Page 1 of 72

DOE Review Release 05/23/2022

Evaluation of Issues in the Use of General 

Area Air Sampling for Argonne National 

Laboratory-West Internal Dose Assessment

ORAUT-RPRT-0089 Rev. 00

Effective Date: 04/19/2022

Supersedes: None

Subject Expert(s): Brian P. Gleckler

Document Owner 

Approval: Signature on File

Pat M. McCloskey, Document Owner

Approval Date: 04/05/2022

Concurrence: Vickie S. Short Signature on File for

Kate Kimpan, Project Director

Concurrence Date: 04/05/2022

Approval: Signature on File

Timothy D. Taulbee, Associate Director for Science

Approval Date: 04/19/2022

FOR DOCUMENTS MARKED AS A TOTAL REWRITE, REVISION, OR PAGE CHANGE, REPLACE THE PRIOR 

REVISION AND DISCARD / DESTROY ALL COPIES OF THE PRIOR REVISION.

New Total Rewrite Revision Page Change
Document No. ORAUT-RPRT-0089 Revision No. 00 Effective Date: 04/19/2022 Page 2 of 72

PUBLICATION RECORD

EFFECTIVE 

DATE

REVISION 

NUMBER DESCRIPTION

04/19/2022 00 New document to evaluate issues raised regarding the use of 

general area air sampling for Argonne National Laboratory-West

internal dose assessments. Training required: As determined by the 

Objective Manger. Initiated by Daniel H. Stempfley and authored by 

Brian P. Gleckler.
Document No. ORAUT-RPRT-0089 Revision No. 00 Effective Date: 04/19/2022 Page 3 of 72

TABLE OF CONTENTS

SECTION TITLE PAGE

Acronyms and Abbreviations ............................................................................................................... 6

1.0 Purpose ................................................................................................................................... 8

2.0 Background and Terminology .................................................................................................. 8

2.1 Background .................................................................................................................. 8

2.2 Terminology................................................................................................................ 10

3.0 Summary of the Issues .......................................................................................................... 11

4.0 Overview of Actinide Air Sampling, 1958 to 1976................................................................... 12

4.1 Source Term and Air Sampling at EBR-I Complex, 1958 to 1975 ............................... 15

4.2 Actinide Air Sampling at EBR-II Complex, 1963 to 1976............................................. 15

4.2.1 Thorium Areas, 1963 to 1967....................................................................... 16

4.2.2 Uranium Areas, 1967 to 1976 ...................................................................... 20

4.2.2.1 Basis for Exposure Period Change............................................ 20

4.2.2.2 Review of Air Sampling Data..................................................... 21

4.2.3 Plutonium Areas, 1970 to 1975 .................................................................... 23

4.2.3.1 Zero Power Plutonium Reactor.................................................. 23

4.2.3.2 Fuel Cycle Facility Areas ........................................................... 23

5.0 Response to Issue 1 .............................................................................................................. 24

5.1 Part 1 of Issue 1 ......................................................................................................... 24

5.2 Part 2 of Issue 1 ......................................................................................................... 25

5.2.1 EBR-II Complex Air Sampling Data, 1963 through 1974 .............................. 26

5.2.2 EBR-II Complex Air Sampling Data, 1975 to 1976 ....................................... 27

6.0 Response to Issue 2 .............................................................................................................. 28

6.1 Applicability of Studies from Other Sites ..................................................................... 28

6.1.1 Key Parameters in ORAUT-RPRT-0097 ...................................................... 28

6.1.2 Air Concentration Levels .............................................................................. 30

6.2 Studies Referenced by SCA-TR-2016-SEC009.......................................................... 31

6.2.1 Caldwell, Potter, and Schnell Study ............................................................. 31

6.2.1.1 Description of the NUMEC Apollo Site....................................... 31

6.2.1.2 Description of the NUMEC Parks Township Site ....................... 32

6.2.1.3 Applicability of NUMEC Sites to ANL-W Actinide-Only 

Areas......................................................................................... 33

6.2.2 Brunskill and Holt Study ............................................................................... 35

6.2.2.1 Description of Windscale Works Site......................................... 35

6.2.2.2 Description of Springfields Works Site....................................... 35

6.2.2.3 Applicability of United Kingdom Facilities to ANL-W 

Actinide-Only Areas................................................................... 36

6.2.3 Summary of Issues with Studies Referenced by SCA-TR-2016-

SEC009 ....................................................................................................... 37

6.3 Studies Evaluated by ORAU Team............................................................................. 38

6.4 Recommended Resolution to the Lack of Parity Issue................................................ 39

6.4.1 Justifying Use of ORAUT-RPRT-0097 for ANL-W Actinide-Only Areas ........ 39

6.4.2 Applicable Scenarios from ORAUT-RPRT-0097 .......................................... 40
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7.0 Unmonitored Actinide Intakes ................................................................................................ 41

7.1 General Approaches Used for Intake Calculations...................................................... 42

7.1.1 Intakes Based on Bounding Air Concentrations ........................................... 43

7.1.2 Intakes Based on Air Sample Data............................................................... 44

7.1.2.1 Intakes Based on Adjusted Air Sample Data ............................. 45

7.1.2.2 Intakes Based on Unadjusted Air Sample Data ......................... 45

7.2 Summary of Air Concentrations for Intake Calculations .............................................. 46

7.3 Calculated Unmonitored Actinide Intakes ................................................................... 47

7.3.1 EBR-I Complex Intakes................................................................................ 47

7.3.2 EBR-II Complex Intakes............................................................................... 47

7.4 Application of Unmonitored Actinide Intakes............................................................... 47

7.4.1 Overlapping Periods of Exposure................................................................. 48

7.4.2 Uncertainties................................................................................................ 48

8.0 Conclusions ........................................................................................................................... 48

8.1 Issues Raised in SCA-TR-2016-SEC009.................................................................... 48

8.2 Additional Details about Unmonitored Actinide Intake Calculations ............................ 50

8.3 Deviations from SEC-00224 Evaluation Report .......................................................... 50

References ........................................................................................................................................ 52

ATTACHMENT A STATISTICAL EVALUATION OF EBR-II COMPLEX AIR SAMPLE DATA...... 58

ATTACHMENT B FIGURES OF THE ANL-W ACTINIDE-ONLY AREAS .................................... 62

LIST OF TABLES

TABLE TITLE PAGE

6-1 ANL-W actinide-only area room sizes .................................................................................... 29

6-2 SNM possession limits for NUMEC Apollo Site ...................................................................... 32

6-3 NUMEC Apollo facility room sizes .......................................................................................... 34

6-4 NUMEC Parks Township facility room sizes........................................................................... 34

6-5 Summary of the combined BZ:GA ratio distributions .............................................................. 41

7-1 EBR-I Complex uranium area air concentrations.................................................................... 46

7-2 EBR-II Complex uranium area air concentrations................................................................... 46

7-3 EBR-II Complex thorium area air concentrations.................................................................... 46

7-4 EBR-II Complex plutonium area air concentrations ................................................................ 46

7-5 Unmonitored uranium intakes at the EBR-I Complex ............................................................. 47

7-6 Unmonitored uranium intakes at the EBR-II Complex ............................................................ 47

7-7 Unmonitored thorium intakes at the EBR-II Complex ............................................................. 47

7-8 Unmonitored plutonium intakes for the EBR-II Complex......................................................... 47

7-9 Unmonitored plutonium intakes for the ZPPR......................................................................... 47

LIST OF FIGURES

FIGURE TITLE PAGE

4-1 Example of radioactive decay on an FCF air sample.............................................................. 14

4-2 Radiation Survey Report for a thoria spill ............................................................................... 17

4-3 Air sample collected during the thoria spill ............................................................................. 18
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4-4 Air sample collected during cleanup of the thoria spill ............................................................ 19

4-5 Example of a post-April 1975 air sample record..................................................................... 22

6-1 BZ:GA ratio by BZ air concentration at the NUMEC plutonium laboratory .............................. 30

6-2 BZ:GA ratio by BZ air concentration at Windscale Works....................................................... 31

A-1 FCF Room 25 (thoria room) air data, August 1963 to November 1967................................... 60

A-2 FCF Cold-Line air data, August 1967 to March 1973.............................................................. 60

A-3 FCF Cold-Line air data, May 1975 to June 1976 .................................................................... 61

A-4 FCF RAS-TREAT sodium-loop air data, April 1970 to April 1973 ........................................... 61

B-1 Early FCF floorplan with dimensions ...................................................................................... 64

B-2 FCF floorplan with room numbers and added dimensions, before August 1967..................... 65

B-3 FCF late Cold-Line era floorplan ............................................................................................ 66

B-4 FCF ventilation flow diagram during Cold-Line era................................................................. 67

B-5 FCF Room 25 (thoria room) ................................................................................................... 68

B-6 FCF Room 22 (sodium-loop glovebox room).......................................................................... 68

B-7 FCF Room 20 (Cold-Line room)............................................................................................. 69

B-8 FCF Room 26 (Cold-Line room)............................................................................................. 69

B-9 ITF .........................................................................................................................................70

B-10 ZPR-III workroom floorplan with added dimensions................................................................ 71

B-11 ZPPR workroom and vault floorplan with added dimensions .................................................. 72
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ACRONYMS AND ABBREVIATIONS 

AC number of air changes 

ADU ammonium diuranate

AEC U.S. Atomic Energy Commission

ANL Argonne National Laboratory

ANL-W Argonne National Laboratory-West

BZ breathing zone

cfm cubic feet per minute

CFR Code of Federal Regulations

cm centimeter

d day

DCAS Division of Compensation Analysis and Support

DOE U.S. Department of Energy

dpm disintegrations per minute

EBR-I Experimental Breeder Reactor-I 

EBR-II Experimental Breeder Reactor-II

FASB Fuel Assembly and Storage Building

FCF Fuel Cycle Facility

ft foot

g gram

GA general area

GM geometric mean

GSD geometric standard deviation

HEU highly enriched uranium

hr hour

in. inch

INL Idaho National Laboratory

ITF Inspection and Testing Facility

kg kilogram

LEU low-enriched uranium

lpm liters per minute

m meter

mCi millicurie

MFP mixed fission product

min minute

MPC maximum permissible concentration

MPL maximum permissible limit

NBS National Bureau of Standards

NIOSH National Institute for Occupational Safety and Health

NRC U.S. Nuclear Regulatory Commission
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NUMEC Nuclear Materials and Equipment Corporation

ORAU Oak Ridge Associated Universities

PAS personal air sampler

pCi picocurie

RAS-TREAT Reactor Analysis and Safety Division-Transient Reactor Experiment and Test 

RCG radiation concentration guide

ROS regression on order statistic

SAS stationary air sampler

SEC Special Exposure Cohort

SNM special nuclear material

SRDB Ref ID Site Research Database Reference 

μCi microcurie Identification (number)

UKAERE United Kingdom Atomic Energy Research Establishment

wk week

yr year

ZPPR Zero Power Plutonium (later Physics) Reactor

ZPR-III Zero Power Reactor-III

μm micrometer
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1.0 PURPOSE

The National Institute for Occupational Safety and Health (NIOSH) evaluation report for Argonne 

National Laboratory-West (ANL-W) Special Exposure Cohort (SEC) Petition SEC-00224 (NIOSH 

2016) proposed using air sampling data to bound potential unmonitored exposures to thorium, 

uranium, and plutonium when mixed fission products (MFPs) were not part of the radioactive source 

term. The areas where these potential unmonitored actinide exposures occurred are generically 

referred to as the “actinide-only areas” throughout this report.

The primary purpose of this report is to address issues raised by SC&A and Salient about the 

unmonitored actinide internal dose approaches NIOSH proposed in the SEC-00224 evaluation report. 

Those issues are documented in SCA-TR-2016-SEC009, Review of Petition Evaluation Report for 

SEC-00224, Argonne National Laboratory-West Regarding the Use of General Area Air Sampling for 

Internal Dose Assessment (SC&A and Salient 2016). Those specific issues are summarized in 

Section 3.0 and are primarily addressed in Sections 5.0 and 6.0. 

A secondary purpose of this report is to provide additional details about the calculation of the 

unmonitored actinide intakes to be used for the ANL-W dose reconstructions. Those additional details 

are provided throughout Sections 2.2, 4.0, 5.0, 6.0, and 7.0. 

As part of the efforts described above, further evaluations of the ANL-W radiological source terms and 

Experimental Breeder Reactor-II (EBR-II) Complex air sample data were performed, which included 

reviewing records that were captured after the SEC-00224 evaluation report was issued. Those 

evaluations indicated that some of the dates for the potential exposure periods needed to be adjusted 

and that the air concentrations in several areas were not bounded by 10% of the maximum 

permissible concentration (MPC) as indicated in the SEC-00224 evaluation report. As a result, 

another purpose of this report is to present the basis for revising some of the unmonitored actinide 

intake and internal dose approaches proposed in the SEC-00224 evaluation report. The bases for 

those deviations from the SEC-00224 evaluation report are provided in Sections 4.1, 4.2.2.1, and 

4.2.3, and the deviations are summarized in Section 8.3.

This report does not address unmonitored actinide exposures in areas where MFPs were also 

present. ORAUT-TKBS-0007-5, Idaho National Laboratory and Argonne National Laboratory-West – 

Occupational Internal Dose, provides an approach for addressing those exposures (ORAUT 2010a).

2.0 BACKGROUND AND TERMINOLOGY

2.1 BACKGROUND

ANL-W facilities occupied two main areas on the Idaho National Laboratory (INL) site, the 

Experimental Breeder Reactor-I (EBR-I) Complex and the EBR-II Complex. Since its beginning in 

1949 until February 2005, ANL-W was operated by the University of Chicago under the U.S. 

Department of Energy (DOE) (and its predecessors) Chicago Operations Office. In February 2005, 

DOE merged ANL-W with INL under the Idaho Operations Office; the operating facilities were 

collectively renamed as the Materials and Fuels Complex.

In addition, ANL-W was originally known as the Idaho Division (ID) of the Argonne National 

Laboratory (ANL) and was also referred to as the “Idaho Site” in ANL documents. In this report, the 

site is referred to as ANL-W. 

The SEC-00224 evaluation determined that workers at the EBR-I Complex could have been exposed 

to dispersible forms of uranium without MFPs being present. It also determined that workers at the 

EBR-II Complex could have been exposed to dispersible forms of thorium, uranium, and plutonium 
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without the MFPs being present. Because there are inadequate actinide bioassay data to estimate 

the potential intakes of thorium, uranium, and plutonium when the MFPs are not present, the SEC00224 evaluation report recommended estimating the potential unmonitored actinide intakes for the 

possible periods of exposure based on the gross alpha radioactivity air sampling results for those 

areas. The following are brief excerpts of what was originally recommended in Sections 7.2.1.2 and 

7.2.2.2 of the SEC-00224 evaluation report (NIOSH 2016). 

EBR-I Complex Uranium

Because there are inadequate uranium bioassay data to estimate the potential intakes 

of uranium without mixed fission products at the EBR-I Complex after 1957, NIOSH will 

bound the potential intakes of uranium by assigning uranium intakes based on 10% of 

the maximum permissible concentration (MPC) that ANL-W was using. A review of the 

available air monitoring data indicated that the airborne alpha radioactivity at the EBR-I 

Complex was typically below 10% of the MPC. Because it cannot be determined when 

and where the workers within the EBR-I Complex worked, these uranium intakes will be 

assessed for all workers any time they were within the EBR-I Complex after 1957. 

EBR-II Complex Uranium

The potential for uranium exposures within ZPPR [Zero Power Plutonium (later Physics) 

Reactor] was minimal because encapsulated and coated fuel sources were used (SRDB 

138139). ZPPR Health Physics used air monitoring data and/or contamination smear 

data to control potential uranium exposure at this reactor during normal operations. 

Loading and unloading procedures were carefully monitored to detect any loose 

contamination on fuel plates. Exposures to uranium without mixed fission products can 

be bound using 10% of MPC (maximum permissible concentration) from available air 

monitoring data. 

Because of FCF [Fuel Cycle Facility] Hot-Line start-up activities, FCF machine shop 

activities, and Cold-Line fuel production at the FCF, ITF [Inspection and Testing Facility], 

and FASB [Fuel Assembly and Storage Building], more-than-incidental intakes of 

depleted and enriched uranium without mixed fission products present could have 

occurred from August 1967 to as late as 1994. For the period of August 1967 through 

June 1983, no uranium bioassay data could be found for ANL-W workers. Beginning in 

July 1983, there is a significant increase in uranium bioassays for ANL-W workers, and 

the bioassay data can be used to estimate the intakes and doses from these uranium 

exposures. 

and

Because it cannot be determined when and where the workers within the EBR-II 

Complex worked, these uranium intakes will be assessed for all workers any time they 

were within the EBR-II Complex during the period of August 1967–June 1983. For 

periods that workers were only assigned EBR-I Complex dosimeters, no additional 

uranium intakes will be assessed using this approach. 

EBR-II Complex Thorium

Because there are inadequate thorium bioassay data to estimate the potential intakes of 

thorium without mixed fission products during the period of August 1963–November 

1967, NIOSH will bound the potential intakes of thorium by assigning thorium intakes 

based on 10% of the maximum permissible concentration (MPC) that ANL-W was using. 

A review of the available air monitoring data indicated that the airborne alpha 

radioactivity in FCF Room 25 was typically below 10% of the MPC. Because it cannot 

be determined when and where the workers within the EBR-II Complex worked, these 
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thorium intakes will be assessed for all workers any time they were within the EBR-II 

Complex during the period of August 1963–November 1967. 

EBR-II Complex Plutonium

The potential for plutonium exposures within ZPPR was minimal because encapsulated 

and coated fuel sources were used (SRDB 138139). ZPPR Health Physics used air 

monitoring data and/or contamination smear data to control potential plutonium exposure 

at this reactor during normal operations. Loading and unloading procedures were 

carefully monitored to detect any loose contamination on fuel plates. Exposures to 

plutonium without mixed fission products can be bound using 10% of MPC (maximum 

permissible concentration) from available air monitoring data. 

Because there isn’t adequate plutonium bioassay data to estimate the potential intakes 

of plutonium without mixed fission products during the period of April 1970–December 

1972, NIOSH will estimate those intakes using the gross alpha radioactivity air sampling 

results that were collected during RAS-TREAT [Reactor Analysis and Safety DivisionTransient Reactor Experiment and Test] sodium-loop experiment work in FCF Rooms 

22 and 27. In the instances where the air samples were counted for alpha radioactivity 

more than once due to the presence of short-lived alpha-emitting radionuclides, the latest 

result for gross alpha radioactivity will be used for this approach, since isotopes of 

plutonium are long-lived alpha-emitting radionuclides. Because it cannot be determined 

when and where the workers within the EBR-II Complex worked, these plutonium intakes 

will be assessed for all workers any time they were within the EBR-II Complex during the 

period of April 1970–December 1972. 

Additional details about the ANL-W site, its operations, and radiological control practices can be found 

in the SEC-00224 evaluation report (NIOSH 2016). Because that information is extensive, it was not 

included in this document for brevity. The information in the SEC-00224 evaluation report provides 

details about the locations, processes, and radioactive source terms for the actinide-only areas at 

ANL-W (NIOSH 2016). In the subsequent sections of this report, additional details are provided: 

1) regarding how the SEC-000224 evaluation report recommendations above were interpreted

(Sections 2.1 and 4.0), 2) how the unmonitored actinide intakes were calculated (Sections 2.2, 4.0, 

5.0, 6.0, and 7.0), and 3) how they will be applied (Section 7.4). 

2.2 TERMINOLOGY

The following paragraphs explain some of the terminology in this report.

The term “significant” means “practically significant” as opposed to “statistically significant.” In other 

words, in this report “significant” means that the difference (effect size) is large enough to influence 

how a parameter is used or treated in practice. 

The term “workroom” refers to a work location having a single discernable airspace. 

For this report, breathing zone (BZ) is defined as the volume of air surrounding a worker where the 

characteristics of the aerosol are close to being identical to the air inhaled by the worker (NRC 1984). 

Thus, the results of a BZ air sample are considered to be close to being identical to the air inhaled by 

the worker. 

The issues being addressed in this report primarily involve the relationship between the general area 

(GA) and BZ air concentrations. GA air concentrations are generally measured using stationary air 

samplers (SASs). Whereas, BZ air concentrations can be measured using SASs, portable air 

samplers, or personal air samplers (PASs). As indicated in NUREG/CR-4033, PAS can be 
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considered an adequate BZ sampling method in almost all situations (NRC 1984, p. 17). However, 

SASs and portable air samplers are only considered to be BZ samplers when their sampling heads 

are within the BZ (NRC 1984, p. 25). A “lapel air sampler” is a specific type of PAS that has the 

sampling head attached to the lapel of the worker’s clothing. 

The relationships between BZ and GA air concentrations are typically discussed in terms of BZ-to-GA 

air concentration ratios, which are abbreviated as BZ:GA ratios throughout this report. 

During the SEC-00224 evaluation years of 1951 to 1979, the applicable air concentration limit values 

were expressed in terms of maximum permissible limit (MPL), MPC, and radiation concentration guide 

(RCG). Based on the available Air Sample Data sheets for ANL-W, uses of the term MPL were found 

as early as April 1959 and as late as December 28, 1959. After that, the term MPC was used on 

those sheets for the period of December 29, 1959 through early 1973. By October 1973, the site was 

using the term RCG on the Air Sample Data sheets and continued using that term beyond the 

evaluation years (i.e., beyond 1979). Based on the available monthly and weekly ANL-W Health 

Physics reports, the term MPC was used in those reports before February 1966. From at least 

January 1967 through the end of the evaluation period, only the term RCG was used in the Health 

Physics reports, even during the periods when MPC was used in the Air Sample Data sheets. 

Because those terms were mostly interchangeable to ANL-W and because the values of whatever 

concentration limit was being applied were typically documented in the Air Sample Data sheets, only 

the terms MPC or air concentration limit are used in this report.

3.0 SUMMARY OF THE ISSUES

The following is a summary of the two main issues raised in Section 4.0 of SCA-TR-2016-SEC009 

(SC&A and Salient 2016). Because the issues were not numbered in SCA-TR-2016-SEC009, they 

have been numbered below to assist with referring to and addressing a specific issue. Excerpts from 

Section 4.0 of SCA-TR-2016-SEC009 are in the italicized text below: 

Issue 1

SC&A’s review of FCF air data, typical daily operations, and assessment of NIOSH’s 

proposed use of GA air sampling data identified two issues of concern. The first concern 

centers around the use of GA air samplers with a low airflow. Low airflow rates required 

sampling times of up to 4 days, which necessarily correspond to periods when no work 

was performed (and no workers were present). Thus, it is reasonable to conclude that 

air concentrations during non-working hours differed significantly from air concentrations 

that would have exposed workers during normal facility operations that were likely limited 

to an 8-hour shift Monday through Friday during many periods.

Issue 2

A second and more serious concern regarding the use of GA air sampling data is the 

generic lack of parity between air concentrations measured by GA air samplers and lapel 

air samplers worn by the individual workers. Study data, including those of two nuclear 

fuel processing facilities cited in this review, have consistently demonstrated the poor 

correlation between GA and BZ air sample data with BZ/GA ratios that spanned several 

orders of magnitude.

In addition, Section 3.2 of SCA-TR-2016-SEC009 recommended multiplying the GA air samples 

results by a factor of 10 to resolve the lack of parity between GA and BZ air concentrations (SC&A 

and Salient 2016).

In this report, Section 4.0 provides an overview of the actinide air sampling at ANL-W. Sections 5.0

and 6.0 address the specific issues. Section 7.0 discusses intake calculation for unmonitored actinide 
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intakes, and Section 8.0 provides the Oak Ridge Associated Universities (ORAU) Team’s 

conclusions. Attachment A provides the statistical evaluation of the ANL-W EBR-II Complex air 

sample data, and Attachment B provides diagrams of the ANL-W actinide-only areas.

4.0 OVERVIEW OF ACTINIDE AIR SAMPLING, 1958 TO 1976

The available information indicates that ANL-W predominantly used air monitoring to demonstrate 

compliance with contamination control requirements, posting requirements, and for determining 

respiratory protection requirements. Based on the low surface contamination levels and low air 

monitoring results for alpha radioactivity in those areas, the probability for a significant unmonitored 

intake of thorium, uranium, or plutonium to have occurred would have been low.

There are five major types of air sampling: (1) fixed GA air sampling, (2) portable GA air sampling, 

(3) fixed BZ air sampling, (4) portable BZ air sampling, and (5) personal BZ air sampling. The term 

“fixed” refers to an air sampler that is not mobile and has a fixed location. The term “portable” refers 

to a mobile air sampler that can be relocated as needed. The term “personal” refers to a portable air 

sampler that is mounted somewhere on a worker. One of the most common types of PASs is the 

lapel air sampler where the sampling head is clipped to a worker’s lapel. The U.S. Nuclear 

Regulatory Commission (NRC) states that air samples collected by lapel air samplers and air 

samplers within about 1 ft of the worker’s head are considered to be representative of the BZ air (NRC 

1992). As a result, the terms BZ air sample, personal air sample, and lapel air sample are 

interchangeable throughout this report. The order of the five types of air sampling listed above 

generally represents the reverse hierarchy for the preferred types of air sampling for estimating 

worker exposures, with personal BZ air sampling usually being the most preferred type. However, 

there are potentially significant sources of error unique to BZ air sampling that can generate 

questionable results. Those sources of error are addressed later in this document. 

In the ANL-W actinide-only areas, actinide air concentrations appear to have been measured using a 

combination of the first four types of air sampling with fixed GA air sampling eventually becoming the 

most common type. For the period in question (i.e., 1958 to 1976), no indication of personal BZ air 

sampling was found in the available radiological safety program documents for ANL-W. However, at 

least some of the samples from portable air samplers were BZ air samples. A 1967 memorandum on 

the health physics aspects of Cold-Line operations provides more than one indication that ANL-W was 

sampling a worker’s BZ for some of the air samples (Stoddart 1967). 

A review of the gross alpha air sampling results for the ANL-W actinide-only areas indicates that the 

majority of the elevated sample results were due to shorter decay periods before the final count was 

performed on the air samples. As a result, the radioactivity on those elevated air samples was still 

dominated by short-lived radionuclides attributable to radon and thoron progeny. This is 

demonstrated in a handful of the air sample results with final recounts performed 1,000 or more 

minutes after the samples were first counted. Numerous samples with final recounts roughly 

200 minutes after the samples were collected show the radioactivity on the air samples dropping by a 

factor of 2 or more. The air sample data indicates that the general practice at ANL-W was to recount 

the air samples only until the results dropped below the level of concern, which was usually 10% of 

the MPC but sometimes below 1% of the MPC. If the final counts on all of the air samples were 

performed several days after the air samples were collected, the mean air concentrations for ANL-W 

facilities would have been significantly lower. The air samples collected in the FCF were affected the 

most by radon and thoron progeny, which was most likely due to the much larger mass of concrete 

used for the construction of the FCF. When Cold-Line fuel production moved from the FCF to the new 

FASB, there was a significant drop in the mean air concentration for the Cold-Line fuel production. 

The majority of that decrease was most likely due to lower levels of radon and thoron progeny in the 

air at the new facility. 
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Figure 4-1 provides one of the more extreme examples of how significant the short-lived radioactivity 

was in ANL-W air samples. It is one of the few air samples with multiple recounts going out to about 

3 days. In Figure 4-1, the alpha radioactivity on the air sample decayed from 2,356 dpm to only 

5 dpm over 4,245 minutes (70.8 hours), which is a reduction by a factor of 471. Most ANL-W air 

samples did not receive a decay count much past the 200-minute mark. For this air sample, the alpha 

radioactivity had reduced by a factor of 1.9 after the first recount for gross alpha radioactivity at 

208 minutes.
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Figure 4

-1. Example of radioactive decay on an FCF air sample (ORAUT 2018a).
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4.1 SOURCE TERM AND AIR SAMPLING AT EBR-I COMPLEX, 1958 TO 1975

The period during which internal uranium exposures might not have been monitored, without MFPs 

being present, was from January 1, 1958, through June 13, 1975 (NIOSH 2016). Section 7.2.1.2 of 

the SEC-00224 evaluation report only indicated that potential intakes of uranium without mixed fission 

products could have occurred at the EBR-I Complex after 1957 (NIOSH 2016). That section did not 

provide an end date for those potential intakes. However, the end of Section 5.1.1.1 in the SEC00224 evaluation report indicated that the decontamination and decommissioning of the EBR-I 

Complex was completed on June 13, 1975 (NIOSH 2016). Therefore, January 1, 1958 through June 

13, 1975 was used as the period when unmonitored uranium intakes could have occurred in the 

EBR-I Complex. Early air sampling at the Zero Power Reactor-III (ZPR-III) was done by putting a 

large piece of filter paper on the (clamp-down) head of a high-volume air mover, not by permanent air 

samplers (ORAUT 2014a). Based on the available air sample datasheets, a variety of high- and lowvolume air samplers were being used at the EBR-I Complex.

The EBR-I Complex consisted of several nuclear reactors and did not contain any radioactive material 

production facilities (NIOSH 2016). As indicated in the SEC-00224 evaluation report, the potential for 

actinide-only internal exposures at the EBR-I Complex was limited to unirradiated uranium at the 

ZPR-III (NIOSH 2016). The unirradiated uranium source terms at the EBR-I Complex were limited to 

uncoated depleted uranium metal plates, unirradiated reactor fuels, and radioactive check and 

calibration sources. The available information indicates that all of the EBR-I Complex reactor fuels 

were either clad in nonradioactive metal or coated with Teflon-based protective coating (called Kel-F) 

(NIOSH 2016), making them an unlikely source for internal exposures. Any uncoated or 

unencapsulated radioactive check and calibration sources were also an unlikely source for internal 

exposures due to how they would have been handled and stored to maintain their usefulness as a 

check and calibration source. Therefore, the primary source term for potential internal uranium 

exposures was the oxidized uranium on the uncoated depleted uranium metal plates. This 

contamination was typically found on the gloved hands of workers handling the bare plates (NIOSH 

2016). However, in August 1961, all of the depleted uranium metal plates from ZPR-III were replaced 

with new plates that had a protective coating. The original depleted uranium metal plates were 

deteriorating and becoming an increasing pyrophoric hazard. The new plates had a protective coating 

to minimize future deterioration and pyrophoric hazards (ANL 1960–1961).

According to a former ZPR-III employee who worked as an operator, supervisor, and experimenter 

while there, they never saw more than trivial contamination levels of depleted uranium. The effort was 

put on contamination control. Air contamination never seemed to be a problem. Most of the dust was 

depleted uranium, but not much was suspended (ORAUT 2014a, 2015a). This information is 

supported by many of the gross alpha radioactivity air concentrations reported by the site for EBR-I 

Complex, which were often reported as being less than 1% of the MPC.

Based on the radioactive source term information and information from worker interviews, the amount 

of potentially dispersible uranium without MFPs being present was low, making internal uranium 

exposures at the EBR-I Complex unlikely. 

4.2 ACTINIDE AIR SAMPLING AT EBR-II COMPLEX, 1963 TO 1976

The following sections provide information about the specific actinide-only areas at the EBR-II 

Complex.

Because of the number of elevated air sample results for the EBR-II Complex thorium and uranium 

areas and the FCF plutonium areas, a closer review of that data was performed for this report. 

Because the evaluated air sampling data were scattered throughout numerous documents in the Site 

Research Database (SRDB), all of those air sample results were put in sequential order and 
Document No. ORAUT-RPRT-0089 Revision No. 00 Effective Date: 04/19/2022 Page 16 of 72

consolidated into three documents (ORAUT 2018a, 2018b, 2018c). One document contains all of the 

thorium-only area air data, one the uranium-only area data, and the third the plutonium-only area data 

for those areas. The original SRDB Reference Identification (SRDB Ref ID) numbers and page 

numbers were added to the top of each page in red font. The three new documents include 

bookmarks, highlights, and notations about how certain records were interpreted or corrected for this 

evaluation. The observations made during the closer review of the air sample results for the EBR-II 

Complex thorium and uranium areas and the FCF plutonium areas resulted in those datasets being 

put through a statistical evaluation.

4.2.1 Thorium Areas, 1963 to 1967

Unmonitored internal thorium exposures could have occurred without MFPs being present from 

August 1963 through November 1967 (NIOSH 2016). During this period, portable Filtronics air 

samplers with a flow rate of 1 cfm were used to collect air samples in the Mold Preparation Room

(FCF Room 25), which was the only location where the thoria (ThO2) was used (ORAUT 2018a). 

The gross alpha radioactivity air concentrations reported by the site for FCF Room 25 were normally 

less than 10% of the MPC even during a thoria spill incident (ORAUT 2018a). Even then much of the 

radioactivity on the air samples was likely still attributable to short-lived radon and thoron progeny 

(ORAUT 2018a). Air concentrations in FCF Room 25 were generally low because only small 

quantities of thoria were being used inside of a ventilation hood. Further, the thoria was mixed with 

ethyl alcohol before coating the casting molds, which would have reduced its ability to become 

airborne (NIOSH 2016).

Figure 4-2 is the Radiation Survey Report for a September 18, 1963 thoria spill (ORAUT 2018a). 

Figures 4-3 and 4-4 are the air sample results for the air samples that were collected during the 

incident and during the cleanup, which was performed immediately after the incident. Even when 

thoria was spilled from the ventilation hood to the floor, the air concentrations never exceeded 10% of 

the MPC of 132 dpm/m3

. It should be noted that ANL-W used MPCs of 66.6 dpm/m3

(3 × 10-11 μCi/cm3

) and 132 dpm/m3 (6 × 10-11 μCi/cm3) for FCF Room 25 versus the MPC for 232Th. 

Based on the 1963 addendum to National Bureau of Standards (NBS) Handbook 69, the MPC for 

232Th was 1 × 10-11 μCi/cm3

, which is equivalent to 22.2 dpm/m3 (NBS 1963). 
Document No. ORAUT-RPRT-0089 Revision No. 00 Effective Date: 04/19/2022 Page 17 of 72

Figure 4-2. Radiation Survey Report for a thoria spill (ORAUT 2018a).
Document No. ORAUT-RPRT-0089 Revision No. 00 Effective Date: 04/19/2022 Page 18 of 72

Figure 4

-3. Air sample collected during the thoria spill (ORAUT 2018a).
Document No. ORAUT-RPRT-0089 Revision No. 00 Effective Date: 04/19/2022 Page 19 of 72

Figure 4

-4. Air sample collected during cleanup of the thoria spill (ORAUT 2018a).
Document No. ORAUT-RPRT-0089 Revision No. 00 Effective Date: 04/19/2022 Page 20 of 72

4.2.2 Uranium Areas, 1967 to 1976

The SEC-00224 evaluation report currently indicates that the period that unmonitored internal uranium 

exposures could have occurred without MFPs being present was from August 1967 through June 

1983 (NIOSH 2016). However, the potential unmonitored uranium exposure period has been reduced 

to August 1967 through June 1976 for this report. This is a result of information that was captured

after the SEC-00224 evaluation report was written. Previously available information combined with 

the new information helped to determine that the FASB only handled encapsulated uranium, which 

means it was not a significant internal dose concern. 

Section 7.2.2.2 of the SEC-00224 evaluation report indicated that the potential for unmonitored 

uranium intakes at ZPPR was minimal and any unmonitored exposures could be bound using 10% of 

the MPC (NIOSH 2016). However, the SEC-00224 evaluation report did not provide any date range 

for when those minimal uranium exposures could have occurred. As part of researching what that 

date range should be, it was determined that the potential for the unmonitored actinide exposures at 

ZPPR would have only been possible when there was a cladding breach in a plutonium fuel element 

or uranium-plutonium fuel element. In the event of a cladding breach for a uranium-plutonium fuel 

element, any internal doses attributable to uranium would have been negligible compared to the 

internal doses attributable to plutonium. Therefore, only unmonitored plutonium intakes will be 

assessed for ZPPR workers.

4.2.2.1 Basis for Exposure Period Change

For the SEC-00224 evaluation report, the primary references for the FASB were the 1981 master site 

plan for ANL-W (ANL-W 1981), a FASB vault safety review from 1972 (Abrams 1972), and a mostly 

unlabeled floorplan for the FASB that was used for the radiological survey reports (ANL-W 1983). 

Page 17 of ANL-W (1981) indicates that the FASB was equipped for the assembly of EBR-II driverfuel pins into fuel elements, fabrication of EBR-II driver and experimental subassemblies, and 

inspection and storage of the elements and finished subassemblies. Page 46 of ANL-W (1981) 

indicates that FASB use was limited to preirradiation inspection, assembly, and testing of EBR-II fuel. 

A preliminary proposal from July 1969, which was not used as a reference for the SEC-00224

evaluation report because it was dated before the FASB was constructed in 1971, indicated that the 

FASB operations were supposed to include all of the Cold-Line fuel production processes (ANL 

1969a). Because of the assembly of EBR-II driver-fuel pins into fuel elements work listed in ANL-W 

(1981) and because of the information in the preliminary proposal document, it was assumed for the 

SEC-00224 evaluation report that all of the Cold-Line processes took place at the FASB, including the 

ones that involved processing unencapsulated uranium. 

Since the SEC-00224 evaluation report was written, two new pieces of information have been found. 

In Appendix E of a 1978 FASB Vault Safety Assessment, which was the firefighting plan for ANL-W, 

statements indicated that only encapsulated uranium was handled at the FASB (ANL-W 1978, 

pp. 151–153). During its Cold-Line fuel production era, the FASB consisted of three major areas: the 

West Room, East Room, and Storage Vault. Based on Appendix E of ANL-W (1978), the East Room 

was primarily for the nonradiological processes (e.g., sodium metal work and subassembly 

production) and the West Room was where the uranium fuel was manufactured. Appendix E of 

ANL-W (1978) also states, “The west room of FASB, including the vault, houses only jacketed or 

contained uranium alloy.” It also indicates that both irradiated and unirradiated samples were 

examined in the West Room, which were potentially small samples of unjacketed uranium (ANL-W 

1978). In addition to the 1978 FASB Vault Safety Assessment, detailed floorplans of the FASB from 

1971 were captured after the SEC-00224 evaluation report was written (ANL-W 1971). Those 

floorplans depicted equipment layouts for the three main parts of the FASB (i.e., East Room, West 

Room, and Vault). There is no equipment for uranium metal alloy melting, pin casting, skull recovery 
Document No. ORAUT-RPRT-0089 Revision No. 00 Effective Date: 04/19/2022 Page 21 of 72

(uranium-bearing melt residue recovery), and pin shearing operations in those diagrams (ANL-W 

1971). Those were the major Cold-Line processes that involved work with unencapsulated uranium.

The continued use of FCF Room 20 and the use of vendor-produced fuel elements appear to have 

eliminated the need for the FASB to have any fuel casting capability. The pin casting work in FCF 

Room 20 was shut down in December 1969 and put on standby until 1982. The available information 

indicates that enough acceptable vendor-produced fuel elements (clad fuel pins) were being provided 

to the site by December 1969. Initially, there were some problems with the vendor producing 

acceptable fuel elements, which is likely why FCF Room 20 was kept on standby (ANL 1968, 1969b, 

1970a, 1970b, 1970c). 

A review of the available FASB air sample data indicates that the FASB gross alpha air sampling data 

support this determination. The 95th percentile of the unadjusted gross alpha air concentrations for 

the FASB is only 3.9% of the MPC. It is also worth noting that many of the higher results occurred 

after May 1977 when the typical sample decay period appears to have been reduced to only 

23 minutes on average. Therefore, a significant fraction of the radioactivity for those samples’ results 

is likely attributable to radon progeny. None of the FASB air samples with sample decay times of 

1,440 minutes (24 hours) or greater had a result above 0.6% of the MPC (ORAUT 2021a). 

A review of the available FASB survey data after 1975 indicates that the GA survey records for the 

FASB also support this determination. GA surveys of the FASB Storage Vault were performed daily 

before April 1976, twice weekly from April 1976 to February 1977, weekly from March 1977 to 

September 1977, and twice monthly after September 1977. For the FASB production areas, GA

surveys appear to have always been performed monthly. The higher contamination survey frequency 

for the Storage Vault indicates that the Storage Vault was considered to be a higher risk area for 

contamination than the FASB production areas. With the exception of a couple FASB Vault surveys, 

all of the FASB GA alpha contamination survey results that were reviewed were reported as 

<50 dpm/100 cm2

, which was the minimum reporting level for alpha contamination in the surveys. 

4.2.2.2 Review of Air Sampling Data

During the period from August 1967 through June 1976, air sampling data for 1970, 1973, and 1975 

were more sparse than other years. This was likely due to the intermittent nature of the Cold-Line fuel 

production work at ANL-W and the transfer of those production facilities to the new FASB. In addition, 

beginning around June 1968, ANL-W started receiving vendor-produced Cold-Line fuel pins and fuel 

elements to initially supplement and eventually replace ANL-W’s Cold-Line fuel pin and fuel

production (ANL 1968). In December 1969, Cold-Line fuel pin and fuel element production at the FCF 

was put on standby until 1982 (NIOSH 2016). Radiological survey records indicate that Cold-Line fuel 

production did not start at the new FASB until at least February 1972 (NIOSH 2016). Even though 

some significant gross alpha radioactivity air concentrations for EBR-II Complex uranium-only areas 

were reported by the site, the majority of the gross alpha radioactivity air concentrations for those 

areas were reported as 10% of the MPC or less (ORAUT 2018b). For the uranium-only areas at the 

EBR-II Complex, ANL-W normally used a MPC of 132 dpm/m3 (6 × 10-11 μCi/cm3

) from 1967 through 

September 1972. After September 1972, a MPC of 220 dpm/m3 (1 × 10-10 μCi/cm3

) was normally 

used.

Beginning in May 1975, ANL-W changed the format of the Air Sample Data sheets, and the Time On 

and Time Off information was no longer being routinely reported for the air samples. However, the 

Run Time for the air samples was still being reported. Based on the few records with sample on/off 

information in the Remarks column of the new form, the Date and Time column on the new forms only 

provide the sample count dates and times. Figure 4-5 is an example of the revised Air Sample Data 

sheets (ORAUT 2018b). Also by May 1975, routine air sample collection times at FCF periodically 

began going up to and over 10,080 minutes (~7 days). 
Document No. ORAUT-RPRT-0089 Revision No. 00 Effective Date: 04/19/2022 Page 22 of 72

Figure 4

-5. Example of a post-April 1975 air sample record (ORAUT 2018b).
Document No. ORAUT-RPRT-0089 Revision No. 00 Effective Date: 04/19/2022 Page 23 of 72

4.2.3 Plutonium Areas, 1970 to 1975

At the EBR-II Complex, there were two distinctly separate sets of plutonium work being performed: 

one at ZPPR and the other at the FCF.

4.2.3.1 Zero Power Plutonium Reactor

Section 7.2.2.2 of the SEC-00224 evaluation report currently states the following (NIOSH 2016): 

The potential for plutonium exposures within ZPPR was minimal because encapsulated 

and coated fuel sources were used. ZPPR Health Physics used air monitoring data 

and/or contamination smear data to control potential plutonium exposure at this reactor 

during normal operations. Loading and unloading procedures were carefully monitored 

to detect any loose contamination on fuel plates. Exposures to plutonium without mixed 

fission products can be bound using 10% of MPC (maximum permissible concentration) 

from available air monitoring data. 

Based on those statements, potential unmonitored intakes of plutonium without MFPs at ZPPR will be 

based on 10% of the MPC value being used for the air monitoring program at ZPPR. Based on the 

available air monitoring data, an MPC of 4.4 dpm/m3 (2 × 10-12 μCi/m3

) was used for the airborne 

alpha radioactivity at ZPPR, which makes the bounding air concentration 0.44 dpm/m3

. 

Because the SEC-00224 evaluation report did not provide a date range for when those potential 

unmonitored intakes were possible, a basis for that date range has been provided as part of this 

report. The earliest indication of plutonium contamination at ZPPR was during the October 9, 1968, 

inspection on the receipt of the original driver fuel. The receipt inspection of that fuel found low levels 

of plutonium contamination on two fuel elements (Koplin 1975). However, ANL-W procedures 

indicate plutonium intakes were unlikely during the receipt inspections because the workers wore 

respirators and because the fuel was handled inside of a hood until it was confirmed as being free of 

contamination (ANL-W 1968a, 1968b). Therefore, plutonium found during receipt inspections was not 

considered a potential intake hazard. 

On September 1 and 2, 1970, a reactor safety survey was performed at ZPPR. Section C.4, “Health 

Physics Procedures,” of the report for that safety survey indicated that no leaky fuel had been 

encountered during the entire operating period of the ZPPR. However, during the safety survey, a 

minor plutonium contamination incident occurred. Some plutonium came loose from a small 

plutonium source on a planchet that was used to check plutonium monitor calibrations. The report 

also indicated that only the source storage container was contaminated (ANL-W 1970). Based on the 

available information, this plutonium contamination was not likely airborne or an intake hazard. In 

addition, the plutonium contamination was likely in the form of nonrespirable flakes. Even though a 

plutonium intake was unlikely from this incident, September 1, 1970, will be used for the start date for 

the potential unmonitored plutonium intakes at ZPPR. 

On July 31, 1975, a plutonium bioassay program was initiated for ANL-W workers (Madison 1975). 

Therefore, the period that unmonitored intakes of plutonium without MFPs were potentially received at 

ZPPR is assumed to be September 1, 1970, through July 31, 1975. 

4.2.3.2 Fuel Cycle Facility Areas

The SEC-00224 evaluation report currently indicates that the period when unmonitored internal 

plutonium exposures could have occurred without MFPs being present was April 1970 through 

December 1972 for FCF. The date range for the potential plutonium exposure period at the FCF was 

based on the period when the Mark-II type RAS-TREAT sodium-loop experiment work was being 
Document No. ORAUT-RPRT-0089 Revision No. 00 Effective Date: 04/19/2022 Page 24 of 72

performed at the FCF (NIOSH 2016). However, the potential unmonitored plutonium exposure period 

for the FCF work has been extended through April 1973 for this report.

At the FCF, the Mark-II type RAS-TREAT sodium-loop work began in April 1970 and was initially 

limited to Rooms 27, 28, and 29. In early 1971, a glovebox was installed in Room 22 (i.e., the former 

FCF Machine Shop) to handle the Mark-II type RAS-TREAT sodium-loop experiments. The glovebox 

became operational sometime after May 1971. After 1972, RAS-TREAT sodium-loop experiment 

work at the FCF appears to have been performed exclusively in the Argon Cell, and the experiments 

were irradiated to much higher levels (Cook et al. 1975). However, additional plutonium air monitoring 

data for FCF Room 22 was found after the SEC-00224 evaluation report was completed. The more 

recently discovered data indicates that plutonium exposures without the MFPs were likely possible at 

the FCF through April 1973 (ORAUT 2018c). Plutonium air monitoring was likely performed in FCF 

Room 22 until the room was decontaminated. Therefore, the potential period for unmonitored 

plutonium exposures at the FCF without MFPs has been extended through April 1973. 

Based on the available air monitoring data, ANL-W used an MPC of 4.4 dpm/m3 (2 × 10-12 μCi/m3

) for 

the FCF plutonium areas. With the exception of the FCF sodium-loop work areas during 1970 to 

1973, the gross alpha radioactivity air concentrations reported by the site for plutonium-only areas 

were normally less than 10% of the MPC (ORAUT 2018c). However, some of the elevated air sample 

results during those years were likely attributable to short-lived radon and thoron progeny. Based on 

the decay times for the final counts on the elevated samples, the final counts for nearly all of the 

elevated air samples were performed before all of the short-lived radon and thoron progeny could 

have decayed. The final counts for the air samples should have been performed after a 3.5-day 

(5,110-minute) decay time to provide sufficient time for all of the short-lived radon and thoron progeny 

to decay out of the air samples. Additionally, some of the elevated results could have been 

attributable to uranium, since uranium work was also periodically performed in those FCF rooms. All 

but one of those results were below 10% of the MPC for uranium.

Until June 1972, portable Filtronics air samplers with a flow rate of 1 cfm were used to collect air 

samples in the plutonium area (FCF Room 22) (ORAUT 2018c). After June 1972, a special alpha 

sampler with a flow rate of 1 cfm was used in the plutonium area (ORAUT 2018c). 

5.0 RESPONSE TO ISSUE 1

There are two distinctly separate parts to Issue 1 in SCA-TR-2016-SEC009 (SC&A and Salient 2016). 

Because of that, each of those parts is addressed separately below.

5.1 PART 1 OF ISSUE 1

The first concern centers around the use of GA air samplers with a low airflow. Low 

airflow rates required sampling times of up to 4 days, which necessarily correspond to 

periods when no work was performed (and no workers were present).

The ORAU Team is not certain how to address the first part of Issue 1 because lapel samplers 

typically have much lower flow rates and because it is contradictory to the argument in SCA-TR-2016-

SEC009 that only lapel air samplers can provide representative BZ air samples (SC&A and Salient 

2016).

At ANL-W, the air samplers in the thorium and uranium areas had flow rates typically ranging between 

1.0 and 2.0 cfm, and the air samplers in the plutonium areas had flow rates of 1 cfm. As a 

comparison, most lapel samplers only operate at 2 lpm, which is equivalent to 0.07 cfm. Minimum 

detectable concentration calculations, based on the site’s 5-minute count time, indicate that the 

ANL-W air samplers could detect <10% of the MPC for an 8-hour sample period and could detect 
Document No. ORAUT-RPRT-0089 Revision No. 00 Effective Date: 04/19/2022 Page 25 of 72

nearly 10% of the MPC for a 4-hour sample period. Whereas, a lapel sampler would need to sample 

for approximately 60 hours to detect 10% of the MPC with a 5-minute count time, or sample for 

24 hours to detect 10% of the MPC with a 60-minute count time. Therefore, the concern that low 

airflow rates required sampling times of up to 4 days for GA air samples is not substantiated. 

5.2 PART 2 OF ISSUE 1

... Thus, it is reasonable to conclude that air concentrations during non-working hours

differed significantly from air concentrations that would have exposed workers during 

normal facility operations that were likely limited to an 8-hour shift Monday through Friday 

during many periods.

The ORAU Team has evaluated this and agrees it is a legitimate concern, even though a number of 

factors would mitigate any biases that are not favorable to claimants by using the reported air 

concentrations as is. The following are a few of the key factors that would mitigate those potential 

biases:

1. Typical low levels of removable contamination in those areas are not indicative of areas with 

significant levels of airborne radioactivity during operational or nonoperational periods, 

2. Not accounting for the actual occupancy in those areas,

3. Not accounting for the respirable fraction of the airborne radioactivity,

4. Not accounting for the times respirators were worn, and

5. ANL-W not allowing for adequate decay of the short-lived radionuclides before performing final 

counts on the air samples. 

With two exceptions, the ANL-W air sampling data for the actinide-only areas were evaluated, and 

calculations were performed, to address the air sample dilution concern. The air sampling data for the 

EBR-I Complex were not evaluated or adjusted because of the small potential for internal uranium 

exposures at the EBR-I Complex. As indicated in Section 4.1, very little of the uranium in the EBR-I 

Complex was dispersible. The air sampling data for the EBR-II Complex ZPPR were not evaluated or 

adjusted because of the small potential for internal plutonium exposures at ZPPR. As indicated in 

Section 7.2.2.2 of the SEC-00224 evaluation report, the potential for plutonium exposures within 

ZPPR was minimal because encapsulated and coated fuel sources were used (NIOSH 2016). These 

determinations are also supported by EBR-I Complex and ZPPR air sampling data, which indicate 

that significant levels of airborne uranium and plutonium contamination without MFPs present were 

rare and that air sampling appears to have mostly been performed as a precautionary measure. 

Therefore, it is reasonable to bound the potential unmonitored uranium and plutonium intakes 

attributable to those facilities based on an average concentration of 10% of the MPC without further 

adjustment.

Only the gross alpha radioactivity air concentrations were evaluated because potential unmonitored 

actinide exposures when MFPs were not present was the only unmonitored internal dose concern at 

the EBR-II Complex. As indicated in Section 4.2, the evaluated air sampling data for each area were 

put in sequential order and consolidated into three documents (ORAUT 2018a, 2018b, 2018c). 

Additionally, a review of the available air sampling data indicates that ANL-W inconsistently applied its 

alpha self-absorption factor of 1.3 to the air sample results for alpha radioactivity. Therefore, 

whenever the air sample results indicate that no alpha self-absorption factor was applied or was not 

known to have been applied (as in the monthly Health Physics Reports), those results were multiplied 

by a factor of 1.3. 
Document No. ORAUT-RPRT-0089 Revision No. 00 Effective Date: 04/19/2022 Page 26 of 72

For the four air sample datasets that were evaluated and adjusted for potential dilution, statistical 

evaluations were performed on each of those datasets to verify that their distributions could be 

represented by a lognormal distribution. The four air sample datasets included the air sample results 

for the EBR-II Complex thorium and uranium areas and the FCF plutonium areas, with the air sample 

results for the uranium areas being divided into two separate datasets. Because all four of those air 

sample datasets had a reasonable fit to a lognormal distribution and because the sample times varied 

significantly with the shortest duration air samples often having the highest results, time-weighted 

geometric mean (GM) air concentrations were calculated for each of those air sample datasets. In the 

instances of overlapping sample periods that were from different work areas for a given dataset, no 

samples were excluded from the calculation of the time-weighted GM air concentration, even though a 

worker could not have been simultaneously at both locations. The primary reasons for not excluding 

overlapping air samples include partial overlaps and differing sample run times. Attachment A 

provides the details on how those statistical evaluations were performed along with the resulting timeweighted GM air concentrations and geometric standard deviations (GSDs) for each of those 

datasets. Sections 5.2.1 and 5.2.2 provide more details about the two different approaches that were 

used to adjust the four air sample datasets for potential dilution. 

5.2.1 EBR-II Complex Air Sampling Data, 1963 through 1974

With the exception of the ZPPR data, all of the available air sampling data from the actinide-only 

areas for August 1963 through December 1974 were evaluated. As discussed above, the ZPPR air 

sample data were not evaluated or adjusted because of the small potential for internal plutonium 

exposures at ZPPR. With the exception of short duration samples, all of the air concentrations for this 

period were adjusted to eliminate sample dilution from sampling during nonoperational periods. To do 

that, the concentrations of airborne alpha radioactivity were assumed to drop to zero during the 

nonoperational periods, and the air sample volumes were reduced to eliminate the volumes of air 

collected during the nonoperational periods. The adjusted air concentrations were calculated based 

on the volumes of air sampled during the operational periods. Those adjusted air sample volumes 

were calculated using the reported sampler flow rates and calculated operational times, which are 

referred to as potential exposure times below and in the calculations (ORAUT 2021a, 2021b, 2021c). 

Based on an FCF worker interview, a second work shift was started at the FCF in the late 1960s that 

continued until the Hot Fuel Examination Facility-North came online in March 1975 (ORAUT 2015b; 

NIOSH 2016). Based on sample start and end times and notes in many of the air sample records, 

two 8-hour shifts were being worked at the FCF by August 1967. Therefore, assuming 16 hours of 

exposure time for each full workday is considered reasonable for the period of August 1967 through 

March 1975. For all other periods, a single 8-hour shift was assumed for each workday. 

Given that ANL-W normally exchanged air samples on Monday to Friday during the first or second 

shift, short duration samples were likely collected outside of that period because of some radiological 

work being performed during off-shift hours. Adjusting the short duration samples to eliminate sample 

dilution could cause the results for the samples collected during off-shift hours to be zeroed out or 

inappropriately biased high. Therefore, the short duration samples with sample times of 18 or less 

hours were not adjusted to eliminate potential sample dilution. The up to 18 hours assumes that two 

8-hour shifts were worked and allows for up to 2 extra hours for additional work outside of the 

scheduled shifts (e.g., performing radiological surveys, collecting air samples). 

For sample times greater than 18 hours, up to 16 hours of exposure time was assumed for each full 

workday (Monday through Friday, excluding federal holidays). Only the observed federal holidays for 

a given year were accounted for in these calculations because no information was found about which 

holidays were observed by ANL-W. The amount of time worked during partial workdays was 

calculated based on the sample exchange times and the assumed workday start and end times. The 

available records indicate that the workday started at 8:00 a.m.
Document No. ORAUT-RPRT-0089 Revision No. 00 Effective Date: 04/19/2022 Page 27 of 72

For FCF air sample results that are only summarized in the monthly health physics reports, all 

exposure periods for those air samples were assumed to be 16 hours, unless additional information 

indicated otherwise. This was assumed for the period before March 1975, because the available 

information indicated that two 8-hour shifts were being worked at the FCF from August 1967 to March 

1975 (ORAUT 2015b; NIOSH 2016). After that, FCF operations returned to a single day shift 

(ORAUT 2015b; NIOSH 2016). However, this exposure period assumption was not reduced to 

8 hours, because no summarized air sample results from the monthly health physics reports were 

used after 1969.

Equations 5-1 and 5-2 are provided to better illustrate how these air concentration adjustments were 

performed. 

(5-1) TO = − T T S N

where

TO = total operating time during the sampling period (min) 

TS = total sampler run time during the sampling period (min) 

TN = total nonoperational time during the sampling period (min) 

Note that the calculation of TN is actually rather complicated. For more details on how TN was 

calculated please refer to the spreadsheets for those calculations, which are in ORAUT (2021a, 

2021b, and 2021c). 

(5-2) ( ) ( ) ( )

( ) ( ) O

A AF CF C

F T Adj =

where

CAdj = adjusted GA air concentration (dpm/m3

) 

A = gross alpha radioactivity on the sample (dpm)

F = sampler flow rate (cfm)

TO = total operating time during the sampling period (min)

AF = 1.3 self-absorption factor for the measured alpha radioactivity, but only included 

it in these calculations when it was omitted in the results reported by ANL-W. 

The ANL-W Air Sample Data sheets clearly indicate when this factor was or 

was not applied.

CF = volume conversion factor used by ANL-W (35.4 ft3

/m3) 

5.2.2 EBR-II Complex Air Sampling Data, 1975 to 1976

As previously indicated, beginning in May 1975 ANL-W changed the format of the Air Sample Data 

sheets; air sample on/off dates and times were no longer routinely reported. Because of that, the air 

sample results from January 1975 through June 1976 could not be adjusted to eliminate the influence 

of the nonoperational periods from the samples. Therefore, instead of adjusting the air 

concentrations, the intake calculations were modified to address the air sample dilution concern. In 

the approach that involved adjusting the air sample results, a major assumption for that approach was 

that the airborne radioactivity dropped to zero during the nonoperational periods at the facility. Using 

that same assumption, the sample dilution concerns can be eliminated by assuming that the workers 

were exposed to unadjusted air concentrations day and night for 1 year (24 hr/d 365 d/yr). It should 

be noted that this approach only works when the daily operational period for the facility is the same as 

the daily exposure period assumed for the workers (i.e., no second or third shifts at the facility). For 
Document No. ORAUT-RPRT-0089 Revision No. 00 Effective Date: 04/19/2022 Page 28 of 72

the period that this approach was being used, the FCF was only operating for one work shift per 

workday. For more details on this approach, including the equations that were used, refer to 

Section 7.1.2.2. 

6.0 RESPONSE TO ISSUE 2

A second and more serious concern regarding the use of GA air sampling data is the 

generic lack of parity between air concentrations measured by GA air samplers and lapel 

air samplers worn by the individual workers. Study data, including those of two nuclear 

fuel processing facilities cited in this review, have consistently demonstrated the poor 

correlation between GA and BZ air sample data with BZ/GA ratios that spanned several 

orders of magnitude.

The end of Section 3.2 in SCA-TR-2016-SEC009 suggests applying a factor of 10 to convert all GA 

air concentrations to BZ air concentrations as a means for addressing the lack of parity between GA 

and BZ air concentrations (SC&A and Salient 2016). The suggested factor of 10 in SCA-TR-2016-

SEC009 is based on the data from only two studies (Brunskill and Holt 1967; Caldwell, Potter, and 

Schnell 1967), both of which were completed in 1967. Those studies represent two of the earliest 

known studies involving lapel air samplers, which is a specific type of BZ PAS device. Because many 

other air sampling studies exist, most of which are more recent than 1967, this report evaluated the 

results from some of those studies. The results from those studies are discussed in the subsections 

below.

6.1 APPLICABILITY OF STUDIES FROM OTHER SITES

There are several highly variable parameters that can affect the outcome of air sampling studies, 

often making most of their results and conclusions only applicable to the specific conditions and 

location of the study. Therefore, it is important to determine the applicability of an air sampling study 

from one site or workroom before applying its results to another site or workroom. 

6.1.1 Key Parameters in ORAUT-RPRT-0097

ORAUT-RPRT-0097, Breathing Zone to General Area Air Concentration Ratios in Small Workrooms

(ORAUT 2021d), identifies several parameters that can influence the BZ:GA ratios in a workroom (see 

Section 2.0 of that reference). As indicated in Section 11.0 of ORAUT-RPRT-0097, there are five key 

parameters that should be considered when justifying the application of BZ:GA ratios from other 

locations to the location of interest. Those parameters include (1) room size, (2) particle size 

distribution for the airborne radioactivity, (3) ventilation rate for the room, (4) room complexity, and 

(5) the presence of dominant particles. The following partial excerpts from Section 2.0 of ORAUTRPRT-0097 briefly explain how each of those parameters can affect the air concentrations and/or 

BZ:GA ratios for a workroom. More detailed discussions can be found in ORAUT-RPRT-0097. 

Total workroom or air space volume is a key parameter that usually influences how much an aerosol 

can or cannot be dispersed. However, with the exception of rooms where elevated releases above 

the typical BZ can occur, a room’s height is usually less of a factor influencing the aerosol 

concentration than its other dimensions because of gravitational settling. Therefore, in most instances 

room sizes can be compared in terms of area for determining applicability of the BZ:GA ratio 

information in this report to a specific room (ORAUT 2021d). At ANL-W, the actinide-only areas were 

relatively small workrooms in comparison with the actinide handling and processing areas at other 

sites. Table 6-1 provides the room size information for each of the areas where unmonitored actinide 

intakes need to be assessed per the SEC-00224 evaluation report (NIOSH 2016). Note that 

Table 6-1 does not contain any area room size information for the Low Bay Area at the ITF, because 

no dimensions or scale drawings were found for that area. The only known size information for the 
Document No. ORAUT-RPRT-0089 Revision No. 00 Effective Date: 04/19/2022 Page 29 of 72

ITF is for the size of the entire building. Section 5.1.2.5 of the SEC-00224 evaluation report indicates 

that the ITF building had a total square footage of 3,455 ft2 (320 m2

) (NIOSH 2016). 

Table 6-1. ANL-W actinide-only area room sizes (ORAUT 2019). 

Process area (actinide)

Total area

(ft2)

Total area

(m2)

FCF Room 20 (U) 672 62.4

FCF Room 22 (U) 672 62.4

FCF Room 23 (U) 528 49.1

FCF Room 25 (Th) 369 34.3

FCF Room 26 (U) 579 53.7

FCF Room 27 (Pu) 222 20.6

FCF Room 28 (Pu) 479 44.5

FCF Room 29 (Pu) 994 92.3

ITF Low Bay Area (U) Unknown Unknown

ZPR-III Workroom (U) 303 28.1

ZPPR Workroom (Pu) 1,662 154.4

Because only respirable size particles (i.e., those particles with aerodynamic diameters smaller than 

about 20 μm) contribute to inhalation intakes, only the respirable fraction of an aerosol is of interest for 

assessing the inhalation intakes of workers. Therefore, the ideal air sample measurements for 

assessing a worker’s inhalation exposure would discriminate against collecting nonrespirable 

particles. The presence of larger nonrespirable particles on any air sample being used to estimate a 

worker’s inhalation intake, whether it is a BZ or GA air sample, could result in an unreasonable 

overestimate of the worker’s intake (ORAUT 2021d). 

The ventilation rate in terms of air changes (ACs) per unit time, usually in units of AC/hr, is a removal 

constant for the aerosol within a workroom. This removal constant is directly related to the depletion 

of the aerosol from an acute, constant, or variable release of aerosol and can be used to estimate the 

average air concentration at different points in time when the release quantity (instantaneous or puff 

releases) or release rate (longer term releases) along with several other parameters are known 

(ORAUT 2021d). 

Factors contributing to the complexity of a workroom include room layout, obstructions, heat sources, 

room ventilation inlet and outlet locations, local exhaust locations, and general flow directions. All of 

these affect the airflow patterns and level of mixing within a room. These things tend to cause air 

streaming, dead air spaces, eddy currents, etc. within a room, which result in larger air concentration 

gradients. The more complex a room is, the more likely larger air concentration gradients are 

(ORAUT 2021d). To facilitate evaluating the complexity of the actinide-only areas at ANL-W, the 

available floorplans were compiled and provided in Attachment B. 

Low numbers of airborne radioactive particles in a workroom are normally a good thing and a 

desirable condition for the workplace atmosphere. However, when those low-in-number radioactive 

particles have a relatively high specific activity and radiotoxicity (i.e., what this report refers to as 

“dominant particles”), that condition can become very problematic and have a very adverse effect on 

the ability to collect meaningful and reproducible air samples. When the numbers of the radioactive 

particles are too low, both BZ and GA air sampling becomes a stochastic process versus a 

deterministic process because there just are not enough radioactive particles to go around. Because 

this normally is not a significant issue of concern for lower specific activity and less radiotoxic 

particles, this situation is primarily relevant to when the low-in-number radioactive particles have a 

relatively high specific activity and radiotoxicity (ORAUT 2021d). 
Document No. ORAUT-RPRT-0089 Revision No. 00 Effective Date: 04/19/2022 Page 30 of 72

6.1.2 Air Concentration Levels

In addition to the parameters above, higher maximum air concentrations can cause larger 

discrepancies between BZ and GA air sample results. For example, if the highest air concentration in 

a room is only 1 MPC, the potential air concentrations in the room can only range from zero to 1 MPC. 

Whereas, when the highest air concentration is only 10% of the MPC, the range is only from zero to 

10% MPC, which means the potential differences between the GA and BZ air concentrations would 

also be smaller. The graphs presented in Caldwell, Potter, and Schnell (1967) and Brunskill and Holt 

(1967) for the Nuclear Materials and Equipment Corporation (NUMEC) plutonium laboratory and the 

Windscale Works show this effect. Those graphs are presented in this report as Figures 6-1 and 6-2. 

As indicated in those graphs, the BZ:GA ratios tend to be closer to unity when the air concentrations 

are low, even for larger workrooms. Given that the reported air concentrations for the ANL-W were 

generally low (ORAUT 2018a, 2018b, and 2018c), air sampling studies performed in facilities with 

significantly higher air concentrations would likely overestimate the BZ:GA ratios for the actinide-only 

workrooms at ANL-W. Summaries of the air concentrations for the ANL-W actinide-only areas and 

applicable periods are provided in Section 7.2. 

Figure 6-1. BZ:GA ratio by BZ air concentration at 

the NUMEC plutonium laboratory (Caldwell, Potter, 

and Schnell 1967). 
Document No. ORAUT-RPRT-0089 Revision No. 00 Effective Date: 04/19/2022 Page 31 of 72

Figure 6-2. BZ:GA ratio by BZ air concentration at Windscale Works 

(Brunskill and Holt 1967).

6.2 STUDIES REFERENCED BY SCA-TR-2016-SEC009

SCA-TR-2016-SEC009 (SC&A and Salient 2016) made recommendations based on the information in 

the two studies Brunskill and Holt (1967) and Caldwell, Potter, and Schnell (1967). However, SCATR-2016-SEC009 does not provide a basis for why those two studies are applicable to the actinideonly areas at ANL-W. 

6.2.1 Caldwell, Potter, and Schnell Study

The Caldwell, Potter, and Schnell (1967) study was performed at the Nuclear Materials and 

Equipment Corporation (NUMEC) uranium plant and plutonium laboratory. Based on ORAUT-TKBS0041, Site Profile for Nuclear Materials and Equipment Corporation, Apollo and Parks Township, 

Pennsylvania, the uranium plant was at the NUMEC Apollo Site and the plutonium laboratory was at 

the Parks Township Site (ORAUT 2017). Additional information about the NUMEC uranium and 

plutonium facilities are summarized in the sections below. Note that, when possible, these summaries 

only provide information that is relevant in time to the 1967 study (i.e., around the years of 1966 and 

1967). 

6.2.1.1 Description of the NUMEC Apollo Site

By 1963, the majority of the Apollo Site was dedicated to the production of low-enriched uranium 

(LEU) and highly enriched uranium (HEU) fuels. The major activities at the Apollo Site included 

(1) the conversion of LEU hexafluoride (UF6) to uranium oxide (UO2) for use in light-water-moderated 

reactors, (2) the production of HEU nuclear fuel for use in the naval reactors program, and (3) the 

processing of unirradiated uranium scrap (including LEU and HEU) from the U.S. Atomic Energy 

Commission (AEC) in the 1960s (ORAUT 2017). 

Production of UO3 from UF6 began with UF6 being converted to UO3 in the Chemical Conversion

Room. The UF6 gas was hydrolyzed to an aqueous solution of UO2F2 and HF. This solution was
Document No. ORAUT-RPRT-0089 Revision No. 00 Effective Date: 04/19/2022 Page 32 of 72

reacted with NH4OH to form a slurry of ammonium diuranate (ADU) [(NH4)2U2O7]. The slurry was

then pumped through a hooded pressure filter. The filter cake was transferred to drying hoods where

the ADU was decomposed to UO3 at a controlled temperature. The UO3 product was transferred in

small polyethylene containers to the Ceramics Fabrication Area for further processing and conversion 

into UO2. By August 1960, the filter cake was being dried by a rotary kiln rather than the earlier fry 

pan method. By June 1961, a calciner had been added for reduction of ADU to U3O8. The dried U3O8

cake was discharged directly from the kiln into a container, eliminating the manual transfer

(ORAUT 2017). 

Reduction of UO3 to UO2 was performed in the Ceramics Fabrication Area. UO2 product was

transferred to the Ceramic Laboratory for additional fabrication. The UO3 was reduced to UO2 in a 

rotary kiln rather than the reduction furnace that had been used earlier. Ceramics fabrication was 

performed in the Ceramics Fabrication Area where UO2 was hammer-milled in a ventilated enclosure 

and then moved to the blender glovebox where Aerowax was added and the mixture was blended. 

The wax-UO2 mixture was then pressed into a cake in a Drake press. The cake was placed in a 

glovebox where it was granulated by hand with screens to give the desired particle size. The UO2

was loaded into shallow metal pans called “firing boats” and sintered. Sintered UO2 was classified, 

weighed, and packaged (ORAUT 2017). 

Uranium scrap was dissolved in two designated areas, CRP-2 and CRP-3. The product solutions 

from the dissolution methods were processed to generate insoluble UF4 and ultimately converted into 

U3O8 or UO2 as the final product (ORAUT 2017). 

Table 6-2 provides a summary of the maximum possession limits for special nuclear material (SNM) 

for the Apollo Site during the 1972 timeframe. Note that these are the maximum possession limits, 

which could be much greater than the actual quantities that were used. The typical amounts of 

uranium in use in an area ranged from milligrams to hundreds of kilograms (ORAUT 2017). 

Table 6-2. SNM possession limits for NUMEC Apollo Site (ORAUT 2017).

Areas Source/chemical or physical form

Maximum 

possession

Processing areas, laboratories, and vaults U-235 enrichment >5% 5,000 kg

Processing areas, laboratories, and vaults U-235 enrichment ≤5% 75,000 kg

Processing areas, laboratories, and vaults Plutonium as fully clad or encapsulated 

material

500 kg

Mass Spectrometry Laboratory Uranium in any enrichment 350 kg

Mass Spectrometry Laboratory Plutonium in any form 0.5 g

Low-level radioactive waste storage areas U-235 within fenced area in approved 

storage containers

35 g

Low-level radioactive waste storage areas U-235 in buildings meeting safeguards and 

security requirements

50 kg

Nuclear Decontamination Corporation Any byproduct material 20 mCi

Nuclear Decontamination Corporation Any source material 20 g

Nuclear Decontamination Corporation Any SNM 20 mCi

6.2.1.2 Description of the NUMEC Parks Township Site

By 1967, the Parks Township Site primarily consisted of two processing buildings, Building A (the 

Plutonium Plant) and Building B (the Metals Complex). The original Building A was 20,000 ft2

. From 

1961 through 1970, a major expansion of Building A was completed in five separate expansions that 

eventually increased the footprint of Building A to 61,000 ft2 (ORAUT 2017). Caldwell, Potter, and 

Schnell (1967) indicated that the plutonium laboratory was 20,000 ft2

. Based on that, Caldwell, Potter, 

and Schnell (1967) appear to only have used Building A for their study. Therefore, only the applicable 
Document No. ORAUT-RPRT-0089 Revision No. 00 Effective Date: 04/19/2022 Page 33 of 72

information for the portions of Building A that existed before and during the study (i.e., before 1968) 

are provided in this section. 

Throughout its history, Building A has been referred to as the Plutonium Plant, Plutonium Laboratory, 

Plutonium Facility, Plutonium Processing Facility, Plutonium Building, and the NUMEC Advanced 

Material Center. The original portion of Building A was designed as a plutonium laboratory to perform 

research and development that led to plutonium-based products (ORAUT 2017). The main 

workrooms in Building A were called fabrication areas, which were also referred to as Fabs or FABs 

for short. Based on the information in a 1979 NRC licensing revision for SNM-414, FAB-1 through 

FAB-5 were the only plutonium production areas present at the time of the 1967 air sampling study. 

FAB-6 through FAB-9 were not added to the Plutonium Plant until 1968 and later (Austin 1979). 

The initial functions of the Parks Township facilities included plutonium fuel fabrication, HEU fuel 

preparation, and the production of zirconium-hafnium bars. That work was performed under the AEC 

and later under NRC License SNM-414, which allowed the handling of plutonium already on the site. 

Before 1968, various plutonium conversion processes were performed in FAB-1 along with some 

repair and maintenance work on contaminated equipment. In FAB-2, plutonium fuel fabrication work 

was performed for the ZPR-III and ZPPR. Before 1968, FAB-3 was primarily used for metallography 

and some unspecified manufacturing operations that involved plutonium. The radioactive materials 

handled in FAB-1 through FAB-3 included plutonium metal, plutonium nitrates, plutonium oxalates, 

plutonium oxide (PuO2), plutonium-beryllium compacted powder, americium metal, and depleted 

uranium oxide (UO2) (ORAUT 2017). The chemical compound formulas were not provided for the 

plutonium nitrates and plutonium oxalates because they vary depending on the valence state of the 

plutonium. 

Before 1968, FAB-4 was primarily used for alpha, beta, gamma, and neutron source fabrication 

involving various or unspecified forms of 7

Be, 60Co, 192Ir, 210Po, plutonium, and americium. FAB-5 was 

used for all forms of plutonium scrap recovery and analytical laboratory work (ORAUT 2017). 

In the late 1960s, the Parks Township Site made an unspecified number of plutonium fuel wafers for 

ZPR-III and 11,500 plutonium fuel plates for the ZPPR, which were both at ANL-W. An additional 

700 plutonium fuel plates were manufactured for ZPPR, but not until after 1967. The ZPPR fuel 

plates were plutonium-uranium-molybdenum ternary alloy plates encapsulated in stainless steel using 

fuel-grade plutonium (11.5% 240Pu) and depleted uranium. From 1961 through 1969, the SNM 

possession limit for Building A was any combination of plutonium and 235U up to 400 kg (ORAUT 

2017). Based on Schedules A and C in the AEC plutonium supply contract with NUMEC, NUMEC 

was committed to processing between 100 and 160 kg of plutonium during most months for just the 

ZPPR fuel production effort (Shapiro 1967). 

6.2.1.3 Applicability of NUMEC Sites to ANL-W Actinide-Only Areas

Caldwell, Potter, and Schnell (1967) indicated that the uranium plant was 40,000 ft2 and the plutonium 

laboratory was 20,000 ft2

. Further research indicates that the 20,000 ft2 and 40,000 ft2 values likely 

represent the footprints for the entire processing areas during the time of the study versus the 

individual workrooms. As indicated above, the 20,000 ft2 appears specifically to be the footprint for 

Building A at the Parks Township Site during that time. That further research also confirmed that the 

majority of the individual workroom sizes for the uranium and plutonium processing areas at the 

NUMEC facilities were significantly larger than the actinide-only areas at ANL-W. Table 6-3 

summarizes the uranium processing room sizes at the Apollo Site, and Table 6-4 summarizes the 

plutonium processing room sizes at the Parks Township Site (ORAUT 2021e). 
Document No. ORAUT-RPRT-0089 Revision No. 00 Effective Date: 04/19/2022 Page 34 of 72

Table 6-3. NUMEC Apollo facility room sizes (ORAUT 2021e). 

Process areaa

Total area

(ft2)

Total area

(m2)

CP-1 6,652 618.0

CRP-1 2,500 232.3

PC-2 785 72.9

PC-3 1,329 123.4

CF-1 1,890 175.6

CRP-4 1,470 136.6

CRP-2/CP-2/CRP-3 (2nd floor)b 7,347 682.6

CP-1 (2nd floor) 1,920 178.4

a. All processing areas are on the first floor, unless indicated otherwise.

b. No drawings depicting walls between these three processing areas were found. 

Therefore, the available information indicated that they shared the same air 

space.

Table 6-4. NUMEC Parks Township facility room sizes (ORAUT 

2021e). 

Process area

Total area

(ft2)

Total area

(m2)

FAB-1 4,951 459.9

FAB-2 1,934 179.6

FAB-3 3,355 311.7

FAB-4 806 74.9

FAB-5 2,182 202.7

As indicated in Table 6-1, the workroom sizes for the actinide-only areas at ANL-W were generally 

92.3 m2 (994 ft2

) or less. The ZPPR Workroom at 154.4 m2 (1,662 ft2) and potentially the ITF Low Bay 

Area were the only exceptions to that, but the air concentrations in those two areas were low. As 

indicated in Section 6.1.2, BZ:GA ratios tend to be closer to unity when the air concentrations are low, 

even for larger workrooms. 

Based on the available information for the NUMEC workrooms involved in the Caldwell, Potter, and 

Schnell (1967) study, the data from that study is not applicable to the ANL-W actinide-only areas for 

the following reasons: 

1. Most of the NUMEC workrooms were much larger than the ANL-W actinide-only areas. 

2. The majority of reported air concentrations for the NUMEC workrooms were significantly 

higher than the air concentrations in the ANL-W actinide-only areas. 

3. The processes at NUMEC involved relatively large quantities of dispersible uranium and 

plutonium; whereas, the processes at ANL-W worked with much smaller quantities. Because 

of that, much greater air concentration gradients were possible at NUMEC than at ANL-W. 

Note that at both NUMEC and ANL-W used engineering controls (e.g., gloveboxes, ventilation 

hood, etc...) to protect the workers from the dispersible forms of plutonium. Therefore, 

airborne plutonium in those workrooms was normally only possible when there was some 

degree of failure in those engineering controls.

4. The particle size distributions are unknown for the NUMEC workrooms. Therefore, it is 

possible that the air samples were influenced by nonrespirable radioactive particles. 
Document No. ORAUT-RPRT-0089 Revision No. 00 Effective Date: 04/19/2022 Page 35 of 72

6.2.2 Brunskill and Holt Study

The Brunskill and Holt (1967) study was performed at the plutonium and uranium plants at the 

Windscale and Springfields Works in the United Kingdom. The study was performed in a number of 

plant and laboratory areas at those two sites. At the time of the study, both sites were under the 

United Kingdom Atomic Energy Authority. In 1981, Windscale was renamed Sellafield. 

In general, there is insufficient information about the processes and workrooms at the Windscale and 

Springfields Works to assess the applicability of the Brunskill and Holt (1967) data to the ANL-W 

actinide-only areas. Sections 6.2.2.1 and 6.2.2.2 provide summaries of the available information on 

the processes and workrooms at those two sites. 

6.2.2.1 Description of Windscale Works Site

The areas of operation for the Brunskill and Holt (1967) study included:

1. A plant for the recovery of plutonium from waste material.

2. A primary separation plant for the extraction of plutonium from irradiated fuel elements.

3. A plutonium finishing plant where pure metal was produced from plutonium nitrate solution.

4. A decontamination center.

5. An active laundry.

6. A facility employing “face standing” or “island” gloveboxes in an open laboratory and used for 

the manufacture of “plutonium oxide/uranium oxide ceramic fuel” (a.k.a. mixed oxide or MOX

fuel) for reactors.

The airborne contamination at Windscale was entirely due to plutonium in one form or another 

(Brunskill and Holt 1967). 

6.2.2.2 Description of Springfields Works Site

The areas of operation for the Brunskill and Holt (1967) study included:

1. A plant for the production of uranyl fluoride (UO2F2) from gaseous uranium hexafluoride (UF6).

2. A laboratory for the preparation of samples of uranium ore concentrates for chemical analysis.

3. A slag-crushing operation in which magnesium fluoride slag from the reduction of uranium 

tetrafluoride (UF4) to metal was crushed in preparation for further treatment.

4. The open shop floor, which was used for the annealing and inspection of uranium metal rods.

5. The UF4 Reduction Plant where the particular operation considered was the weighing of 

UF4-magnesium pellets after delivery from the pelleting press. 

The airborne contamination at Springfields was entirely due to uranium and its compounds (Brunskill 

and Holt 1967). 
Document No. ORAUT-RPRT-0089 Revision No. 00 Effective Date: 04/19/2022 Page 36 of 72

6.2.2.3 Applicability of United Kingdom Facilities to ANL-W Actinide-Only Areas

Based on the limited information for the air sampling locations at the Windscale and Springfields 

Works, the workrooms evaluated in Brunskill and Holt (1967) were full-scale plutonium and uranium 

production and processing areas. One of the plants at Windscale covered an area of about 300 by 

300 ft (90,000 ft2 or 8,361.3 m2

) (Brunskill and Holt 1967). No other size information was found for the 

workrooms that were evaluated at the Windscale and Springfields Works. However, their room sizes 

are likely comparable to large-scale plutonium and uranium processing facilities at DOE sites like 

Hanford, Feed Materials Production Center, Savannah River Site, and Weldon Spring Plant. Based 

on the available information, the workrooms at the Windscale and Springfields Works would have had 

large quantities of dispersible plutonium and uranium present; the ANL-W actinide-only areas did not. 

In general, there is insufficient information about the processes and workrooms at the Windscale and 

Springfields works to assess the applicability of the data in the Brunskill and Holt (1967) study to the 

actinide-only areas at ANL-W. In addition, there are significant issues with the Brunskill and Holt 

(1967) study that make all of its BZ:GA ratio information questionable. 

A review of Brunskill and Holt (1967) indicates that the terms “static samplers” and “static 

environmental samplers” are used interchangeably and therefore mean the same thing. The terms 

“static samplers” and “static environmental samplers” have also been interpreted to represent GA air 

samplers. The terms “uranium areas” and “uranium-active areas” are also used interchangeably 

throughout the study, and are assumed to mean the same thing. 

Brunskill and Holt (1967) contains contradicting statements regarding the correlation of BZ and GA air 

sample data. The following is stated in the first two sentences in Section 6.1:

An analysis of results from all uranium and plutonium areas investigated has indicated 

quite clearly that static environmental samplers are in themselves incapable of indicating 

the true conditions in the breathing zone of operators. Results from uranium-active areas 

have generally shown good correlation between the data from static and personal air 

samplers.

The following is stated in Section 4.1:

In uranium-active areas there is a significant correlation between the personnel and 

static data at the 99% confidence level. In plutonium-active areas there is virtually no 

positive correlation.

The following is stated in Section 6.2:

In the plutonium-active areas the general absence of correlation between the personal 

and static air-sampling data rules out any possibility of control using static sampler data.

The following is stated in Section 6.3:

In the plutonium-active areas, under normal conditions of operation, there is evidence 

that the nature of the airborne activity in the region of the static samplers is basically 

similar to that in the breathing zone.

Given the above, Brunskill and Holt (1967) had significant difficulty determining when there was not a 

good correlation between BZ and GA air sample results. Based on the majority of the statements, the 

correlation between the BZ and GA air sample results in the uranium areas was good. However, 

there seems to be confusion when it comes to the plutonium facilities. Based on the statement in 
Document No. ORAUT-RPRT-0089 Revision No. 00 Effective Date: 04/19/2022 Page 37 of 72

Section 6.3 of that study, the correlation between the BZ and GA air sample results in the plutonium 

areas was good under normal conditions of operation; whereas, statements in Section 4.1 and 6.1 of 

that study indicated there was virtually no correlation. 

In 1970, the results for another study that was performed at the Springfields Works were published. 

The purpose of that study was to investigate why air contamination measured with PAS devices is 

usually greater than indicated by fixed samplers in the working area. That study concluded that the 

dust released from protective clothing can contribute appreciably to the air contamination measured 

by PAS (Butterworth and Donoghue 1970). 

In 1988, a review of the experience with plutonium exposure assessment methodologies at the 

Windscale/Sellafield site was completed (Strong 1988). The Discussion of Results section included 

the following statements (Strong 1988, pp.124-125): 

i there is no obvious correlation between PAS data, urine sample data, and faecal 

sample data

ii biological sample results generally imply intakes smaller than indicated by personal 

air sampler... 

a the personal air sampler behaves as a statistical sampling device when operated in 

an environment having only a few to a few tens of particles per m3

, 

b trivial levels of surface contamination, once transported to the air sampler, can 

appear to indicate significant inhalation... 

However, as is evident from the data presented, there is no clear relationship between 

significant PAS results and the evidence from biological sampling. Indeed this is a 

source of confusion.

These statements make it clear that through at least some of 1988 the Safety and Medical Services 

staff at Windscale/Sellafield were still unaware of the of the potentially significant sources of error with 

personal air sampling and were thoroughly confused as to why their bioassay data generally yielded 

intake estimates lower than the PAS data. The statements quoted above indicate that nonrespirable 

dominant particles were likely biasing the Windscale/Sellafield PAS results high, which would also 

cause the intakes based on those results to be overestimated. This is supported by the statement 

that the biological sample results generally imply intakes smaller than indicated by the PAS results. 

Based on the information in Butterworth and Donoghue (1970) and Strong (1988), the BZ:GA air 

concentration ratios in Brunskill and Holt (1967) are likely biased high because of sampling 

resuspended from clothing and the presence of nonrespirable dominant particles on the samples. 

Therefore, the ORAU Team does not consider the Brunskill and Holt (1967) study to be a credible 

source of BZ:GA ratio information for any site.

6.2.3 Summary of Issues with Studies Referenced by SCA-TR-2016-SEC009

The determinations made from these two 1967 studies rest heavily on BZ air sample data that was 

collected using lapel air samplers (Brunskill and Holt 1967; Caldwell, Potter, and Schnell 1967). Both 

of these studies were completed before some of the potentially significant sources of error with lapel 

sampling were known. 

In the years after those two studies were completed, subsequent air sampling studies indicated that a 

major cause of the discrepancies between GA or BZ air samples from SASs and BZ air samples from 
Document No. ORAUT-RPRT-0089 Revision No. 00 Effective Date: 04/19/2022 Page 38 of 72

lapel air samplers are potentially due to the lapel air samplers sampling the contamination 

resuspended from a worker’s protective clothing (Butterworth and Donoghue 1970; Cohen et al. 1982; 

Martinelli et al. 1983; Cohen, Harley, and Lippmann 1984; Bohne and Cohen 1985). In 1967, when 

that was unknown, steps were not likely taken during those two studies to prevent or minimize the 

sampling of contamination from the workers’ clothing by the lapel air samplers. 

In addition, subsequent air sampling studies indicated that the presence of larger nonrespirable size 

particles (i.e., >10 μm) and/or dominant radioactive particles on the filter media for a BZ air sample 

are also likely causes for the discrepancies between GA and BZ air samples. Because larger 

nonrespirable size particles do not travel as far as smaller size particles, BZ air concentrations can be 

higher than GA air concentrations due to the presence of those larger particles only being airborne 

within the BZ. The presence of nonrespirable radioactive particles in an air sample can result in the 

overestimate of a worker’s intake, because only the respirable fraction of the airborne radioactivity 

needs to be accounted for and used to estimate internal exposures. When only smaller respirable 

particles are in the air, the discrepancies between GA and personal BZ air sample results tend to be 

smaller because the respirable particles are more readily spread throughout the room (Alvarez, 

Bennett, and Davidson 1994). Further, much of the disparity between air sample data and bioassay 

data is likely due to the nonrespirable fraction of airborne radioactivity in the air samples (Fischoff 

1963; Alvarez, Bennett, and Davidson 1994). When the nonrespirable fraction of airborne 

radioactivity is significant, air sample data should overestimate a worker’s intake. When that is the 

case, intakes based on air sample data often do not correlate well with bioassay data (Jones et al. 

1983; NRC 1984). However, when the air samplers can discriminate against the nonrespirable 

particles, the air sample data and bioassay data correlate much better (Fischoff 1963). 

The presence of dominant radioactive particles on the filter media for an air sample is also likely a 

cause of significant temporal and spatial variations between GA and BZ air concentrations. Even 

though this potential issue with BZ air samples was known before 1967, neither of the studies 

referenced in SCA-TR-2016-SEC009 appear to have considered this as a potential cause for the 

discrepancies between their GA and BZ air sample results. A 1967 paper on the development of air 

sampling technology by the United Kingdom Atomic Energy Research Establishment (UKAERE) 

stated that, in more than half of the air samples from PASs that were examined, much of the activity 

could be attributed to one or a few active particles (Lister 1967). A 1983 study at the UKAERE 

facilities indicated that about half of the significant air sample results from their PAS devices were 

attributable to a single hot particle on the filter media. In the cases that study investigated, there was 

little to no correlation between the elevated sample results and early fecal excretion data (Jones et al. 

1983). 

Because of the reasons presented above, the ORAU Team does not recommend using the 

information in Brunskill and Holt (1967) and Caldwell, Potter, and Schnell (1967) for the actinide-only 

areas at ANL-W or any other site. 

6.3 STUDIES EVALUATED BY ORAU TEAM

The ORAU Team has evaluated five air sampling studies that it considers to be more applicable 

to the actinide-only areas at ANL-W than the studies cited in SCA-TR-2016-SEC009. Those 

studies were much more recent than the ones cited in SCA-TR-2016-SEC009. The results of the 

ORAU Team evaluation have been documented in ORAUT-RPRT-0097 (ORAUT 2021d). The 

studies evaluated for ORAUT-RPRT-0097 were limited to the air sampling studies that were 

completed in small workrooms or the mockups of small workrooms, which is more consistent with 

the sizes of the actinide-only workrooms at ANL-W. 
Document No. ORAUT-RPRT-0089 Revision No. 00 Effective Date: 04/19/2022 Page 39 of 72

6.4 RECOMMENDED RESOLUTION TO THE LACK OF PARITY ISSUE

To address the potential lack of parity issue between BZ and GA air sampling results, the ORAU 

Team recommends using the recommendations and BZ:GA ratio information in ORAUT-RPRT-0097

(ORAUT 2021d). 

ORAUT-RPRT-0097 recommends adjusting the GA air concentrations when the median of the 

BZ:GA ratio distribution becomes significantly greater than 1 or the GSD becomes large. Based 

on the evaluations in ORAUT-RPRT-0097, the GA air concentrations in most small workrooms 

should be adjusted to make them equivalent to BZ air concentrations (ORAUT 2021d). 

6.4.1 Justifying Use of ORAUT-RPRT-0097 for ANL-W Actinide-Only Areas

As indicated in ORAUT-RPRT-0097, the use of the BZ:GA ratio information in that report should be 

justified on a case-by case basis (ORAUT 2021d). The following discussion is the justification for 

using the ORAUT-RPRT-0097 BZ:GA ratio information for the actinide-only areas at ANL-W. 

As indicated in Section 11.0 of ORAUT-RPRT-0097, there are five key parameters that should be 

considered when justifying the application of BZ:GA ratios from other locations to the location of 

interest. Those parameters include (1) room size, (2) particle size distribution for the airborne 

radioactivity, (3) ventilation rate for the room, (4) room complexity, and (5) the presence of dominant 

particles (ORAUT 2021d). The following list addresses each of those five key parameters and 

provides the reasons why the ORAU Team considers the BZ:GA ratio information in ORAUT-RPRT0097 to be applicable to the actinide-only areas at ANL-W: 

1. Room size. The room sizes in the ORAUT-RPRT-0097 studies ranged from 24.0 to 105.0 m2

(258 to 1,130 ft2

), making them more comparable to the room sizes for the actinide-only areas 

at ANL-W (ORAUT 2019, ORAUT 2021d). As indicated in Table 6-1, the room sizes for the 

actinide-only areas at ANL-W were generally 92.3 m2 (994 ft2

) or less. The ZPPR Workroom 

was the only confirmed exception to that at 154.4 m2 (1,662 ft2

), but airborne releases at ZPPR 

were relatively rare. Because the room size information for the Low Bay Area at the ITF is 

unknown, the Low Bay Area might be another exception to that. Given that the ITF building 

had a total square footage of 320 m2 (3,455 ft2

), and given that the Low Bay Area likely 

occupied less than half of the ITF, the size of the Low Bay Area was likely less than 160 m2

(1,727 ft2

). In addition, only six air samples from the ITF Low Bay Area were used for 

unmonitored uranium intake calculations and all six had gross alpha radioactivity results that 

were <10% of the MPC (ORAUT 2018b). As indicated in Section 6.1.2, BZ:GA ratios tend to 

be closer to unity when the air concentrations are low, even for larger workrooms. Therefore, 

the ORAUT-RPRT-0097 BZ:GA ratio information is still applicable to the larger workroom at 

ZPPR and the potentially larger workroom at the ITF. 

2. Particle size distribution. The particle size distributions were known for all five of the ORAUTRPRT-0097 studies, and the majority of the particles were in the respirable size range

(ORAUT 2021d). Therefore, the BZ air sample results from those studies are not significantly 

biased high by nonrespirable particles. There was no indication of any particle sizing 

measurements being performed on the airborne particles in the ANL-W actinide-only areas, 

and there was no indication that the ANL-W air samplers were capable of discriminating 

against nonrespirable particles. Therefore, the ANL-W air sample results are likely biased 

high to some unknown degree because those air samples likely contained some nonrespirable 

radioactive particles. However, the presence of any nonrespirable particles on the ANL-W air 

samples would result in overestimates of the unmonitored actinide intakes, which is favorable 

to claimants. 
Document No. ORAUT-RPRT-0089 Revision No. 00 Effective Date: 04/19/2022 Page 40 of 72

3. Ventilation rate. ORAUT-RPRT-0097 considers the BZ:GA ratio information in it to be 

applicable to rooms with all but the most extreme air change rates. The workrooms in the 

studies evaluated for ORAUT-RPRT-0097 had air change rates that ranged from 6 AC/hr to 

90 AC/hr (ORAUT 2021d). Even though the air change rates for the actinide-only areas at 

ANL-W were not found, it is doubtful that any of the air change rates in those areas were 

significantly greater than 90 AC/hr. 

4. Room complexity. With the exception of the Charuau (1987) study evaluated in ORAUTRPRT-0097 (ORAUT 2021d), the complexity of the rooms and ventilation flow patterns 

evaluated in these studies were comparable to the actinide-only areas at ANL-W. 

Attachment B provides a compilation of the available floorplans for the actinide-only areas at 

ANL-W, some of which depict equipment and process locations for evaluating a room's 

complexity. Even though the room evaluated in the Charuau (1987) study had much more 

complex ventilation flow patterns than the ones at ANL-W, the use of the Charuau (1987) 

study data would likely result in a reasonable overadjustment of the ANL-W air sample data. 

5. Dominant particles. As indicated in ORAUT-RPRT-0097, dominant particles were not present 

in the workrooms associated with the evaluated studies. For ORAUT-RPRT-0097, studies 

were excluded from being evaluated if there was any indication of the presence of dominant 

particles (ORAUT 2021d). Additionally, after several trips to capture data at the site, no 

records were identified that would have indicated that airborne dominant particles were 

present in the actinide-only areas at ANL-W. Therefore, the BZ:GA ratio information and air 

sample data being used for the unmonitored actinide intakes in this report was unaffected by 

dominant particles.

6.4.2 Applicable Scenarios from ORAUT-RPRT-0097

ORAUT-RPRT-0097 identified four common air sampling scenarios for small workrooms and provided 

BZ:GA ratio distributions that should be applied to the GA air concentrations to make them equivalent 

to BZ air concentrations and to account for the increased uncertainty in those air concentrations. The 

following are the two primary scenarios that were evaluated in ORAUT-RPRT-0097 (ORAUT 2021d): 

• Scenario 1. Assumes the worker (BZ location) was always located at the same X,Y

coordinates in the room as the release point (i.e., colocated worker and release point).

• Scenario 2. Assumes the worker (BZ location) was not necessarily located at the same X,Y

coordinates in the room as the release point (i.e., variable worker location relative to the 

release point).

As indicated in ORAUT-RPRT-0097, Scenario 1 represents what is typically the worst-case scenario 

because it normally yields the highest BZ:GA ratios. In most instances, this scenario only applies to 

acute exposures because workers often move around and are not always located at the release point. 

Scenario 2 represents radiological processing areas that have multiple workstations and multiple 

workers moving around in each room rather than one worker being at a single stationary location for 

an entire work shift. For Scenario 2, it was assumed that the potential BZ locations had the same 

probability of being anywhere in the room (ORAUT 2021d). 

Because the generated data for Scenario 1 also indicated there was a significant difference between 

the BZ:GA ratios when the workroom is generally open in the middle of the room rather than when the 

workroom has significant obstructions in the middle of the room, the Scenario 1 BZ:GA ratio 

distributions were divided into two subgroups and reevaluated. The first subgroup represented the 

scenario when the workroom is generally open in the middle of the room, and the second represented

the scenario when the workroom has significant obstructions in the middle of the room. For 
Document No. ORAUT-RPRT-0089 Revision No. 00 Effective Date: 04/19/2022 Page 41 of 72

Scenario 2, the generated data indicated that significant obstructions in the middle of the room had 

little effect on the BZ:GA ratio distribution. Therefore, no subgroups were evaluated for Scenario 2 for 

ORAUT-RPRT-0097 (ORAUT 2021d). 

The BZ:GA ratio distributions associated with those scenarios are presented in Table 6-5. The 

information in Table 6-5 comes from Table 10-1 in ORAUT-RPRT-0097 (ORAUT 2021d). 

Table 6-5. Summary of the combined BZ:GA ratio distributions (ORAUT 2021d).

Scenario

BZ:GA ratio 

GM

BZ:GA ratio 

GSD

Scenario 1 2.95 5.03

Scenario 1 – open in the middle 1.39 3.27

Scenario 1 – obstructions in the middle 12.0 4.11

Scenario 2 1.08 4.02

Section 10 of ORAUT-RPRT-0097 also provides the following guidance (ORAUT 2021d): 

For acute exposure scenarios, the decision to use the results from Scenario 1 or 

Scenario 2 will be dependent on whether or not the worker was present at the release 

locations. Because it is unlikely that a worker would be located at the release location 

for every release event, the results from Scenario 1 would normally just be applicable to 

acute exposure scenarios, such as when the worker was at the release location during 

a radiological accident or excursion. Therefore, the results from Scenario 2 would 

usually be the more appropriate BZ:GA ratio distribution to use for assessing chronic 

exposure scenarios. However, in the rare instances when a chronic exposure scenario 

is more consistent with the description for Scenario 1, the results from Scenario 1 should 

be used.

For the unmonitored intakes in the actinide-only areas at ANL-W, the BZ:GA ratio distribution 

associated with Scenario 2 was determined to be the most appropriate BZ:GA ratio distribution for the 

following reasons: 

1. The unmonitored actinide intakes will be assessed as chronic exposures; 

2. Most of the actinide-only workrooms included multiple workstations; and

3. Based on the work being performed in the actinide-only areas, it is unlikely that workers would 

have spent the majority of their workdays at a single workstation or location. This is especially 

true given the intermittent nature of most of the work in those areas. In most situations, those 

workers would have moved between multiple workstations during their workdays, including 

moving in and out of those workrooms. 

Therefore, a BZ:GA ratio distribution with a GM of 1.08 and a GSD of 4.02 will be applied to the 

air concentrations being used to calculate the potential unmonitored actinide intakes in this 

report. The application of that BZ:GA ratio distribution should make the ANL-W actinide-only 

area air concentrations equivalent to BZ air concentrations and should adequately account for 

the increased uncertainty in those air concentrations.

7.0 UNMONITORED ACTINIDE INTAKES

This section describes the approaches used to calculate the unmonitored actinide intakes for 

ANL-W, summarizes the air concentrations used for those intake calculations, summarizes the 

resulting intake values, and describes how some of the application issues will be addressed. 
Document No. ORAUT-RPRT-0089 Revision No. 00 Effective Date: 04/19/2022 Page 42 of 72

7.1 GENERAL APPROACHES USED FOR INTAKE CALCULATIONS

As indicated in ORAUT-RPRT-0097, the GA air concentrations in small workrooms should be 

adjusted using a BZ:GA ratio distribution to make them equivalent to the BZ air concentrations 

(ORAUT 2021d). Because the air samples at ANL-W were predominately collected using GA air 

samplers, the bounding and reported air concentrations needed to be converted into equivalent BZ air 

concentrations for the intake calculations. 

The following assumptions and parameters were used for the intake calculations: 

• Daily exposure time was 8 hours per workday; 

• 5 days worked per week; 

• 50 weeks worked per year; 

• 250 workdays per year (DWrkYr), based on prior assumptions; and

• 365 days per year (DCalYr) represents the average number of calendar days per year. 

The 8 hr/d, 5 d/wk, 50 wk/yr is equivalent to a 2,000-hr exposure year. Even though some ANL-W 

workers might have worked more than 8 hours per workday, it is unlikely that they were in the 

actinide-only areas for more than 8 hours per workday, due to breaks and the intermittent nature of 

the work in the actinide-only workrooms. Table 6-1 lists the workrooms where actinide-only 

exposures were possible at ANL-W, and all of those are limited to specific workrooms in specific 

buildings. They also represent a very small fraction of the ANL-W site. Because of that, any time that 

a worker spent in the hallways or other rooms would be time that a worker was not in an actinide-only 

workroom. Therefore, on average, the potential for unmonitored actinide exposures did not likely 

exceed 8 hours per workday. 

Based on the available information, process operations in the actinide-only areas only took place 

during the Monday through Friday work week. Activities that did not involve process operations did 

take place during some weekends, but those activities were even more intermittent than the process 

operations. Therefore, the 5 days worked per week assumption is reasonable for all types of work 

that might have taken place in the actinide-only areas. 

The 50 weeks worked per year assumes that the ANL-W workers only took 2 weeks of vacation each 

year. Because this assumption does not factor in the observed holidays, it is favorable to claimants. 

For the intake periods addressed in this report, the number of observed holidays occurring during the 

workweek typically ranged from 8 to 10 days for a given calendar year. Accounting for the observed 

holidays that occurred during a weekday would reduce the weeks worked assumption by up to 

2 weeks for a total of 48 weeks worked per year. 

On average, there are actually 365.2465 days per calendar year, and 365 days represents a rounded 

value. For how it is used in the equations below, rounding that value to 365 days is favorable to 

claimants. 

The following sections show how the unmonitored actinide intakes were calculated. When a specific 

value was used for one of the equation variables, that value was provided in parentheses at the end 

of the variable definition. For example in Equation 7-5, TDayExp is defined as the “daily exposure period 

(8 hr/day)”. The “(8 hr/day)” part of that definition indicates that 8 hr/day was always used for TDayExp

when using Equation 7-5. 
Document No. ORAUT-RPRT-0089 Revision No. 00 Effective Date: 04/19/2022 Page 43 of 72

7.1.1 Intakes Based on Bounding Air Concentrations

The bounding air concentrations in the SEC-00224 evaluation report were only expressed in terms of 

percent of the applicable air concentration limits (NIOSH 2016). For the intake calculations, the 

following formula was used to convert those bounding air concentrations into maximizing GA air 

concentrations: 

(7-1) ( ) ( ) GA CL

MaxGA

CL P C

CF =

where

CMaxGA = maximizing GA air concentration (pCi/m3) 

CLGA = the applicable air concentration limit value used during the applicable period, as 

measured by a GA sampler (dpm/m3

) 

PCL = percent of the CL maintained by the site during the applicable period (1-100%)

CF = conversion factor (2.22 dpm/pCi × 100 percent/fraction = 222)

Because the air concentrations being used for these calculations represent upper-bounds or 

maximizing air concentrations, the resulting maximizing GA air concentrations were treated as 

constants. To convert the maximizing GA air concentrations to maximizing BZ air concentrations, 

Equations 11-5 and 11-6 in ORAUT-RPRT-0097 were used in conjunction with the BZ:GA ratio 

distribution for Scenario 2 in ORAUT-RPRT-0097 (ORAUT 2021d): 

( ) ( ) (7-2) CBZ MaxBZ MaxGA = = C C BZ:GA

(7-3) GSDBZ = = GSDMaxBZ GSDBZ:GA

where

CBZ = GM of the BZ air concentration distribution (pCi/m3

) 

CMaxBZ = GM of maximizing BZ air concentration distribution (pCi/m3

) 

CMaxGA = maximizing GA air concentration (pCi/m3

) 

BZ:GA = GM of the applicable BZ:GA ratio distribution (1.08)

GSDBZ = GSD of the BZ air concentration distribution

GSDMaxBZ = GSD of the maximizing BZ air concentration distribution

GSDBZ:GA = GSD of the applicable BZ:GA ratio distribution (4.02)

Equations 7-4 and 7-5 were then used for the intake calculations. Because air sampling data can only 

be used to estimate potential intakes due to inhalation, potential intakes due to ingestion were 

estimated based on the recommendations in OCAS-TIB-009, Estimation of Ingestion Intakes (NIOSH 

2004). Equation 7-5 was derived from the information in OCAS-TIB-009. Because the Mode 2 

ingestion intake factor in OCAS-TIB-009 was based on an assumed daily exposure period of 8 hours 

(i.e., a typical workday), the Mode 2 equation was modified to accommodate different exposure 
Document No. ORAUT-RPRT-0089 Revision No. 00 Effective Date: 04/19/2022 Page 44 of 72

periods. Equations 7-5 and 7-6 also contain an adjustment to convert from an intake per workday to 

an intake per calendar day: 

( ) ( )( ) (7-4) D I C BR T

D

WrkYr

Inhal BZ DayExp

CalYr

  =    

where

IInhal = daily inhalation intake (pCi/calendar day)

CBZ = GM of the BZ air concentration distribution (pCi/m3

) 

BR = breathing rate (1.2 m3

/hr)

TDayExp = daily exposure period (8 hr/d) 

DWrkYr = days per work year (250 d/yr)

DCalYr = days per calendar year (365 d/yr)

( ) ( ) (7-5) DayExp WrkYr

Ingest BZ Ingest Ingest

CalYr

2 3

8hr

T D I C M M

D

      =     +        

where

IIngest = daily ingestion intake (pCi/calendar day)

CBZ = GM of the BZ air concentration distribution (pCi/m3

) 

TDayExp = daily exposure period (8 hr/day) 

M2Ingest = Mode 2 ingestion intake factor based on an 8-hour workday (0.0985 m3

) 

M3Ingest = Mode 3 ingestion intake factor based on a 24-hour day (0.1004 m3

) 

DWrkYr = days per work year (250 days/yr)

DCalYr = days per calendar year (365 days/yr)

7.1.2 Intakes Based on Air Sample Data

As indicated in Section 5.2, the evaluated sets of air sample concentrations in the ANL-W actinideonly areas can be represented by lognormal distributions. To convert the lognormally distributed timeweighted GA air concentrations to BZ air concentrations, Equations 7-6 and 7-7 were used in 

conjunction with the BZ:GA ratio distribution for Scenario 2 in ORAUT-RPRT-0097 (ORAUT 2021d). 

Equations 7-6 and 7-7 are based on Equations 11-1 and 11-2 in ORAUT-RPRT-0097 (ORAUT 

2021d). Before using Equation 7-6, the time-weighted GMs for the GA air concentration distributions 

were converted from units of dpm/m3 to pCi/m3

. 

( ( ) ( )) (7-6) CBZ = exp ln ln BZ:GA CGA +

( ) ( ) (7-7) ( ) 2 2

BZ exp ln GA BZ:GA GSD = GSD +ln GSD

where

CBZ = GM of the BZ air concentration distribution (pCi/m3

) 

CGA = time-weighted GM of the GA air concentration distribution (pCi/m3

) 

BZ:GA = GM of the applicable BZ:GA ratio distribution (1.08)

GSDBZ = GSD of the BZ air concentration distribution

GSDGA = GSD of the GA air concentration distribution

GSDBZ:GA = GSD of the applicable BZ:GA ratio distribution (4.02)
Document No. ORAUT-RPRT-0089 Revision No. 00 Effective Date: 04/19/2022 Page 45 of 72

7.1.2.1 Intakes Based on Adjusted Air Sample Data

When the individual air sample results in the dataset were adjusted to address the air sample 

dilution concern, Equations 7-4 and 7-5 above were used, respectively, for the inhalation intake 

(IInhal) and ingestion intake (IIngest) calculations. 

7.1.2.2 Intakes Based on Unadjusted Air Sample Data

When the individual air sample results in the dataset could not be adjusted to address the air sample 

dilution concern, the concern was addressed by modifying the intake calculations. Equation 7-8 is the 

modified inhalation intake calculation that addresses the sample dilution concern when the individual 

air sample results could not be adjusted. For the inhalation intake calculation, no adjustment was 

needed in Equation 7-8 to convert the intake into an intake per calendar day because the worker was 

assumed to be continuously exposed to the applicable air concentration during work and outside of 

work (i.e., for 24 hr/d and 365 d/yr). Note that this calculation is only equivalent to the approach taken 

for the adjusted air sample data when the air concentration is assumed to drop to zero during the 

nonoperational periods and when the daily operational period for the facility is the same as the daily 

exposure period assumed for the workers (i.e., no second or third work shifts at the facility). 

( ) ( )( ) (7-8) Inhal BZ DayExp I = C BR T

where

IInhal = daily inhalation intake (pCi/calendar day)

CBZ = GM of the unadjusted BZ air concentration distribution (pCi/m3

) 

BR = breathing rate (1.2 m3

/hr)

TDayExp = daily exposure period (24 hr/d)

As indicated above, potential intakes due to ingestion need to be estimated based on the 

recommendations in OCAS-TIB-009 (NIOSH 2004). Because OCAS-TIB-009 is not set up to account 

for ingestion intakes for a 24 hr/d exposure scenario and might underestimate the ingestion intakes if 

the BZ air concentration from Equation 7-6 is used, a modified calculation was performed, which 

included an additional calculation step. By adding a step to calculate an equivalent to the adjusted BZ 

air concentration, Equation 7-5 can then be used to calculate the ingestion intake. The equivalent to 

the adjusted BZ air concentration was calculated using Equation 7-9, which yields a BZ air 

concentration that is equivalent to a BZ air concentration that was adjusted to account for the air 

sample dilution concern. 

(7-9)

( )( )

Inhal

AdjBZ

WrkYr

DayExp

CalYr

I C

D BR T

D

=      

where

CAdjBZ = equivalent to the GM for the adjusted BZ air concentration distribution (pCi/m3

) 

IInhal = daily inhalation intake (pCi/d)

BR = breathing rate (1.2 m3

/hr)

TDayExp = daily exposure period (8 hr/d)

DWrkYr = days per work year (250 d/yr)

DCalYr = days per calendar year (365 d/yr)
Document No. ORAUT-RPRT-0089 Revision No. 00 Effective Date: 04/19/2022 Page 46 of 72

By using CAdjBZ for CBZ, Equation 7-5 was then used to calculate the ingestion intake. A daily 

exposure period of 8 hr/d was also used for that calculation. 

7.2 SUMMARY OF AIR CONCENTRATIONS FOR INTAKE CALCULATIONS

Tables 7-1 through 7-4 summarize the air concentrations that were used for the unmonitored actinide 

intake calculations. The air concentrations in those tables are the GA air concentrations before the 

application of the BZ:GA ratio distribution. In all instances, the bounding air concentrations were 

based on 10% of the MPC values that ANL-W was using for those areas. The MPC values being 

used by ANL-W were obtained from the available Air Sample Data sheets for those areas. When 

10% of the MPC might not have been bounding, the available air sample data for those areas were 

compiled and a statistical evaluation was performed. The details and results of that statistical 

evaluation are provided in Attachment A. In all instances, the distributions for the evaluated air 

sample data could be represented by a lognormal distribution.

Table 7-1. EBR-I Complex uranium area air concentrations.

a

Actinide-only area

Applicable

period

Distribution

type

Air

concentrationb

(dpm/m3) GSD

ZPR-III Workroom 01/01/1958–07/31/1961 Constant 7.0 N/A

ZPR-III Workroom 08/01/1961–06/13/1975 Constant 13.2 N/A

a. N/A = not applicable. 

b. When the distribution type is constant (i.e., no distribution), a bounding air concentration was used for the intake 

calculations. Per the SEC-00224 evaluation report, the bounding air concentrations for the uranium areas at the EBR-I 

Complex were 10% of the MPC values that ANL-W was using for those areas.

Table 7-2. EBR-II Complex uranium area air concentrations. 

Actinide-only area

Applicable

period

Distribution

type

Air

concentrationa

(dpm/m3) GSD

Cold-Line Areasb 08/01/1967–12/31/1974 Lognormal 3.37 4.92

Cold-Line Areasb 01/01/1975–06/30/1976 Lognormal 2.28 5.20

a. When the distribution type is lognormal, this air concentration is the GM for that air concentration distribution. 

b. In the EBR-II Complex, the uranium areas for the Cold-Line fuel work were in the FCF and the ITF buildings.

Table 7-3. EBR-II Complex thorium area air concentrations.

Actinide-only area

Applicable

period

Distribution

type

Air

concentrationa

(dpm/m3) GSD

FCF Room 25 08/01/1963–11/30/1967 Lognormal 16.9 2.29

a. When the distribution type is lognormal, this air concentration is the GM for that air concentration distribution.

Table 7-4. EBR-II Complex plutonium area air concentrations.

a

Actinide-only area

Applicable

period

Distribution

type

Air

concentrationb

(dpm/m3) GSD

ZPPR 09/01/1970–07/31/1975 Constant 0.44 N/A

FCF 04/01/1970–04/30/1973 Lognormal 1.35 5.73

a. N/A = not applicable. 

b. When the distribution type is constant (i.e., no distribution), a bounding air concentration was used for the intake 

calculations. Per the SEC-00224 evaluation report, the bounding air concentration for the plutonium area at ZPPR was 

10% of the MPC value that ANL-W was using for that area. When the distribution type is lognormal, this air 

concentration is the GM for that that air concentration distribution.
Document No. ORAUT-RPRT-0089 Revision No. 00 Effective Date: 04/19/2022 Page 47 of 72

7.3 CALCULATED UNMONITORED ACTINIDE INTAKES

Using the approaches described in Section 7.1 and the air concentration values and distributions 

in Section 7.2, the unmonitored actinide intakes were calculated for ANL-W. These intakes were 

calculated using the Excel workbook named “ANL-W - Unmonitored Actinide Intake Calculations” 

(ORAUT 2021f). The following sections describe the results.

7.3.1 EBR-I Complex Intakes

For the EBR-I Complex, uranium was the only type of unmonitored actinide exposure that was 

possible, and the potential unmonitored uranium intakes are presented in Table 7-5. 

Table 7-5. Unmonitored uranium intakes at the EBR-I Complex.

Applicable period

Inhalation 

intake 

(pCi/d)

Ingestion 

intake 

(pCi/d) GSD

01/01/1958–07/31/1961 2.07E+01 4.30E-01 4.02

08/01/1961–06/13/1975 3.91E+01 8.10E-01 4.02

7.3.2 EBR-II Complex Intakes

At the EBR-II Complex, the potential unmonitored actinide exposures included exposures to 

uranium, thorium, and plutonium. Tables 7-6 through 7-9 summarize the potential unmonitored 

actinide intakes at the EBR-II Complex and their applicable periods. 

Table 7-6. Unmonitored uranium intakes at the EBR-II Complex.

Period

Inhalation 

intake 

(pCi/d)

Ingestion 

intake 

(pCi/d) GSD

08/01/1967–12/31/1974 1.08E+01 2.23E-01 8.29

01/01/1975–06/30/1976 3.19E+01 6.62E-01 8.65

Table 7-7. Unmonitored thorium intakes at the EBR-II Complex.

Period

Inhalation 

intake 

(pCi/d)

Ingestion 

intake 

(pCi/d) GSD

08/01/1963–11/30/1967 5.41E+01 1.12E+00 5.05

Table 7-8. Unmonitored plutonium intakes for the EBR-II Complex.

Period

Inhalation 

intake 

(pCi/d)

Ingestion 

intake 

(pCi/d) GSD

04/01/1970–04/30/1973 4.32E+00 8.28E-02 9.32

Table 7-9. Unmonitored plutonium intakes for the ZPPR.

Period

Inhalation 

intake 

(pCi/d)

Ingestion 

intake 

(pCi/d) GSD

09/01/1970–07/31/1975 1.4E+00 2.9E-02 4.02

7.4 APPLICATION OF UNMONITORED ACTINIDE INTAKES

The following sections address some application issues that are unique to the unmonitored 

actinide intakes for ANL-W. 
Document No. ORAUT-RPRT-0089 Revision No. 00 Effective Date: 04/19/2022 Page 48 of 72

7.4.1 Overlapping Periods of Exposure

Given that all workers at the EBR-I and EBR-II Complexes were required to wear dosimeters to 

monitor their external doses, work at each of those complexes can be identified by the location code 

information in their external dosimetry records. Therefore, for periods when a worker had both EBR-I 

and EBR-II Complex dosimeters, only the unmonitored intakes for the complex resulting in the highest 

dose would be assigned for that period. 

At the EBR-II Complex, the actinide-only areas were distinctly separate air spaces, which means 

that a worker could not be simultaneously exposed to thorium, uranium, and plutonium. In 

addition, unmonitored actinide intakes for ZPPR workers and non-ZPPR workers will be handled 

separately. Because ZPPR workers can be identified by the available external dosimetry records 

in the SRDB and because it is unlikely that ZPPR workers would have entered the other actinideonly areas in the EBR-II Complex during the operational periods for the EBR-II fuel production 

work in those areas, only the unmonitored plutonium intakes specific to ZPPR will be assessed 

for the ZPPR workers (i.e., the intakes in Table 7-9). Additionally, because of security 

safeguards for the plutonium, it was unlikely that non-ZPPR workers would have been allowed in 

the ZPPR Workroom while the plutonium fuel was being handled. Therefore, the ZPPR 

plutonium intakes in Table 7-9 do not need to be assessed for non-ZPPR workers. 

For the non-ZPPR workers in the EBR-II Complex, potential unmonitored actinide intakes will be 

assessed based on the intake information in Tables 7-6, 7-7, and 7-8. Because the unmonitored 

actinide intakes are based on gross alpha analysis results and because a non-ZPPR worker in 

the EBR-II Complex could not be simultaneously exposed to all three actinides, only the 

unmonitored actinide intake resulting in the highest dose would be assigned for periods when 

more than one type of actinide intake was possible. 

7.4.2 Uncertainties

To account for uncertainties associated with the unmonitored actinide intakes, each unmonitored 

actinide intake would be assessed as a lognormal distribution with the appropriate GSD value, as 

indicated above in Tables 7-5 through 7-9. 

8.0 CONCLUSIONS

The following sections provide the conclusions for each of the three main purposes of this report. 

8.1 ISSUES RAISED IN SCA-TR-2016-SEC009

SCA-TR-2016-SEC009 indicates a concern that there are factors that caused the ANL-W air 

monitoring data to not be representative of the BZ air concentrations, and it concludes that NIOSH’s 

proposed value of 10% MPC as a bounding value for internal dose assessment lacks credibility. 

However, SCA-TR-2016-SEC009 does not identify factors specific to or applicable to ANL-W that 

could cause those air monitoring results to be unrepresentative of the BZ air concentrations (SC&A 

and Salient 2016). 

The dose reconstruction regulation (42 CFR Part 82) and Section 5.2 of OCAS-IG-002 (NIOSH 2002) 

still allow the use of workplace monitoring data such as BZ air samples, GA air samples, and surface 

contamination surveys to estimate an individual’s internal dose when bioassay data is not adequate. 

However, Section 5.2 of OCAS-IG-002 requires that consideration be given to any factors that could 

create a difference between GA and BZ air concentrations when using GA air sampling data (NIOSH 

2002). After an extensive data capture effort to find information on the ANL-W air monitoring program 

and find air sampling results for the actinide-only areas at ANL-W, those air monitoring records have 
Document No. ORAUT-RPRT-0089 Revision No. 00 Effective Date: 04/19/2022 Page 49 of 72

been reviewed to identify potential factors that could create a difference between the GA and BZ air 

concentrations. There is no information that indicates that there might have been a significant 

difference between the GA and BZ air concentrations in the actinide-only areas at ANL-W. In 

addition, as indicated in Section 6.1, differences between the GA and BZ air concentrations would 

have been minimized by the relatively small size of the air spaces and normally low air concentrations 

associated with the ANL-W actinide-only areas. Therefore, the OCAS-IG-002 requirement to give 

consideration to those potential factors had been met by the SEC-00224 evaluation report (NIOSH

2016). 

However, NIOSH has since decided to investigate the lack of parity issue further and tasked the 

ORAU Team to evaluate that issue for small workrooms. The results of that evaluation have been 

documented in ORAUT-RPRT-0097 (ORAUT 2021d). The studies evaluated for ORAUT-RPRT-0097 

were limited to the air sampling studies that were completed in small workrooms or the mockups of 

small workrooms, which is more consistent with the sizes of the actinide-only workrooms at ANL-W. 

ORAUT-RPRT-0097 recommends adjusting the GA air concentrations when the median of the BZ:GA 

ratio distribution becomes significantly greater than 1 or the GSD becomes large. Based on the 

evaluations in ORAUT-RPRT-0097, the GA air concentrations in most small workrooms should be 

adjusted to make them equivalent to BZ air concentrations. 

During the process of making the adjustments to correct for sample dilution, the ORAU Team 

determined that an air concentration equivalent to 10% MPC would not be bounding for all actinides 

and periods because of a few elevated air sample results for those periods. To address this, the timeweighted mean of the adjusted air concentrations will be used to bound the unmonitored internal 

actinide doses for those actinides and periods (for more details see Attachment A). As indicated in 

Section 5.2, it is still reasonable to bound the potential unmonitored uranium intakes at the EBR-I 

Complex and unmonitored plutonium intakes at ZPPR based on a concentration of 10% of the MPC. 

The following are a few of the key factors that would ensure that the approaches described above and 

in Attachment A would not underestimate the unmonitored thorium, uranium, and plutonium intakes at 

ANL-W: 

1. A significant number of the elevated air sampling results were biased high because ANL-W did 

not allow enough time for all of the short-lived radon and thoron progeny to decay before 

performing the final counts on those air samples. As indicated in Section 4.0, some of the 

reported air concentrations could be as much a 171 times higher than what they actually were 

due to the final counts being performed on the air samples before all of the short-lived radon 

and thoron progeny could decay.

2. The actual occupancy in the actinide-only areas and intermittent nature of the operations in 

those areas are not accounted for. In general, radiological workers do not spend 100% of their 

workday in radiological areas for several reasons (e.g., breaks, lunches, meetings, training). 

At ANL-W, another factor that affects the occupancy is the intermittent nature of much of the 

work. Because the Cold-Line HEU fuel production work and RAS-TREAT sodium-loop 

experiment work with plutonium source terms were very intermittent, potential operational 

exposures to those source terms were also very intermittent. In addition, occupancy in the 

uranium areas (due to HEU) and plutonium areas was likely minimal during the nonoperational 

portions of the workdays due to accountability and security reasons associated with the 

protection of SNM. During all nonoperational periods, all of the HEU and plutonium source 

terms were probably locked up.

3. The respirable fraction of the airborne radioactivity was not accounted for. The unmonitored 

thorium intakes would be the most affected by this factor because all of the thoria was 
Document No. ORAUT-RPRT-0089 Revision No. 00 Effective Date: 04/19/2022 Page 50 of 72

assumed to be respirable even though the majority of the thoria powder being used was likely 

nonrespirable. 

4. A review of the radiological work permits for ANL-W indicated that the site often required 

respiratory protection when there was any potential for internal exposure. Not accounting for 

when a worker wore respirator protection would result in an overestimate of their actinide 

intakes by a factor of 10 or more.

5. The air sample results that were adjusted to account for potential sample dilution were 

adjusted based on the assumption that the concentrations of airborne radioactivity dropped to 

zero during the nonoperational periods. Because the concentrations of airborne radioactivity 

did not likely drop all of the way to zero outside of the normal work shift hours, those 

adjustments likely resulted in an overestimate of the air concentrations for the operational 

periods. 

8.2 ADDITIONAL DETAILS ABOUT UNMONITORED ACTINIDE INTAKE CALCULATIONS

Section 7.0 provides the additional details about the unmonitored actinide intakes calculations that the 

SEC-00224 evaluation report did not go into. The resulting unmonitored actinide intakes are provided 

in Section 7.3. Section 7.4 provides some additional guidance on how the unmonitored actinide 

intakes in Section 7.3 will be applied for ANL-W claims. 

8.3 DEVIATIONS FROM SEC-00224 EVALUATION REPORT

As indicated in Section 1.0, one of the purposes of this report was to provide the basis for revising 

some of the unmonitored actinide intake and internal dose approaches proposed in the SEC-00224

evaluation report. This section provides a summary of those changes.

Rather than using the recommendation in the SEC-00224 evaluation report to bound all of the 

unmonitored intakes by using 10% of the MPC, the actual air concentration distribution information 

was used for four of the air sample datasets. As indicated in Sections 4.2 and 5.2, four of the air 

sample datasets were put through a formal statistical evaluation, because the numbers of elevated air 

sample results indicated that 10% of the MPC may not be bounding. The four air sample datasets 

included the air sample results for the EBR-II Complex thorium and uranium areas and the FCF 

plutonium areas, with the air sample results for the uranium areas being divided into two separate 

datasets. 

The potential unmonitored actinide intakes for ZPPR and non-ZPPR workers in the EBR-II Complex 

are now being handled separately. Previously, it could not be determined which facility a worker in 

the EBR-II Complex worked at. Because of that, it had to be assumed that all EBR-II Complex could 

have received an unmonitored actinide intake in any of the actinide-only areas in the EBR-II Complex. 

As discussed in Section 7.4.1, ZPPR workers can now be identified using the available external 

dosimetry records in the SRDB. When the SEC-00224 evaluation report was issued, the external 

dosimetry records for some of years had not been captured yet. 

As indicated in Section 4.2.2, the need to assess unmonitored uranium intakes at ZPPR has been 

eliminated. Only unmonitored plutonium intakes need to be assessed for ZPPR workers.
Document No. ORAUT-RPRT-0089 Revision No. 00 Effective Date: 04/19/2022 Page 51 of 72

The remaining deviations were either revisions to the unmonitored actinide intake date ranges in the 

SEC-00224 evaluation report or determined start and/or end dates when not provided in that 

evaluation report. The following is a summary of those revised dates.

1. For the unmonitored uranium intakes at the EBR-I Complex, June 13, 1975 was determined to 

be an appropriate end date for those intakes, as discussed in Section 4.1. 

2. For the unmonitored uranium intakes for the Cold-Line areas at the EBR-II Complex, the 

intake end date was reduced from June 1983 to June 30, 1976, as discussed in Section 4.2.2. 

3. For the unmonitored plutonium intakes at ZPPR, no start or end dates were provided in the 

SEC-00224 evaluation report for these potential intakes. As discussed in Section 4.2.3.1, the 

appropriate intake start and end dates were determined to be September 1, 1970, through 

July 31,1975.

4. For the unmonitored plutonium intakes at the FCF, the intake end date was extended from 

December 1972 to April 30, 1973, as discussed in Section 4.2.3.2. 
Document No. ORAUT-RPRT-0089 Revision No. 00 Effective Date: 04/19/2022 Page 52 of 72

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Concentration Calculations,” Rev. 0, spreadsheets, Oak Ridge, Tennessee, October 26. 

[SRDB Ref ID: 168581]

ORAUT (Oak Ridge Associated Universities Team), 2021d, Breathing Zone to General Area Air 

Concentration Ratios in Small Facilities, ORAUT-RPRT-0097, Rev. 00, Oak Ridge, 

Tennessee, March 29. [SRDB Ref ID: 185576]

ORAUT (Oak Ridge Associated Universities Team), 2021e, NUMEC – Room Sizes for the Uranium 

and Plutonium Processing Areas, Oak Ridge, Tennessee, June 18. [SRDB Ref ID: 187123]

ORAUT (Oak Ridge Associated Universities Team), 2021f, “ANL-W - Unmonitored Actinide Intake 

Calculations,” Rev. 0, spreadsheets, Oak Ridge, Tennessee, October 31. [SRDB Ref ID: 

168580]

ORAUT (Oak Ridge Associated Universities Team), 2021g, Statistical Evaluations of Several ANL-W 

Air Sample Datasets That Were Used for ORAUT-RPRT-0089, Rev. 0, Oak Ridge, 

Tennessee, August 3. [SRDB Ref ID: 172409]

SC&A and Salient, 2016, Review of Petition Evaluation Report for SEC-00224, Argonne National 

Laboratory-West Regarding the Use of General Air Sampling for Internal Dose Assessment, 

SCA-TR-2016-SEC009, Rev. 0 (Draft), Vienna, Virginia, and Jefferson, Maryland, July. 

[SRDB Ref ID: 168432]

Shapiro, Z. M., 1967, “Plutonium Supply Contract No. AT(11-1)-1661,” memorandum to K. A. Dunbar 

(U.S. Atomic Energy Commission), Nuclear Materials and Equipment Corporation, Apollo, 

Pennsylvania, April 30. [SRDB Ref ID: 142890]

Stoddart, P. G., 1967, “Health Physics Aspects of Cold Line Operations: Initial Progress Report,” 

memorandum to E. D. Graham, University of Chicago, Argonne National Laboratory, Idaho 

Falls, Idaho, August 29. [SRDB Ref ID: 145181]
Document No. ORAUT-RPRT-0089 Revision No. 00 Effective Date: 04/19/2022 Page 57 of 72

Strong, R., 1988, “A Review of Experience with Plutonium Exposure Assessment Methodologies at 

the Nuclear Fuel Reprocessing Site of British Nuclear Fuels PLC,” pp. 121–131, Proceedings 

of the Second Conference on Radiation Protection Dosimetry, Held in Orlando, Florida, 

October 31-November 3, 1988, ORNL/TM-10971, Martin Marietta Energy Systems, Oak Ridge 

National Laboratory, Oak Ridge, Tennessee, November. [SRDB Ref ID: 172406]
Document No. ORAUT-RPRT-0089 Revision No. 00 Effective Date: 04/19/2022 Page 58 of 72

ATTACHMENT A 

STATISTICAL EVALUATION OF EBR-II COMPLEX AIR SAMPLE DATA

TABLE OF CONTENTS

SECTION TITLE PAGE

A.1 Purpose ................................................................................................................................. 59

A.2 Approach Used ...................................................................................................................... 59

LIST OF FIGURES

FIGURE TITLE PAGE

A-1 FCF Room 25 (thoria room) air data, August 1963 to November 1967................................... 60

A-2 FCF Cold-Line air data, August 1967 to March 1973.............................................................. 60

A-3 FCF Cold-Line air data, May 1975 to June 1976 .................................................................... 61

A-4 FCF RAS-TREAT sodium-loop air data, April 1970 to April 1973 ........................................... 61
Document No. ORAUT-RPRT-0089 Revision No. 00 Effective Date: 04/19/2022 Page 59 of 72

ATTACHMENT A

STATISTICAL EVALUATION OF EBR-II COMPLEX AIR SAMPLE DATA (continued)

A.1 PURPOSE

This attachment provides details on the statistical evaluations of several of the gross alpha 

radioactivity in air datasets for the ANL-W actinide-only areas. Those evaluations were only 

performed on the ANL-W air monitoring datasets for the actinide-only areas with the highest potential 

for internal exposures.

A.2 APPROACH USED

The regression on order statistic (ROS) is a method that can be used to calculate the GM and GSD of 

a lognormal distribution fit to a dataset (see ORAUT 2006 and ORAUT 2014b for technical details). 

ROS can handle datasets that have singly or multiply left-censored data and provides a graphic 

output that can be used to judge if a lognormal distribution indeed fits the dataset. In ROS the logs of 

the empirical quantiles (the observed air monitoring results) are plotted against the theoretical 

quantiles from a standard normal distribution (the z-scores or standard deviations). If the points fall 

approximately along a regression line, the data are taken to be lognormally distributed with the log of 

the GM given by the y-intercept of the line and the log of the GSD given by its slope.

The R code was used to perform the ROS calculations for this report. R is a language and 

environment for statistical computing and graphics. R is described as an integrated suite of software 

facilities for data manipulation, calculation, and graphical display. It is capable of performing a wide 

variety of statistical (linear and nonlinear modeling, classical statistical tests, time-series analysis, 

classification, clustering, and others) and graphical techniques, and is highly extensible.

ROS was performed on the ANL-W gross alpha radioactivity air monitoring datasets for the actinideonly areas with the highest potential for internal exposures. The air concentration data for those 

calculations was obtained from three spreadsheets (ORAUT 2018a, 2018b, and 2018c). Because the 

natural logarithm of zero cannot be taken, the ORAU Team evaluated results of 0 dpm/m3 with a 

single censoring level of <0.01 dpm/m3

. Note that a portion of the reported air sample results also 

included several other censoring levels imposed by ANL-W (i.e., <1% of MPC, <10% of MPC, etc.). 

Except for the May 1975 through June 1976 uranium data (which had a better fit using an unweighted 

regression), a weighted linear regression was used to fit the data where the sampling time was the 

weight. Each plot lists the GM in units of dpm/m3

, the GSD, the total number of results used in the 

regression, and the number of uncensored results used in the regression. Figures A-1 through A-4 

contain quantile-quantile plots of those datasets along with the resulting GM and GSD values. The 

files associated with the ROS, which contain the details for those calculations, can be found in 

ORAUT (2021g).

Based on the plots in Figures A-1 through A-4, a lognormal distribution provides a reasonable fit for 

the four datasets. The FCF Cold-Line air data from August 1967 through March 1973, depicted in 

Figure A-2, has the worst fit for a lognormal distribution between the four datasets. The nonlinear 

shape depicted by the air concentration data is due to the much larger quantity of data in this dataset 

that was left-censored by ANL-W (e.g., only reported as <1% of MPC, <10% of MPC, etc.), and the 

fact that some of the censoring levels were greater than many of the other reported results. As a 

result, the distribution of that dataset has a bimodal component. However, a lognormal distribution is 

still considered to be a reasonable fit for the FCF Cold-Line air data from August 1967 through March 

1973. 
Document No. ORAUT-RPRT-0089 Revision No. 00 Effective Date: 04/19/2022 Page 60 of 72

ATTACHMENT A

STATISTICAL EVALUATION OF EBR-II COMPLEX AIR SAMPLE DATA (continued)

Figure A-1. FCF Room 25 (thoria room) air data, August 

1963 to November 1967 (ORAUT 2021g).

Figure A-2. FCF Cold-Line air data, August 1967 to 

March 1973 (ORAUT 2021g).
Document No. ORAUT-RPRT-0089 Revision No. 00 Effective Date: 04/19/2022 Page 61 of 72

ATTACHMENT A

STATISTICAL EVALUATION OF EBR-II COMPLEX AIR SAMPLE DATA (continued)

Figure A-3. FCF Cold-Line air data, May 1975 to June 

1976 (ORAUT 2021g).

Figure A-4. FCF RAS-TREAT sodium-loop air data, April 

1970 to April 1973 (ORAUT 2021g).
Document No. ORAUT-RPRT-0089 Revision No. 00 Effective Date: 04/19/2022 Page 62 of 72

ATTACHMENT B 

FIGURES OF THE ANL-W ACTINIDE-ONLY AREAS

TABLE OF CONTENTS

SECTION TITLE PAGE

B.1 Purpose ................................................................................................................................. 63

B.2 Summary of Figures............................................................................................................... 63

LIST OF FIGURES

FIGURE TITLE PAGE

B-1 Early FCF floorplan with dimensions ...................................................................................... 64

B-2 FCF floorplan with room numbers and added dimensions, before August 1967..................... 65

B-3 FCF late Cold-Line era floorplan ............................................................................................ 66

B-4 FCF ventilation flow diagram during Cold-Line era................................................................. 67

B-5 FCF Room 25 (thoria room) ................................................................................................... 68

B-6 FCF Room 22 (sodium-loop glovebox room).......................................................................... 68

B-7 FCF Room 20 (Cold-Line room)............................................................................................. 69

B-8 FCF Room 26 (Cold-Line room)............................................................................................. 69

B-9 ITF .........................................................................................................................................70

B-10 ZPR-III workroom floorplan with added dimensions................................................................ 71

B-11 ZPPR workroom and vault floorplan with added dimensions .................................................. 72
Document No. ORAUT-RPRT-0089 Revision No. 00 Effective Date: 04/19/2022 Page 63 of 72

ATTACHMENT B 

FIGURES OF THE ANL-W ACTINIDE-ONLY AREAS (continued) 

B.1 PURPOSE

This attachment contains several figures associated with the ANL-W actinide-only areas. The 

purpose of this attachment is to consolidate these figures into one location, to provide a basis for 

some of the information in the main body of this report, and to facilitate review of this report.

B.2 SUMMARY OF FIGURES

Complete as-built room dimensions for the ANL-W actinide-only areas could not be found. However, 

a floorplan of the FCF from sometime between 1962 and 1964, which was potentially before 

construction of the FCF was completed, contained some dimensional information and is provided as 

Figure B-1 below (ANL-W 1962–1964). Figure B-2 is a floorplan of the FCF from before August 1967

that appears to be drawn to scale and also provides the room numbers. In August 1967, FCF 

Rooms 19 and 20 were combined into a single room that was identified as just Room 20. In addition, 

a number of dimensions were added to Figure B-2. The added dimensions were based on a 

combination of the dimensional values in Figure B-1 and dimensions that were scaled from 

Figure B-2. Using the dimensional information in Figure B-1 and scaling the room dimensions from 

objects in Figure B-2, reasonable estimates of the room sizes could be made. Figure B-3 is a late 

Cold-Line production era floorplan of the FCF that also provides the room numbers. Figure B-4 is a 

ventilation flow diagram for the FCF during the Cold-Line production era, which depicts where the 

ventilation air inlets and outlets were located. 

Figures B-5 through B-8 depict specific FCF rooms and the equipment locations in those rooms. 

These figures were obtained from the various survey records for the FCF and might not be drawn to 

scale. 

Figure B-9 is the only known depiction of the layout of the ITF. The figure was obtained from some 

1976 survey records and might not be drawn to scale. Figure B-10 is a floorplan for ZPR-III. It also 

contains added dimensions for the Workroom, which was the actinide-only area at the EBR-I 

Complex. Figure B-11 is the floorplan for the Workroom and Vault at ZPPR, which also contains 

added dimensions. The actinide-only area at ZPPR was the Workroom. 

The room sizes for the actinide-only areas in Figures B-1 through B-11 have been calculated based 

on the best available information and are documented in ORAUT (2019). Because no scale floorplan 

or room dimension information could be found for the ITF, no room sizes were calculated for the ITF.
Document No. ORAUT-RPRT-0089 Revision No. 00 Effective Date: 04/19/2022 Page 64 of 72

ATTACHMENT B 

FIGURES OF THE ANL-W ACTINIDE-ONLY AREAS (continued) 

Figure B-1. Early FCF floorplan with dimensions (ANL-W 1962–1964).
Document No. ORAUT-RPRT-0089 Revision No. 00 Effective Date: 04/19/2022 Page 65 of 72

ATTACHMENT B 

FIGURES OF THE ANL-W ACTINIDE-ONLY AREAS (continued) 

Figure B-2. FCF floorplan with room numbers and added dimensions, before August 1967

(ORAUT 2019).
Document No. ORAUT-RPRT-0089 Revision No. 00 Effective Date: 04/19/2022 Page 66 of 72

ATTACHMENT B 

FIGURES OF THE ANL-W ACTINIDE-ONLY AREAS (continued) 

Figure B-3. FCF late Cold-Line era floorplan (Forrester et al. 1989).
Document No. ORAUT-RPRT-0089 Revision No. 00 Effective Date: 04/19/2022 Page 67 of 72

ATTACHMENT B 

FIGURES OF THE ANL-W ACTINIDE-ONLY AREAS (continued) 

Figure B-4. FCF ventilation flow diagram during Cold-Line era (ERDA 1977).
Document No. ORAUT-RPRT-0089 Revision No. 00 Effective Date: 04/19/2022 Page 68 of 72

ATTACHMENT B 

FIGURES OF THE ANL-W ACTINIDE-ONLY AREAS (continued) 

Figure B-5. FCF Room 25 (thoria room) (ANL-W 1963).

Figure B-6. FCF Room 22 (sodium-loop glovebox room) (ANL-W 1974).
Document No. ORAUT-RPRT-0089 Revision No. 00 Effective Date: 04/19/2022 Page 69 of 72

ATTACHMENT B 

FIGURES OF THE ANL-W ACTINIDE-ONLY AREAS (continued) 

Figure B-7. FCF Room 20 (Cold-Line room) (ANL-W 1976).

Figure B-8. FCF Room 26 (Cold-Line room) (ANL-W 1976).
Document No. ORAUT-RPRT-0089 Revision No. 00 Effective Date: 04/19/2022 Page 70 of 72

ATTACHMENT B 

FIGURES OF THE ANL-W ACTINIDE-ONLY AREAS (continued) 

Figure B-9. ITF (ANL-W 1976).
Document No. ORAUT-RPRT-0089 Revision No. 00 Effective Date: 04/19/2022 Page 71 of 72

ATTACHMENT B 

FIGURES OF THE ANL-W ACTINIDE-ONLY AREAS (continued) 

Figure B-10. ZPR-III workroom floorplan with added dimensions (ORAUT 2019).
Document No. ORAUT-RPRT-0089 Revision No. 00 Effective Date: 04/19/2022 Page 72 of 72

ATTACHMENT B 

FIGURES OF THE ANL-W ACTINIDE-ONLY AREAS (continued)

Figure B-11. ZPPR workroom and vault floorplan with added dimensions 

(ORAUT 2019).