 886 and 875 Airhead Calibrations; Rocky Flats Plant; January, 

March, and August-September 1988; SRDB Ref ID: 152339 
White Paper: Reassessment of Internal Dose from Sources at the RFP CML

Page 43 of 51

This is a working document prepared by NIOSH’s Division of Compensation Analysis and Support (DCAS) or its contractor for use in discussions 

with the ABRWH or its Working Groups or Subcommittees. Draft, preliminary, interim, and White Paper documents are not final NIOSH or 

ABRWH (or their technical support and review contractors) positions unless specifically marked as such. This document represents preliminary 

positions taken on technical issues prepared by NIOSH or its contractor. NOTICE: This report has been reviewed to identify and redact any 

information that is protected by the Privacy Act 5 USC §552a and has been cleared for distribution. 

Calibrations, 1989a, Building 886 and 875 Airhead and SAAM Calibrations; Rocky Flats Plant; 

January, February, and April 1989; SRDB Ref ID: 152512 

Calibrations, 1989b, Building 886 and 875 Airhead Calibrations; Rocky Flats Plant; MayAugust 1989; SRDB Ref ID: 152513 

DOE (U.S. Department of Energy), 2011, Historic American Engineering Record Rocky Flats 

Site Building 865 Metal Research and Development Laboratory; U.S. Department of Energy 

Legacy Management; November 28, 2011; SRDB Ref ID: 105198 

Experiment Run 2-8-170; Undated; SRDB Ref ID: 157750 

Experiments at RFNSF, 1965 to April 19, 1977, 1977a; April 19, 1977; SRDB Ref ID: 157752 

Experiments at RFNSF, 1965 to April 19, 1977, 1977b; April 19, 1977; SRDB Ref ID: 157757 

ICRP (International Commission on Radiological Protection), 1995, Dose Coefficients for 

Intakes of Radionuclides by Workers; ICRP Publication 68; Oxford, New York: Pergamon Press; 

1995; SRDB Ref ID: 22731 

Introducing Nuclear Reactor Theory and Experiments 2-8-169 and 2-8-170; May 4, 1977; 

SRDB Ref ID: 157748 

Log, 1967, Console Log #3, January 3 and 6, 1967; Rocky Flats Plant; January 3 and 6, 1967; 

SRDB Ref ID: 157753 

Log, 1976, Building 886 Fission Due to Experiment Gamma Level Oy Shells, Rocky Flats Plant; 

April 29, 1976; SRDB Ref ID: 157751 

NIOSH, 2006, SEC Petition Evaluation Report for Petition SEC-00030, Rocky Flats Plant; 

National Institute for Occupational Safety and Health (NIOSH); April 7, 2006; SRDB Ref ID: 

132776 

ORAUT-OTIB-0004, Estimating the Maximum Plausible Dose to Workers at Atomic Weapons 

Employer Facilities, Revision 3, PC-2; ORAU Team Dose Reconstruction Project for NIOSH; 

December 6, 2006; SRDB Ref ID: 36191 

ORAUT-OTIB-0054, Fission and Activation Product Assignment for Internal Dose-Related 

Gross Beta and Gross Gamma Analyses, Revision 3; ORAU Team Dose Reconstruction Project 

for NIOSH; February 6, 2015; SRDB Ref ID: 315479 

Palmer, 2004, “Minimally Reflected Cylinders of Highly Enriched Solutions of Uranyl Nitrate;” 

NEA/NSC/DOC(95)03/II, Volume II; HEU-SOL-THERM-001, from the International 

Handbook of Evaluated Criticality Safety Benchmark Experiments; Brian Palmer, Idaho 

Chemical Processing Plant; September 30, 2004; SRDB Ref ID: 142464 
White Paper: Reassessment of Internal Dose from Sources at the RFP CML

Page 44 of 51

This is a working document prepared by NIOSH’s Division of Compensation Analysis and Support (DCAS) or its contractor for use in discussions 

with the ABRWH or its Working Groups or Subcommittees. Draft, preliminary, interim, and White Paper documents are not final NIOSH or 

ABRWH (or their technical support and review contractors) positions unless specifically marked as such. This document represents preliminary 

positions taken on technical issues prepared by NIOSH or its contractor. NOTICE: This report has been reviewed to identify and redact any 

information that is protected by the Privacy Act 5 USC §552a and has been cleared for distribution. 

Personal Communication, 2015a, Personal Communication with former Rocky Flats Worker; 

telephone interview ORAU Team, ABRWH, DCAS, and Advocates; October 13, 2015; SRDB 

Ref ID: 150603 

Personal Communication, 2015b, Personal Communication with former Rocky Flats Worker; 

telephone interview ORAU Team and DCAS; October 13, 2015; SRDB Ref ID: 150602 

Radiation Monitoring Technical Support Programs, 1976, Supplied Breathing Air Procedures 

Radiation Monitoring and Routine Air Sampling Procedures; HS-RM4.1; Rocky Flats Plant; 

March 1976; SRDB Ref ID: 24235 

Radiological Surveys, 1980a, Daily Control Point, Dry Swipe, Gamma and Neutron and 

Radiation Monitoring Surveys Building 886; Rocky Flats Plant; January 1980; SRDB Ref ID: 

156807 

Radiological Surveys, 1980b, Daily Control Point, Dry Swipe and Gamma Neutron Surveys 

Buildings 875 and 886; Rocky Flats Plant; February 1980; SRDB Ref ID: 156808 

Radiological Surveys, 1980c, Daily Control Point, Dry Swipe, Gamma Neutron and Radiation 

Monitoring Survey Results Buildings 886 and 875; Rocky Flats Plant; March 1980; SRDB Ref 

ID: 156809 

Radiological Surveys, 1980d, Daily Control Point, Dry Swipe and Gamma Neutron Surveys 

Buildings 875 and 886; Rocky Flats Plant; April 1980; SRDB Ref ID: 156810 

Radiological Surveys, 1980e, Daily Control Point and Dry Swipe Surveys Buildings 875 and 

886; Rocky Flats Plant; May 1980; SRDB Ref ID: 156811 

Radiological Surveys, 1980f, Daily Control Point and Dry Swipe Surveys Buildings 875 and 

886; Rocky Flats Plant; June 1980; SRDB Ref ID: 156812 

Radiological Surveys, 1980g, Daily Control Point, Dry Swipe, Gamma Neutron and Radiation 

Monitoring Surveys Buildings 875 and 886; Rocky Flats Plant; July 1980; SRDB Ref ID: 156813 

Radiological Surveys, 1980h, Daily Control Point, Dry Swipe and Gamma Neutron Surveys 

Building 886; Rocky Flats Plant; August 1980; SRDB Ref ID: 156814 

Radiological Surveys, 1980i, Daily Control Point, Dry Swipe and Gamma Neutron Surveys 

Building 886; Rocky Flats Plant; September 1980; SRDB Ref ID: 156815 

Radiological Surveys, 1980j, Daily Control Point, Dry Swipe and Gamma Neutron Surveys 

Building 886; Rocky Flats Plant; October 1980; SRDB Ref ID: 156816 

Radiological Surveys, 1980k, Daily Control Point, Dry Swipe and Gamma Neutron Surveys 

Building 886; Rocky Flats Plant; November 1980; SRDB Ref ID: 156817 
White Paper: Reassessment of Internal Dose from Sources at the RFP CML

Page 45 of 51

This is a working document prepared by NIOSH’s Division of Compensation Analysis and Support (DCAS) or its contractor for use in discussions 

with the ABRWH or its Working Groups or Subcommittees. Draft, preliminary, interim, and White Paper documents are not final NIOSH or 

ABRWH (or their technical support and review contractors) positions unless specifically marked as such. This document represents preliminary 

positions taken on technical issues prepared by NIOSH or its contractor. NOTICE: This report has been reviewed to identify and redact any 

information that is protected by the Privacy Act 5 USC §552a and has been cleared for distribution. 

Radiological Surveys, 1980l, Daily Control Point, Dry Swipe and Gamma Neutron Surveys 

Building 875 and 886; Rocky Flats Plant; December 1980; SRDB Ref ID: 156818 

Radiological Surveys, 1981a, Building 886 Radiological Surveys; Rocky Flats Plant; October

1981; SRDB Ref ID: 152233 

Radiological Surveys, 1981b, Building 886 Radiological Surveys; Rocky Flats Plant; November

1981; SRDB Ref ID: 152245 

Radiological Surveys, 1981c, Building 886 Radiological Surveys; Rocky Flats Plant; January 

1981; SRDB Ref ID: 152247 

Radiological Surveys, 1981d, Building 886 Radiological Surveys; Rocky Flats Plant; December 

1981; SRDB Ref ID: 152248 

Radiological Surveys, 1981e, Building 886 Radiological Surveys; Rocky Flats Plant; February 

1981; SRDB Ref ID: 152250 

Radiological Surveys, 1981f, Building 886 Radiological Surveys; Rocky Flats Plant; March 

1981; SRDB Ref ID: 152251 

Radiological Surveys, 1981g, Building 886 Radiological Surveys; Rocky Flats Plant; April 1981; 

SRDB Ref ID: 152253 

Radiological Surveys, 1981h, Building 886 Radiological Surveys; Rocky Flats Plant; May 1981; 

SRDB Ref ID: 152255 

Radiological Surveys, 1981i, Building 886 Radiological Surveys; Rocky Flats Plant; June 1981; 

SRDB Ref ID: 152257 

Radiological Surveys, 1981j, Building 886 Radiological Surveys; Rocky Flats Plant; July 1981; 

SRDB Ref ID: 152259 

Radiological Surveys, 1981k, Building 886 Radiological Surveys; Rocky Flats Plant; August 

1981; SRDB Ref ID: 152261 

Radiological Surveys, 1981l, Building 875 Radiological Surveys; Rocky Flats Plant; September 

1981; SRDB Ref ID: 152262 

Radiological Surveys, 1981m, Building 886 Radiological Surveys; Rocky Flats Plant; September

1981; SRDB Ref ID: 152263 

Radiological Surveys, 1982a, Building 886 Radiological Surveys; Rocky Flats Plant; July 1982; 

SRDB Ref ID: 152264 
White Paper: Reassessment of Internal Dose from Sources at the RFP CML

Page 46 of 51

This is a working document prepared by NIOSH’s Division of Compensation Analysis and Support (DCAS) or its contractor for use in discussions 

with the ABRWH or its Working Groups or Subcommittees. Draft, preliminary, interim, and White Paper documents are not final NIOSH or 

ABRWH (or their technical support and review contractors) positions unless specifically marked as such. This document represents preliminary 

positions taken on technical issues prepared by NIOSH or its contractor. NOTICE: This report has been reviewed to identify and redact any 

information that is protected by the Privacy Act 5 USC §552a and has been cleared for distribution. 

Radiological Surveys, 1982b, Building 886 Radiological Surveys; Rocky Flats Plant; August 

1982; SRDB Ref ID: 152266 

Radiological Surveys, 1982c, Building 886 Radiological Surveys; Rocky Flats Plant; September 

1982; SRDB Ref ID: 152267 

Radiological Surveys, 1982d, Building 886 Radiological Surveys; Rocky Flats Plant; October 

1982; SRDB Ref ID: 152268 

Radiological Surveys, 1982e, Building 886 Radiological Surveys; Rocky Flats Plant; November 

1982; SRDB Ref ID: 152269 

Radiological Surveys, 1982f, Building 886 Radiological Surveys; Rocky Flats Plant; December 

1982; SRDB Ref ID: 152270 

Radiological Surveys, 1982g, Building 886 Radiological Surveys; Rocky Flats Plant; February 

1982; SRDB Ref ID: 152276 

Radiological Surveys, 1982h, Building 886 Radiological Surveys; Rocky Flats Plant; March 

1982; SRDB Ref ID: 152281 

Radiological Surveys, 1982i, Building 886 Radiological Surveys; Rocky Flats Plant; April 1982; 

SRDB Ref ID: 152288 

Radiological Surveys, 1982j, Building 886 Radiological Surveys; Rocky Flats Plant; May 1982; 

SRDB Ref ID: 152292 

Radiological Surveys, 1982k, Building 886 Radiological Surveys with Survey of Sr-90 Sulfate 

Containers; Rocky Flats Plant; June 1982; SRDB Ref ID: 152295 

Radiological Surveys, 1982l, Floor, Daily Control and Gamma Neutron Dry Swipe Surveys 

Building 886; Rocky Flats Plant; January 1982; SRDB Ref ID: 156820 

Radiological Surveys, 1983a, Building 886 Radiological Surveys; Rocky Flats Plant; November 

1983; SRDB Ref ID: 152271 

Radiological Surveys, 1983b, Building 886 Radiological Surveys; Rocky Flats Plant; December 

1983; SRDB Ref ID: 152272 

Radiological Surveys, 1983c, Daily and Monthly Control Point, Controlled Access Area and 

Offices Dry Swipe Results Building 886; Rocky Flats Plant; January 1983; SRDB Ref ID: 

156822 
White Paper: Reassessment of Internal Dose from Sources at the RFP CML

Page 47 of 51

This is a working document prepared by NIOSH’s Division of Compensation Analysis and Support (DCAS) or its contractor for use in discussions 

with the ABRWH or its Working Groups or Subcommittees. Draft, preliminary, interim, and White Paper documents are not final NIOSH or 

ABRWH (or their technical support and review contractors) positions unless specifically marked as such. This document represents preliminary 

positions taken on technical issues prepared by NIOSH or its contractor. NOTICE: This report has been reviewed to identify and redact any 

information that is protected by the Privacy Act 5 USC §552a and has been cleared for distribution. 

Radiological Surveys, 1983d, Daily, Weekly and Monthly Control Point, Controlled Access Area 

and Offices Dry Swipe Results Building 886; Rocky Flats Plant; February 1983; SRDB Ref ID: 

156824 

Radiological Surveys, 1983e, Daily, Weekly and Monthly Control Point, Controlled Access Area 

and Offices Dry Swipe Results Building 886; Rocky Flats Plant; March 1983; SRDB Ref ID: 

156825 

Radiological Surveys, 1983f, Daily and Monthly Control Point, Controlled Access Area and 

Offices Dry Swipe Results Building 886; Rocky Flats Plant; April 1983; SRDB Ref ID: 156826 

Radiological Surveys, 1983g, Contamination Surveys Building 886; Rocky Flats Plant; June 

1983; SRDB Ref ID: 159203 

Radiological Surveys, 1983h, Contamination Surveys Building 886; Rocky Flats Plant; July 

1983; SRDB Ref ID: 159204 

Radiological Surveys, 1983i, Contamination Surveys Building 886; Rocky Flats Plant; August 

1983; SRDB Ref ID: 159205 

Radiological Surveys, 1983j, Contamination Surveys Building 886; Rocky Flats Plant; 

September 1983; SRDB Ref ID: 159207 

Radiological Surveys, 1983k, Contamination Surveys Building 886; Rocky Flats Plant; October 

1983; SRDB Ref ID: 159210 

Radiological Surveys, 1984a, Building 875 Radiological Surveys; Rocky Flats Plant; October, 

1984; SRDB Ref ID: 152229 

Radiological Surveys, 1984b, Building 886 Radiological Surveys; Rocky Flats Plant; January 

1984; SRDB Ref ID: 152273 

Radiological Surveys, 1984c, Building 886 Radiological Surveys; Rocky Flats Plant; February 

1984; SRDB Ref ID: 152277 

Radiological Surveys, 1984d, Building 886 Radiological Surveys; Rocky Flats Plant; March 

1984; SRDB Ref ID: 152280 

Radiological Surveys, 1984e, Building 886 Radiological Surveys; Rocky Flats Plant; April 1984; 

SRDB Ref ID: 152284 

Radiological Surveys, 1984f, Building 886 Radiological Surveys; Rocky Flats Plant; May 1984; 

SRDB Ref ID: 152286 
White Paper: Reassessment of Internal Dose from Sources at the RFP CML

Page 48 of 51

This is a working document prepared by NIOSH’s Division of Compensation Analysis and Support (DCAS) or its contractor for use in discussions 

with the ABRWH or its Working Groups or Subcommittees. Draft, preliminary, interim, and White Paper documents are not final NIOSH or 

ABRWH (or their technical support and review contractors) positions unless specifically marked as such. This document represents preliminary 

positions taken on technical issues prepared by NIOSH or its contractor. NOTICE: This report has been reviewed to identify and redact any 

information that is protected by the Privacy Act 5 USC §552a and has been cleared for distribution. 

Radiological Surveys, 1984g, Building 886 Radiological Surveys; Rocky Flats Plant; June 1984; 

SRDB Ref ID: 152289 

Radiological Surveys, 1984h, Building 886 Radiological Surveys; Rocky Flats Plant; July 1984; 

SRDB Ref ID: 152293 

Radiological Surveys, 1984i, Building 886 Radiological Surveys; Rocky Flats Plant; August 

1984; SRDB Ref ID: 152299 

Radiological Surveys, 1984j, Building 886 Radiological Surveys; Rocky Flats Plant; September 

1984; SRDB Ref ID: 152301 

Radiological Surveys, 1984k, Building 886 Radiological Surveys; Rocky Flats Plant; October 

1984; SRDB Ref ID: 152302 

Radiological Surveys, 1984l, Building 886 Radiological Surveys; Rocky Flats Plant; November 

1984; SRDB Ref ID: 152304 

Radiological Surveys, 1984m, Building 886 Radiological Surveys; Rocky Flats Plant; December 

1984; SRDB Ref ID: 152318 

Radiological Surveys, 1985a, Building 886 Radiological Surveys; Rocky Flats Plant; January 

1985; SRDB Ref ID: 152323 

Radiological Surveys, 1985b, Building 886 Radiological Surveys; Rocky Flats Plant; February 

1985; SRDB Ref ID: 152325 

Radiological Surveys, 1985c, Building 886 Radiological Surveys; Rocky Flats Plant; March 

1985; SRDB Ref ID: 152331 

Radiological Surveys, 1985d, Building 886 Radiological Surveys; Rocky Flats Plant; April 1985; 

SRDB Ref ID: 152335 

Radiological Surveys, 1985e, Building 886 Radiological Surveys; Rocky Flats Plant; May 1985; 

SRDB Ref ID: 152340 

Radiological Surveys, 1985f, Building 875 Radiological Surveys; Rocky Flats Plant; June 1985; 

SRDB Ref ID: 152342 

Radiological Surveys, 1985g, Building 886 Radiological Surveys; Rocky Flats Plant; June 1985; 

SRDB Ref ID: 152343 

Radiological Surveys, 1985h, Building 886 Radiological Surveys; Rocky Flats Plant; July 1985; 

SRDB Ref ID: 152344 
White Paper: Reassessment of Internal Dose from Sources at the RFP CML

Page 49 of 51

This is a working document prepared by NIOSH’s Division of Compensation Analysis and Support (DCAS) or its contractor for use in discussions 

with the ABRWH or its Working Groups or Subcommittees. Draft, preliminary, interim, and White Paper documents are not final NIOSH or 

ABRWH (or their technical support and review contractors) positions unless specifically marked as such. This document represents preliminary 

positions taken on technical issues prepared by NIOSH or its contractor. NOTICE: This report has been reviewed to identify and redact any 

information that is protected by the Privacy Act 5 USC §552a and has been cleared for distribution. 

Radiological Surveys, 1985i, Building 886 Radiological Surveys; Rocky Flats Plant; August 

1985; SRDB Ref ID: 152346 

Radiological Surveys, 1985j, Radiation Monitoring Surveys, Weekly, Monthly, Control Points 

and Dry Swipe Results Buildings 886, 875, and 880; Rocky Flats Plant; September 1985; SRDB 

Ref ID: 157168 

Radiological Surveys, 1985k, Radiation Monitoring Surveys, Weekly, Monthly, Control Points 

and Dry Swipe Results Buildings 886, 875, and 880; Rocky Flats Plant; October 1985; SRDB 

Ref ID: 157170 

Radiological Surveys, 1985l, Radiation Monitoring Surveys, Weekly, Monthly, Control Points, 

Gamma Neutron Surveys and Dry Wipe Results Buildings 886, 875, and 880; Rocky Flats Plant; 

November 1985; SRDB Ref ID: 157172 

Radiological Surveys, 1985m, Radiation Monitoring Surveys, Weekly, Monthly, Control Points 

and Dry Swipe Results Buildings 886, 875, and 880; Rocky Flats Plant; December 1985; SRDB 

Ref ID: 157173 

Radiological Surveys, 1986a, Building 886 Radiological Surveys; Rocky Flats Plant; April 1986; 

SRDB Ref ID: 152298 

Radiological Surveys, 1986b, Building 886 Radiological Surveys; Rocky Flats Plant; May 1986; 

SRDB Ref ID: 152306 

Radiological Surveys, 1986c, Building 886 Radiological Surveys; Rocky Flats Plant; June 1986; 

SRDB Ref ID: 152308 

Radiological Surveys, 1986d, Building 886 Radiological Surveys; Rocky Flats Plant; July 1986; 

SRDB Ref ID: 152310 

Radiological Surveys, 1986e, Building 886 Radiological Surveys; Rocky Flats Plant; August 

1986; SRDB Ref ID: 152312 

Radiological Surveys, 1986f, Building 886 Radiological Surveys; Rocky Flats Plant; September 

1986; SRDB Ref ID: 152314 

Radiological Surveys, 1986g, Building 886 Radiological Surveys; Rocky Flats Plant; SeptemberOctober 1986; SRDB Ref ID: 152317 

Radiological Surveys, 1986h, Building 886 Radiological Surveys; Rocky Flats Plant; November 

1986; SRDB Ref ID: 152320 

Radiological Surveys, 1986i, Building 886 Radiological Surveys; Rocky Flats Plant; December 

1986; SRDB Ref ID: 152326 
White Paper: Reassessment of Internal Dose from Sources at the RFP CML

Page 50 of 51

This is a working document prepared by NIOSH’s Division of Compensation Analysis and Support (DCAS) or its contractor for use in discussions 

with the ABRWH or its Working Groups or Subcommittees. Draft, preliminary, interim, and White Paper documents are not final NIOSH or 

ABRWH (or their technical support and review contractors) positions unless specifically marked as such. This document represents preliminary 

positions taken on technical issues prepared by NIOSH or its contractor. NOTICE: This report has been reviewed to identify and redact any 

information that is protected by the Privacy Act 5 USC §552a and has been cleared for distribution. 

Radiological Surveys, 1986j, Building 886 Radiological Surveys; Rocky Flats Plant; January 

1986; SRDB Ref ID: 152351 

Radiological Surveys, 1986k, Building 886 Radiological Surveys; Rocky Flats Plant; February 

1986; SRDB Ref ID: 152353 

Radiological Surveys, 1986l, Building 886 Radiological Surveys; Rocky Flats Plant; March 

1986; SRDB Ref ID: 152355 

Radiological Surveys, 1987a, Building 886 Cold Area Radiological Surveys; Rocky Flats Plant; 

1987; SRDB Ref ID: 152327 

Radiological Surveys, 1987b, Building 886 Control Point Radiological Surveys; Rocky Flats 

Plant; 1987; SRDB Ref ID: 152328 

Radiological Surveys, 1987c, Building 886 Dry Alpha Sweep Radiological Surveys; Rocky Flats 

Plant; 1987; SRDB Ref ID: 152329 

Radiological Surveys, 1987d, Building 886 Neutron-Photon Radiological Surveys; Rocky Flats 

Plant; 1987; SRDB Ref ID: 152330 

Radiological Surveys, 1987e, Building 886 Equipment Release Survey; Rocky Flats Plant; June 

1987; SRDB Ref ID: 152332 

Radiological Surveys, 1988a, Buildings 886 and 875 SAAM Alarm Log Sheets; Rocky Flats 

Plant; April-May 1988; SRDB Ref ID: 152341 

Radiological Surveys, 1988b, Building 886 Cold Area Radiological Surveys; Rocky Flats Plant; 

January-November 1988; SRDB Ref ID: 152348 

Radiological Surveys, 1988c, Building 886 Control Point Radiological Surveys; Rocky Flats 

Plant; January-November 1988; SRDB Ref ID: 152357 

Radiological Surveys, 1988d, Building 886 Dry Swipe Radiological Surveys; Rocky Flats Plant; 

January-October 1988; SRDB Ref ID: 152359 

Radiological Surveys, 1988e, Building 886 Neutron-Photon Radiological Surveys; Rocky Flats 

Plant; February and August-September 1988; SRDB Ref ID: 152360 

Radiological Surveys, 1989a, Building 886 Cold Area Radiological Surveys; Rocky Flats Plant; 

January-April 1989; SRDB Ref ID: 152362 

Radiological Surveys, 1989b, Building 886 Contamination Surveys; Rocky Flats Plant; JanuaryJune and October-December 1989; SRDB Ref ID: 152500 
White Paper: Reassessment of Internal Dose from Sources at the RFP CML

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This is a working document prepared by NIOSH’s Division of Compensation Analysis and Support (DCAS) or its contractor for use in discussions 

with the ABRWH or its Working Groups or Subcommittees. Draft, preliminary, interim, and White Paper documents are not final NIOSH or 

ABRWH (or their technical support and review contractors) positions unless specifically marked as such. This document represents preliminary 

positions taken on technical issues prepared by NIOSH or its contractor. NOTICE: This report has been reviewed to identify and redact any 

information that is protected by the Privacy Act 5 USC §552a and has been cleared for distribution. 

Radiological Surveys, 1989c, Building 886 Control Point Radiological Surveys; Rocky Flats 

Plant; January October 1989; SRDB Ref ID: 152502 

Radiological Surveys, 1989d, Building 886 Neutron-Photon Radiological Surveys; Rocky Flats 

Plant; January-November 1989; SRDB Ref ID: 152507 

Radiological Surveys, 1989e, Building 886 Radiological Surveys; Rocky Flats Plant; MayNovember 1989; SRDB Ref ID: 152508 

Radiological Surveys, 1989f, Building 886 Office Area Surveys; Rocky Flats Plant; AprilOctober 1989; SRDB Ref ID: 152509 

Radiological Surveys, 1989g, Building 886 Smear Counter Performance Tests; Rocky Flats 

Plant; December 1989; SRDB Ref ID: 152511 

Radiological Surveys, 1990a, Building 886 Contamination Surveys; Rocky Flats Plant; JanuaryJune and October-December1990; SRDB Ref ID: 152515 

Radiological Surveys, 1990b, Building 886 Gamma and Neutron-Photon Surveys; Rocky Flats 

Plant; January-July and September-December 1990; SRDB Ref ID: 152516 

Rothe, 2005, A Technically Useful History of the Critical Mass Laboratory at Rocky Flats, LAUR-05-3247, R. E. Rothe; Los Alamos National Laboratory; May 2005; SRDB Ref ID: 21358 

Summary of All Nuclear Safety Experiments; April 19, 1977; SRDB Ref ID: 157749 

Tuck, 1977, Total Fissions Produced by Critical Approach Experiments Done at the Rocky Flats 

Plant Nuclear Facility from September 1965 to June 1, 1977; Grover Tuck, Rocky Flats Nuclear 

Safety Facility; June 3, 1977; SRDB Ref ID: 157872