ORAU Team

NIOSH Dose Reconstruction Project

Technical Basis Document: Basis for Development of an Exposure 

Matrix for the Mallinckrodt Chemical Company St. Louis Downtown 

Site, St. Louis, Missouri, Period of Operation: 1942–1958

Document Number:

ORAUT-TKBS-0005

Effective Date: 10/24/2003

Revision No.: 00

Controlled Copy No.:______

Page 1 of 124

Subject Expert: Janet L. Westbrook

Document Owner:

Approval: _____Signature on File_________ Date: 10/24/2003

Jeri L. Anderson, TBD Team Leader

Approval: __Signature on File____________ Date: 10/24/2003 Judson L. Kenoyer, Task 3 Manager

Concurrence: __Signature on File__________ Date: 10/24/2003 Richard E. Toohey, Project Director

Approval: _____Signature on File__________ Date: 10/24/2003 James W. Neton, OCAS Health Science Administrator

Supersedes:

None

TABLE OF CONTENTS

Section

Record of Issue/Revisions ...................................................................................................................

Page

6

Acronyms and Abbreviations ...............................................................................................................7

1.0 Purpose and Scope ....................................................................................................................8

2.0 Introduction................................................................................................................................. 9

3.0 History of Site Use .................................................................................................................... 10

4.0 Description of the Mallinckrodt Uranium Refining Process and Other Processes ...................... 15

4.1 The Basic Process .............................................................................................................. 15

4.2 The Wartime Period (April 1942– April 1945) ...................................................................... 17

4.3 The Early Postwar Period (May 1945–December 1949)...................................................... 17

4.4 The Later Postwar Period (1950–1958)............................................................................... 18

4.5 Other Processes ................................................................................................................. 20

4.6 Ores and Other Feed Forms ............................................................................................... 22

4.7 Residues and Effluents ....................................................................................................... 23

5.0 Radiological Characteristics, Conditions, Considerations, and Available Data .......................... 25

5.1 Units, Limits, and Recommendations .................................................................................. 25

5.2 Radioactivity Content and Handling of the Ore, Uranium Products, and Residues .............. 27

5.2.1 Ore ............................................................................................................................. 27

5.2.2 Uranium Products....................................................................................................... 28

5.2.3 Residues (Wastes) ..................................................................................................... 29

5.3 Internal Dose Considerations .............................................................................................. 30

5.3.1 Particle Size, Solubility, and Composition Considerations; Sampling Methods ........... 30
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5.3.2 Airborne Dust Levels .................................................................................................. 32

5.3.3 Respirator Use ........................................................................................................... 35

5.3.4 Radon......................................................................................................................... 36

5.3.5 Surface Contamination ............................................................................................... 37

5.3.6 Information and Available Data Regarding Urinalyses ................................................ 39

5.3.7 Information and Available Data Regarding Other Types of Bioassay.......................... 41

5.4 External Dose Considerations ............................................................................................. 42

5.4.1 Gamma, Beta (Electron), and Nonspecific Beta-Gamma Exposures .......................... 42

5.4.2 Neutrons..................................................................................................................... 44

5.4.3 Information and Available Data Regarding Film Badges and Extremity 

Dosimeters ................................................................................................................. 44

5.4.3.1 Film Badge Monitoring Periods .......................................................................... 44

5.4.3.2 Film Badge Technical and Processing Information............................................. 45

5.4.3.3 Film Badge Record Types, Arrangement, and Availability.................................. 45

5.4.3.4 Extremity Dosimeters......................................................................................... 46

5.4.4 Information and Available Data Regarding Occupational X-Ray Examinations ........... 46

5.4.5 Skin Contaminations and Other Radiological Incidents............................................... 46

5.5 Other Data of Dosimetric Interest ........................................................................................ 47

5.5.1 Number of Workers .................................................................................................... 47

5.5.2 Number of Hours Worked per Week........................................................................... 47

5.5.3 Job Types, Work Areas, and Work and Access Practices........................................... 48

5.5.4 Miscellaneous Information .......................................................................................... 49

6.0 Determination of Radioactivity Intakes and Internal Doses........................................................ 50

6.1 Assumptions ....................................................................................................................... 51

6.2 Estimating Intake by Using Surrogate Worker or Estimated Thorium Intake Data 

(Tables 31 and 32).............................................................................................................. 52

6.2.1 Using the Surrogate Worker Intake Data (Table 31) ................................................... 52

6.2.2 Using the Estimated Thorium Intake Data (Table 32) ................................................. 53

6.3 Estimating Intake by Using Time-Weighted Daily Average Exposure Data 

(Tables 21-24)..................................................................................................................... 53

6.4 Calculation of Internal Doses for Missing Periods or for Comparison .................................. 54

7.0 Determination of External Doses............................................................................................... 55

7.1 General Considerations for Mallinckrodt External Dose Reconstruction .............................. 55

7.2 Unmonitored Workers ......................................................................................................... 55

7.3 Application of Dose Data..................................................................................................... 55

7.3.1 Incorporation of Available Film Badge Dose Monitoring Data ..................................... 55

7.3.2 External Exposure Geometries................................................................................... 56

7.3.2.1 Photons in the 50% AP/50% ROT Category ...................................................... 57

7.3.2.2 Photons in the 25% AP/75% ROT Category ...................................................... 57

7.3.2.3 Photons in the 75% AP/75% ROT Category ...................................................... 58

7.3.2.4 Photons in the 90% AP/10% ROT Category (Process Workers) ........................ 58

7.3.2.5 Photons: Non-Process Workers......................................................................... 58

7.3.3 Photon Energy Ranges............................................................................................... 58

7.4 Reconstructed External Dose.............................................................................................. 58

7.4.1 Estimating Dose, Unmonitored Period 1942-1945: Workers With Subsequent 

Dose Monitoring Records ........................................................................................... 58

7.4.2 Estimating Dose, Unmonitored Period 1942-1945: Workers Without 

Subsequent Dose Monitoring Records ....................................................................... 60

7.4.2.1 Dose During the 15-Week Monitoring Period ..................................................... 61
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7.4.2.2 Gaps in Data...................................................................................................... 61

7.4.2.3 Non-Specific Summary Results.......................................................................... 61

7.4.2.4 Unmonitored Workers ........................................................................................ 61

7.5 X-Ray Dose......................................................................................................................... 62

7.5.1 Mallinckrodt-Specific Information ................................................................................ 62

7.5.2 General Information.................................................................................................... 62

7.6 Other Dose Contributions.................................................................................................... 63

7.6.1 Extremity Dose ........................................................................................................... 63

7.6.2 Submersion Dose....................................................................................................... 63

7.6.3 Shallow Dose ............................................................................................................. 63

8.0 Determination of Exposures Due to Residual Contamination Remaining From MED/AEC 

Operations ................................................................................................................................65

References ........................................................................................................................................65

References ........................................................................................................................................66

ATTACHMENT A Tables.............................................................................................................77

APPENDIX A Notes on How the "Tolerance" or "Preferred" Level for Insoluble 

Uranium Compounds in Air Was Calculated in 1948 (From AEC 1949) ........ 123
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LIST OF TABLES

Table

1 Plants and buildings used at the St. Louis Downtown Site for AEC/MED 

uranium processing work ..........................................................................................................

Page

79

2 Summary chronology of Mallinckrodt site operations ................................................................ 82

3 Principal changes made in sites, processes, and equipment..................................................... 83

4 Types and quantities of material produced in association with Mallinckrodt 

uranium refining ........................................................................................................................ 84

5 Functional and process keywords and codes............................................................................ 88

6 Thorium and daughter content of the AM-7 residue .................................................................. 91

7 Airborne uranium particle size in process areas........................................................................ 92

8 Uranium dust concentrations by process, Plant 4, 1943–1952.................................................. 92

9 Uranium dust concentrations by operation, Plant 6, 1946–1957................................................ 92

10 Uranium dust concentrations by process, Plant 6E, 1950–1958................................................ 93

11 Uranium dust concentrations by process, Plant 7, 1952–1958.................................................. 93

12 Airborne uranium dust concentrations in Plant 4 Areas, 1948: AEC versus 

Mallinckrodt measured data ...................................................................................................... 94

13 Plant 4 measured daily weighted average exposure concentrations ......................................... 95

14 Plant 6 measured daily weighted average exposure concentrations ......................................... 96

15 Plant 6E measured daily weighted average exposure concentrations ....................................... 98

16 Plant 7 measured daily weighted average exposure concentrations ......................................... 99

17 Plant 7E measured daily weighted average exposure concentrations ....................................... 99

18 Average and highest airborne dust concentrations in the laundry ........................................... 100

19 Job titles and classifications, with function class codes and geometry factors......................... 101

20 Relationship between major processes and job functions ....................................................... 106

21 Uranium dust exposure levels, Plant 4.................................................................................... 107

22 Uranium dust daily weighted average exposure levels, Plant 6............................................... 108

23 Uranium dust daily weighted average exposure levels, Plant 6E............................................. 109

24 Uranium dust daily weighted average exposure levels, Plant 7 (including the 

Slag Separation Plant) and Plant 7E....................................................................................... 110

25 Measured radon concentrations at various indoor and outdoor areas, in units of 

1 x 10-10 Ci/L ........................................................................................................................... 111

26 Middlesex ore storage worker radon exposures, January 1944–January 1949 ....................... 113

27 Potential radon exposures from thorium processing................................................................ 113

28 Ratio of alpha airborne activity to surface contamination......................................................... 113

29 Contamination levels on various surfaces and pieces of equipment in the 

laundry.................................................................................................................................... 114

30 Contamination levels and associated dose rates from work clothing....................................... 114

31 Surrogate (comparable) worker inhalation intake values ......................................................... 115

32 Estimate of annual intake of thorium dust during Plant 7E operation....................................... 117

33 Exposure rates (mR/hr) from drums of K-65 and Q-11 residue in a railcar .............................. 117

34 Middlesex ore storage worker gamma-beta doses, 1944–1949 .............................................. 118

35 Measured dose rates at various positions in Plant 6 ............................................................... 118

36 Weekly external dose values, April 1942–December 1945...................................................... 119

37 Organ doses (rem) for a beam quality (HVL) of 2.5 mm Al and 80 kVp (1942-

1958) ...................................................................................................................................... 119

38 Decontamination and decommissioning years, 1959-1961: maximum annual 

doses ...................................................................................................................................... 120

39 Postoperations years, 1962-1995: maximum annual doses .................................................... 121
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40 Summary of postoperational years exposure data as measured by two survey 

groups..................................................................................................................................... 122
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RECORD OF ISSUE/REVISIONS

ISSUE 

AUTHORIZATION 

DATE

EFFECTIVE 

DATE

REV. 

NO. DESCRIPTION

Draft 08/18/2003 00-A New document to establish the technical basis for 

the development of a radiation exposure matrix 

for the Mallinckrodt Chemical Company St. Louis 

Downtown Site, St. Louis, Missouri. Initiated by 

Janet L. Westbrook.

Draft 10/17/2003 00-B Draft revision to incorporate NIOSH review 

comments. Initiated by Janet L. Westbrook.

10/24/2003 10/24/2003 00 First approved issue. Initiated by Janet L. 

Westbrook.
Effective Date: 10/24/2003 Revision No. 00 Document No. ORAUT-TKBS-0005 Page 7 of 124

ACRONYMS AND ABBREVIATIONS

AEC Atomic Energy Commission

AP anterior-to-posterior

Ci/L curies per liter

DWE levels time-weighted daily average exposure levels

EEOICPA Energy Employees Occupational Illness Compensation Program Act

FMFL fluorinated MFL

HASL Health and Safety Laboratory (of the Atomic Energy Commission)

LOD limit of detection

MAC Maximum Allowable Concentration

ME or MEP Minor Elements Production

MED Manhattan Engineer District

MFL MgF2 liner

MPC maximum permissible concentration

mR milliroentgen

mrad millirad

mrep millirep

NBS National Bureau of Standards

NCRP National Council on Radiation Protection and Measurements

NRC Nuclear Regulatory Commission

NYOO New York Operations Office (of the Atomic Energy Commission)

ORGDP Oak Ridge Gaseous Diffusion Plant

RBE relative biological effectiveness

RMF reject MgF2 material

ROT rotational

SLAPS or SLAPSS St. Louis Airport (Storage) Site

SLDS St. Louis Downtown Site

UNH uranyl nitrate hexahydrate

VLE very low enrichment

WL working level

WLM working level month

Code terms for the various uranium forms (e.g., TA-7) are given in Table 5 in Attachment A.
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This technical basis is established for the reconstruction of radiation doses to workers at the 

Mallinckrodt Chemical Company St. Louis Downtown Site, which refined uranium under contract to 

the United States government from 1942 to 1958. The two principal purposes of this technical basis 

document are (1) to provide information sufficient to enable dose reconstructors to estimate claimantfavorable doses for these workers on an individual basis under the provisions of the Energy 

Employees Occupational Illness Compensation Program Act (EEOICPA) and (2) to allow claimants, 

federal assessors, and others to understand the information sources and assumptions on which the 

dose estimations are based.

PURPOSE AND SCOPE

This document covers workers in Mallinckrodt Plants 1, 2, 4, 6, 6E, and 7 (including 7E) over the time 

period of the start of contract operations for the Atomic Energy Commission (and its predecessor 

agency) through the cessation of these operations, which is different for each plant. The whole period 

of such operations covers April 1942 through July 1958. Additionally, exposure due to residual 

contamination left in these plants after decontamination and decommissioning is considered for the 

period 1959-1995. Operations at the St. Louis Airport (Storage) Site (SLAPS or SLAPSS), a waste 

repository site north of the former St. Louis Municipal Airport, are not included.

1 
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The Manhattan Engineer District (MED), the predecessor agency of the Atomic Energy Commission 

(AEC), asked the Mallinckrodt Chemical Works in April 1942 to begin research on uranium refining 

and processing operations to lead to large-scale uranium production operations (Fleishman-Hillliard 

1967; FUSRAP 1996; FUSRAP Undated b). The work began immediately and by July 1942 

Mallinckrodt was producing almost a ton of UO2 per day (Fleishman-Hillliard 1967; Mallinckrodt 1994; 

FUSRAP 1996; FUSRAP Undated b). Later in 1942, Mallinckrodt started production of UF4 and in 

July 1953, it started the first uranium metal plant (Fleishman-Hillliard 1967; Mallinckrodt 1994). It is 

estimated that between 1942 and 1957, Mallinckrodt processed more than 50,000 tons of natural 

uranium products at the St. Louis facilities (FUSRAP 1996).

INTRODUCTION

The first three Mallinckrodt plants (1, 2, and 4) were not built for the purpose of uranium refining but 

were converted on an urgent basis from other uses and were intended to operate for only six to eight 

months. The permanent status of the Mallinckrodt site was not established until 1946 (Mason 1958a); 

a full-scale health program was not authorized until 1947 and did not get under way until 1948, as a 

joint AEC-Mallinckrodt effort (Mason 1958a). Film badging started only in late 1945 and urinalysis 

some time later. Thus there is little data on radioactivity concentrations in air or on internal and 

external doses prior to the late 1940's.

Although Mallinckrodt had its own industrial hygiene group and did most of its own safety work 

(Eisenbud 1975; MCW 1955; AEC 1950a), AEC's Health and Safety Laboratory (HASL) also did 

periodic air sampling and other surveys for the operating plants. While some records of the 

Mallinckrodt surveys survive, they essentially cover the same work and time periods as AEC's 

surveys. In this document preferential use is made of the AEC air concentration data because AEC 

developed and used the standard survey methods. During the early years, film badge servicing, 

urinalysis, and breath radon analyses were done under the aegis of HASL (AEC-NYOO) (Eisenbud 

1975); later, Mallinckrodt took over this work for its own plants and created the corresponding records. 

Eventually, however, HASL resumed doing the urinalysis and breath radon analysis (MCW 1955). It 

appears that most of the records of this monitoring are available. Thus the external dose records 

needed for dose reconstruction are mostly available from 1946 on, but are missing for the period 

1942-1945, and the internal dose records (as bioassays) are mostly available from 1948 on, but are 

missing for the period 1942-1947.

In this document, the context for interpretation of the existing records is established, along with the 

basis on which to determine missing doses for periods in which records do not exist.
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At their height, uranium-refining operations took place in about 60 buildings on the so-called St. Louis 

Downtown Site (SLDS), of which about 20 were left after decontamination (Applied Nuclear Safety 

1986; FUSRAP 2003a). All of these 20 buildings and most of the other 40 buildings are listed in Table 

1 along with their principal uses. (See Attachment A for all tables). A summary chronology of the use 

of the site is shown in Table 2.

HISTORY OF SITE USE

Plants 1 and 2 were already in existence and in use by Mallinckrodt at the time uranium-refining 

operations stated. Uranium-refining operations began at Plant 2 in April 1942 and by July 1942 it was 

producing approximately a ton of uranium oxide (UO2) per day (Mallinckrodt 1994). Facilities for 

batch production were installed in Buildings 50, 51, 51A, and 52 to produce uranium trioxide (UO3) 

from ore concentrates (Mallinckrodt 1994). The concentrates were digested in nitric acid in Building 

51 to produce uranyl nitrate (Mallinckrodt 1994; Fleishman-Hilliard 1967), which was then transferred 

to Building 52 to be purified by ether extraction to uranyl nitrate hexahydrate (UNH) (MED 1944b; 

Mallinckrodt 1994). The UNH was converted in Building 51A first to UO3 and then to UO2 

(Mallinckrodt 1994). Building 50 was used as a warehouse to store incoming feed materials, outgoing 

product material, and tanks of process liquids (Mallinckrodt 1994). Building 55 contained the 

“shotgun” laboratory that tested samples. In the spring of 1945, an annex to Building 52 (presumably 

52A) was added to serve as a pilot plant for a continuous countercurrent ether extraction process to 

replace the existing batch process (Fleishman-Hilliard 1967). Work at Plant 2 continued until 1946, 

when the plant was closed and the work moved to the newly built Plant 6 (Mason 1958a; Eisenbud 

1975; ORNL 1981; Mallinckrodt 1994).

Developmental work at the laboratory level to support Plant 2 and Plant 4 took place in Plant 1, 

specifically the Building 25 laboratory and the alley between Buildings K1E and 25 (Mallinckrodt 

1994). Experimental processing of radium-containing pitchblende ores began in Plant 1 in 1944 

(Mallinckrodt 1994), with the laboratory for the work in Building 25-2 and the pilot plant to test radiumextraction methods in Building K-1E (Mallinckrodt 1994) and in the alley (Fleishman-Hilliard 1967). 

Building 25 also contained the project offices and Buildings “P” and “Z” contained the engineering and 

other offices (MED 1944b). Buildings 40, 45, 45A, and 47 were used as warehouse buildings for raw, 

in-process, and finished materials (MED 1944b). Plant 1 was not used after about 1945 (ORNL 

1981); the offices and laboratories apparently moved to Plant 6.

Late in 1942, Plant 4, a former lumber sash and door works, was hastily converted for uranium 

refining (Mason 1958a) and dubbed "the metal plant" (AEC 1949). In April 1943 (Fleishman-Hilliard 

1967) or July 1943 (Mallinckrodt 1994), production of uranium tetrafluoride (UF4) began in Plant 4. 

The metal production took place in Buildings 400 and 401B and the UF4 production in Building 400 

(Mallinckrodt 1994). Production of metal moved from Plant 4 to Plant 6E in 1946 (Mallinckrodt 1994) 

and the UO2-to-UF4 process moved to Plant 7 in 1951 (Fleishman-Hilliard 1967; Mallinckrodt 1994). 

In about 1950 (ORNL 1981 says 1956), Plant 4 was refitted as an experimental development and a 

metallurgical pilot plant processing uranium metal and was then referred to as the “Pilot Plant” (ORNL

1981; FUSRAP 1996; Deblois undated). The "dingot" metal production process was developed and 

conducted at Plant 4 in the mid-1950's (AEC 1956b; Fleishman-Hilliard 1967), along with sporadic 

ordinary metal "derby" production on a developmental basis. Plant 4 was used until 1956 (ORNL 

1981; Mallinckrodt 1994; FUSRAP 1996; FUSRAP 2003c). (Note that Plant 4, in its post-AEC uses, 

was referred to as “Plant 10”.)

Due to the need to increase production and also due to the recognition by MED and Mallinckrodt of 

significant safety problems with dust and external doses (partly arising from the prospective extensive 

use of radium-containing pitchblende ore (ORNL 1981; Mallinckrodt 1994)), Plant 6 was built in 1945-
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46 on a large site fronting on Destrehan Street (Mason 1958a; Fleishman-Hilliard 1967; Mallinckrodt 

1994) and was then referred to as “the refinery” (AEC 1949). Most of the administrative offices, 

laboratories, and support facilities for the uranium refining operations were located there. The ore-toUO2 part of the refining process was moved there in early 1946 from Plant 2 and apparently the 

laboratory work from Plant 1, while the UO2-to-metal production remained at Plant 4 (FUSRAP 1996; 

FUSRAP 2003c; AEC 1949; Mason 1958a). However, some reduction of UO3 to UO2 appears to 

have been done at Plant 4 also, perhaps as part of pilot plant operations, as indicated by Mason 

(1958) and AEC (1950m). The incoming ore arrived by rail and was stored in Building 110; however, 

in late 1950 or early 1951 an outdoor ore storage area was added for pitchblende ore (Q-11) (AEC 

1950h). Building 104 housed the continuous process equipment, which replaced the batch process 

equipment that had been used in Plant 2, and processed mostly pitchblende ore (Mallinckrodt 1994); 

in 1949 a second digest line was added in this building to process uranium ore concentrates 

(Mallinckrodt 1994). Most of the UO2 produced at Plant 6 was trucked over to Plant 4, with the rest 

going by rail to Harshaw and Linde for some years (AEC 1949). When equipment was added to Plant 

7 to allow continuous UO3-to-UF4 conversion, Plant 6 began to produce only UO3 (Mallinckrodt 1994). 

Milling of UO3 and pre-digestion ore grinding, both conducted at Plant 6, were discontinued in 1950 

and 1955 respectively (Mason 1958a). Pitchblende ore continued to be used as feed until early 1955 

(AEC 1959). In 1957, a pilot-scale fluid-bed denitration reactor from Argonne was installed for testing 

and developmental improvements at the Destrehan site, presumably at Plant 6 (Fleishman-Hilliard 

1967).

In 1949 it became clear that process improvements to Plants 4 and 6 were not enough to bring about 

satisfactory control of dust and other hazards (AEC 1949). AEC became concerned and authorized 

funding for dust control and mechanization improvements, which were installed in 1949-1950 (see 

Table 3). Plant 6 was shut down for a time in late 1949-1950 for this purpose. It is not clear what the 

workers did during the time the plant was shut down, although it seems likely that those who had the 

necessary skills worked on the installation project.

These improvements were regarded as an interim measure, however. Mallinckrodt and AEC agreed 

to build a new pair of plants that were located at the Destrehan Street site (AEC 1949; Mason 1958a). 

The first was Plant 6E, the new metal plant, which went into operation in late 1950. Metal production 

(UF4-to-U metal) operations at Plant 4 moved to Plant 6E, which was from then on referred to in 

records as "the metal plant”. Metal production took place in Building 116 (Mallinckrodt 1994). Building 

116C was built in 1954 to recycle magnesium fluoride slag (Mallinckrodt 1994).

The second new plant was Plant 7, the green salt plant, which went into operation at some point in the 

first half of 1951. Note that FUSRAP (1996) and AEC (1949) appear to indicate that Plant 7 

operations began in late 1952, while Mason 1958a indicates a 1948 start date. The confusion may be 

due to the nature of the new direct or continuous UO3-to-UF4 process used at Plant 7 and the 

apparent use of Plant 4 facilities to test out the process, as a pilot plant, and to work out difficulties 

even after Plant 7 went into production. At Plant 7, a continuous process replaced the batch-type 

process used at Plant 4; later, equipment was added to allow the continuous production of UF4 from 

UO3 directly (Mallinckrodt 1994). Uranium metal recovery and some storage operations were moved 

to Plant 7 in 1952 (FUSRAP 1996; Mason 1958a). Some reversion of UF4 to UO2 or UO3 was done in 

1954 and perhaps later. A new wet slag (interim residue) recovery operation was added in late 1955 

in Building 701 (Mallinckrodt 1994. In 1955, very low enrichment uranium (probably only a small 

amount) as UF4 was processed at Plant 7 (AEC 1955b); about 5.5 kg of 20%-enriched uranium was 

processed in August 1956, presumably in Plant 7 (MCW 1956); and at some time in the late 1950’s 

machining of (nuclear) reactor core (elements) was done on a temporary basis in a fabrication facility 

in Building 700 (Mallinckrodt 1994).
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Plant 7E, regarded administratively as part of Plant 7, was used from 1955-1957 to process 

pitchblende raffinate (solids removed during uranium refining by wet filtration) to produce a 

concentrated thorium solution (FUSRAP Undated b, FUSRAP 2003a) by an acid digestion process 

similar to the uranium ore digestion (AEC 1955c). The concentrate was sent to the Mound site for 

further processing.

In 1957, all uranium refining operations ceased at all the plants (FUSRAP 1996; FUSRAP 2003c; 

Deblois undated; DOE 1981) and the work moved to Fernald and Weldon Spring. Eisenbud (1975) 

stated that all Plant 6 work was transferred to Weldon Spring in March 1957 (steps through UO3

production, at that time). However, Mason (1958a) stated that some Plant 7 production operations 

continued up to July 1958, when they were transferred to Weldon Spring; this was probably to use up 

the store of orange oxide that had been produced. This is supported by the statement of Mallinckrodt 

(1994) that the last of the site was put on standby in 1958.

Decontamination and surveys were performed at Plants 1 and 2 in 1948-1950 by Mallinckrodt 

personnel, applying AEC criteria (ORNL 1981); however, Mallinckrodt (1994) stated that further 

decontamination took place in 1954 and 1970. In 1951, these plants were released for unrestricted 

use to Mallinckrodt and no further AEC work was performed there (FUSRAP 1996; Deblois undated, 

DOE 1981). Some of these buildings were still extant as late as 1990. Plants 4, 6, 6E, and 7 were 

surveyed and decontaminated or taken down by AEC in 1957-1961 and released to Mallinckrodt 

without restriction in 1962 (FUSRAP 1996); the buildings removed during that time, included all of the 

Plant 4 buildings and all wet-process buildings in Plant 6 (ORNL 1981; Mallinckrodt 1994) (see Table 

1 for details). These decontaminations were done to the AEC standards then in force, i.e., not to 

background (ORNL 1981; FUSRAP 1996; Deblois undated; DOE undated). The remaining Plants 6, 

6E, and 7 buildings were decontaminated to modern standards in the 1990's and demolished in 1997 

(FUSRAP 1996).

Note that the term "Destrehan Street site" seems to have been used in the literature sometimes in 

reference to all the Mallinckrodt St. Louis facilities and sometimes just to those on the Destrehan 

Street site proper. In this technical basis, the former interpretation will be used, i.e., Plants 1, 2, 4, 6, 

6E, and 7 will be covered. For clarity in using the references, note that Plant 4 fronted on Broadway 

Street and although Plants 1 and 2 did not, either Plant 4 or all three plants could be referred to as 

“the Broadway site”. Thus the "Broadway" and "Destrehan" terms appear to be a loose reference to 

geographical location and not necessarily to a division of function or operations. Also, as will be 

explained later, due to a rotation practice, Mallinckrodt workers were apt to have worked in multiple 

plants over the course of their employment.

To summarize, work performed at the Mallinckrodt downtown facilities in St. Louis on a continuing 

basis included the following.

1. Production of UO2 and UO3 from ore, with some being shipped to Harshaw and other sites and 

with some natural and low-enrichment UO3 coming from Hanford and other sites in the 1950's 

(FUSRAP 1996; DOE 1997; Mallinckrodt 1994; MED 1946a)

2. Production of UF4, with some being shipped to Harshaw and K-25 (FUSRAP 1996; DOE 1997; 

Mallinckrodt 1994; MED 1946a; FUSRAP Undated b)

3. Production of uranium derby metal and vacuum recasting of ingot metal, with the ingots being 

the final product shipped to other sites (FUSRAP 1996; DOE 1997; Mallinckrodt 1994; MED 

1946a; FUSRAP Undated b)
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4. Recovery of scrap uranium metal, some from other sites, such as Hanford (FUSRAP Undated 

b; DOE 2000; Mallinckrodt 1994)

Other work performed for a limited period or on an occasional basis included the following.

1. Production of dingots (Fleishman-Hilliard 1967; AEC 1956b)

2. Machining of metal rods for reactor fuel slugs (Mallinckrodt 1994; FUSRAP Undated b)

3. Reversion of UF4 to UO2 or UO3 (FUSRAP Undated b, which says UO2 or U3O8; Mallinckrodt 

1994; AEC 1954g)

4. Production of UO2F2 (FUSRAP Undated b)

5. Extraction and concentration of Th-230 from pitchblende raffinate (FUSRAP Undated a; 

FUSRAP Undated b; Mallinckrodt 1994; AEC 1955c)

6. Experimental processing of very low enrichment UF4 (FUSRAP Undated b; Mallinckrodt 1994; 

ORNL 1981) and some 20% uranium metal (MCW 1956)

7. Conversion of small research quantities of aqueous uranyl nitrate hexahydrate to UO3 (DOE 

2000)

From 1946-1955, personnel provided by Ledoux and Co. of New York performed uranium ore assays 

at the Mallinckrodt downtown site in St. Louis (DOE 2001). This work was done in the "Ledoux 

Laboratory" located in Buildings 110A and 111 of Plant 6 (FUSRAP 2003b) and possibly in other labs 

at the site as well. The Ledoux Laboratory is also mentioned in Mallinckrodt film badge records, so 

presumably the Ledoux and Co. personnel were badged under Mallinckrodt's aegis.

MED (1944b) stated that non-MED Mallinckrodt employees and employees of the St. Louis Sash & 

Door Works (the former operator of Plant 4’s Building 400 in its pre-MED days) had occasion to enter 

respectively Buildings 45, 45A, and 47 in Plant 2 and “Building 1” at Plant 4 (presumably Building 

400) respectively. These employees were presumably not monitored in any way other than having 

their access controlled for security reasons. They are not considered to be atomic workers for the 

purposes of this dose reconstruction.

Mallinckrodt processing of black oxide (pre-milled ore), sodium salt, UO3 (orange oxide), and UO2

(brown oxide) was done under AEC/MED Contract W7401-Eng 1; UO2 and UF4 (green salt) under 

Contract W7405-Eng 29, 1st Phase; and UF4, biscuit metal (U), slag processing, ingots (billets), 

croppings, sawdust processing, and other metal production under Contract W7405-Eng 29, 2nd 

Phase (MED 1945a; MED 1944b).

Starting in 1952, some recycling of uranium was done by the AEC and its contractors nationwide. 

Thus the question arises whether this was done at the Mallinckrodt facilities in St. Louis. However, 

Mallinckrodt (as a company) did not begin to receive recycled uranium until 1962, which was after the 

St. Louis facilities had been shut down and their work had shifted to Fernald and Weldon Spring (the 

latter run by Mallinckrodt) (FUSRAP 2003b). Also, ORNL (1981) stated that in its pre-survey review of 

the site, including interviews with Mallinckrodt old-timers, no indications were found that there had 

ever been any process conducted under AEC contracts involving the purification or working of Th232, highly enriched uranium, fission products, or byproduct material. Thus it is assumed that no 
Effective Date: 10/24/2003 Revision No. 00 Document No. ORAUT-TKBS-0005 Page 14 of 124

recycled uranium or any of the materials listed by ORNL (1981) was handled at Mallinckrodt’s St. 

Louis facilities as part of AEC work.

In 1949, AEC requested that Mallinckrodt produce in their AEC-owned facilities 200 pounds of 

uranium metal for "nonproject and educational uses" (MCW 1949c); however, there is no indication 

that Mallinckrodt ever did so. Mallinckrodt apparently also carried on some commercial processing of 

euxenite (an ore) at Plant 5 at some point in the 1950's. 

Waste residues produced at Mallinckrodt were taken by truck to the St. Louis Airport (Storage) Site 

(SLAPS or SLAPSS), a waste repository site north of the former St. Louis Municipal Airport, also 

known as Lambert Field (FUSRAP 1996). The SLAPS site was operated by MED/AEC from 1946 

until July 1953 when the operation was turned over to Mallinckrodt (AEC 1959). Guards were 

maintained at the site from 1946 to 1951 (AEC 1959). It is not clear whether the truck drivers and 

guards were Mallinckrodt employees or not, although there are a few mentions of "airport" in 

urinalysis records that indicated work assignment or area. Because this remains to be clarified, 

SLAPS operations are not included in the scope of this TBD. 
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It is important to understand the Mallinckrodt uranium refining process in order to understand the 

radiological hazards, to follow the changes in source terms and exposure potential, and finally to 

estimate the doses to individual workers. The basic process will be described here. Then particulars 

will be discussed for three cases: the early wartime period, the late wartime and early postwar period, 

and the later postwar period. Details regarding the types and quantities of ore, uranium products, and 

residues are shown in Table 4; Table 5 is provided as a reference for technical terms and keywords. 

An additional table is given within the text of Section 4.4 to illustrate the variety of feed forms that was 

used in later years. 

DESCRIPTION OF THE MALLINCKRODT URANIUM REFINING PROCESS AND

OTHER PROCESSES

THE BASIC PROCESS

First the ore was prepared for uranium extraction, as follows. Milled ore, as ore concentrate ("black 

oxide"), was taken from its storage location(s), thawed if necessary in an enclosed "thaw oven" in the 

thaw house, and dried. It was then "digested" (dissolved) in nitric acid in a digestion vessel or tank 

(MED 1946a; AEC 1949; AEC 1978; Fleishman-Hilliard 1967). Sulfuric acid was then added to the 

solution in the vessel to precipitate the radium and lead (MED 1946a; AEC 1949; AEC 1978; 

Fleishman-Hilliard 1967). The uranium was left in solution as uranyl nitrate and the precipitate was 

filtered out (MED 1946a; AEC 1949; Fleishman-Hilliard 1967) using a string-discharge rotary vacuum 

filter (AEC 1978) referred to by the name of its manufacturer, Feinc (or FEinc). The filtered solids 

formed a radium-bearing residue also called raffinate or "cake" and referred to as the K-65 residue 

(AEC 1949; AEC 1978) or GLC ("gangue lead cake") (MED 1946a). Next, a slurry of barium 

carbonate was added to the uranyl nitrate solution to remove the sulfates, when present (MED 1946a; 

AEC 1949; AEC 1978; Fleishman-Hilliard 1967). To remove the solids, the mixture was run in a 

continuous solid-bowl centrifuge (also referred to by the name of its manufacturer, Bird) (MED 1946a; 

AEC 1949; AEC 1978). The uranium remained in solution as uranyl nitrate and the precipitate formed 

a barium sulfate cake. The uranyl nitrate solution -- the "liquor" -- was then boiled (MED 1946a). 

Calcium nitrate was added to the solution to assure nitrate saturation (MED 1946a), then the solution 

was filtered to remove any solids formed. The acidity of the solution was adjusted as needed by 

adding acid (AEC 1978; Fleishman-Hilliard 1967).

Second, uranium oxide as UO3 was extracted from the solution, as follows. In the two-step ether 

extraction process, diethyl ether was added to the solution in extraction columns. The first step was 

the acid ether extraction, including ether addition, nitric acid addition, and re-extraction into water 

(also called a "water wash"), while the second step was a neutral ether extraction followed by reextraction into water (MED 1946a; AEC 1949; AEC 1978; Fleishman-Hilliard 1967). Eisenbud (1975) 

states that in the ether extraction process, the isotopes of thorium and protactinium were left in "the 

aqueous phase" while the uranium was stripped off by the solvent (ether), so presumably the thorium 

and protactinium were separated from the uranium in the first step of the ether extraction process, 

before or as part of the water wash. After the ether extraction was complete, the uranyl nitrate extract 

solution was boiled to the molten salt to form uranyl nitrate hexahydrate ("hex liquor" or "OK hex 

liquor") (MED 1946a; AEC 1949; Fleishman-Hilliard 1967). The molten salt was then directed in batch 

form to gas-fired denitration "pots" or boildown vessels, which were continuously heated and agitated 

so as to produce the dissociation of the hexahydrate and the formation of UO3 (also called orange 

oxide or QM-2) (MED 1946a; AEC 1949; Fleishman-Hilliard 1967). The UO3 lumps were often broken 

up by crude mashing, but there was also a formal double grinding process requiring the material to be 

moved to the grinding area (MED 1944a).
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Third, the UO3 was placed on thin trays, placed in airtight boxes, and loaded into a batch electric 

(muffle) furnace to react with dissociated (cracked) ammonia to form UO2 (also called brown oxide or 

LF-9), a step that took about 7 hours (MED 1949b; MED 1946a; AEC 1949; Fleishman-Hilliard 1967). 

Usually the UO2 was further processed at Mallinckrodt, but some was shipped to non-Mallinckrodt 

sites such as Harshaw (DOE 1997). 

Fourth, the UO2 was converted to UF4, as follows. The UO2 was removed from the fiber containers 

and placed into large stainless steel drums for temporary storage (AEC 1949). As needed, it was 

removed onto steel (later monel) trays and weighed (MED 1946a; AEC 1949; Fleishman-Hilliard 

1967). The trays were then set into airtight graphite or nickel boxes placed in a hydrofluorination 

reactor (MED 1946a; AEC 1949; Fleishman-Hilliard 1967). (Note that this was not a nuclear/atomic 

type of reactor but rather a chemical reaction vessel.) In the reactor, hydrogen fluoride gas was 

passed over the UO2, forming UF4 (also called green salt or TA-7) and water (MED 1946a; FleishmanHilliard 1967). The UF4 was unloaded and put through a pulverizer or mill and then a blender, 

followed by packing into metal containers (AEC 1949). Most of the UF4 was further processed at 

Mallinckrodt, but some was regularly shipped to non-Mallinckrodt sites (DOE 1997; AEC 1949); MED 

(1946), however, implies that all of the UF4 produced at Mallinckrodt was also further processed there.

Fourth, the UF4 was converted to uranium metal, as follows. The UF4 was blended with magnesium 

powder and mixed in a tumbler (AEC 1949; Fleishman-Hilliard 1967). A special firing container (the 

"bomb") was lined with a refractory material, variously reported to be lime (1945, per MED 1946a); 

dolomite (1946 on, per FUSRAP Undated a and FUSRAP 1996, or 1942 on, per Fleishman-Hilliard 

1967); or recyclable magnesium fluoride from about 1954 on (FUSRAP Undated a; FUSRAP 1996; 

Fleishman-Hilliard 1967). The bomb was "jolted" (shaken) on a mechanical jolter until the liner was 

packed hard enough. The mixture (the "charge") was then placed in the bomb. After the bomb was 

sealed, it was placed in a gas-fired furnace and heated until it "fired"; the magnesium reduced the UF4

to uranium metal in this process (MED 1946a; AEC 1949; Fleishman-Hilliard 1967). The molten mass 

formed a "derby" or "biscuit" of solidified metal as it cooled (MED 1946a; AEC 1949; FleishmanHilliard 1967). (The derby form was also called KB-2.) After the bomb cooled, the derby was taken 

out and the slag on the derby was chipped off pneumatically and sent to uranium salvage (recovery) 

(MED 1946a; AEC 1949; Fleishman-Hilliard 1967).

Finally, the metal derbies were remelted and cast in an induction-heated, high-vacuum furnace (a 

process also called vacuum recasting). In this process, the derbies were placed in a crucible, the 

crucible was placed over a mold, and the crucible-mold assembly was placed in a quartz shield, 

sealed, evacuated, and placed into the furnace (MED 1946a; Fleishman-Hilliard 1967). There was a 

stopper rod at the bottom of the crucible that was removed when the critical temperature was 

reached, allowing the molten metal to flow into the mold and form a cylindrical ingot or billet (MED 

1946a; AEC 1949; Fleishman-Hilliard 1967). (The billet form was also called YM-5.) The billet was 

then removed from the mold by separating its parts. The porous impurity-heavy top section was

cropped off and sent for recovery; the other surfaces were cleaned of liner ("C-liner") slag and other 

impurities (MED 1946a; AEC 1949; Fleishman-Hilliard 1967). A sample was taken from the billet by 

power hacksaw (AEC 1949). The billets were then packed and shipped to the appropriate nonMallinckrodt site for further processing (MED 1946a; AEC 1949).

The nitric oxides produced as offgasses in the pots were sent to a recovery system (MED 1946a; AEC 

1949; FUSRAP 2003b) where the nitric oxides were converted back to nitric acid. The GLC residue 

and the barium sulfate cake were usually leached with sodium carbonate to remove the residual 

uranium (AEC 1978).
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The precipitate from the first extraction column was "on occasion" dewatered using a Sperry 

Filterpress, producing a supernate and a batch of cake called Sperry cake (AEC 1978). The 

supernate from the press and the aqueous uranium tails from the wash were de-etherized and treated 

with a hydrated lime slurry (AEC 1978). This was passed through a continuous rotary vacuum leaf 

filter (referred to by the name of its maker, Niagara); the supernate was discharged to the local river 

and the limed fraction became the so-called "airport cake" (AEC 1978), from its being stored for 

several years at the AEC's SLAPS site that had been an airfield; however, it must be noted that 

several other types of cake were also referred to loosely as "airport cake" due to their being sent there 

for storage.

THE WARTIME PERIOD (APRIL 1942– APRIL 1945)

Initially, the work consisted primarily of the production of UO2 and UO3 from mined ore marginally 

milled (FUSRAP Undated b); some preliminary milling to "black oxide" (a form of U3O8 concentrate) 

thus had to be done, some of it at Mallinckrodt (DOE 1997). Little information is available as to how 

or where the early milling operations were performed, but it was likely in Plant 2 (see Table 1). Some 

“soda salt” (sodium diuranate) was also used (DOE 1997).

Plant 4 began operation in about October 1942 and from that point on, production-level work in 

converting UO2 to UF4 and then to uranium metal took place there. The ore→UO3→UO2 conversion 

continued to be done in Plant 2.

In the first months of refining by Mallinckrodt, a different extraction process from the ether extraction 

was used since the latter had not yet been developed. No information could be found as to the details 

of the first process; however, the ether extraction process seems to have been in the process of 

development at Mallinckrodt from April 1942 on and in use from July 1942 on (US DOE 1997; AEC 

1978; Fleishman-Hilliard 1967), so the first process could have been used for about three months at 

most (mid-April to mid-July). The magnesium reduction process for metal production was developed 

by Iowa State by about mid-July 1943 (DOE 1997) and Mallinckrodt established the first metal plant 

(Plant 4) the same month (Fleishman-Hilliard 1967).

THE EARLY POSTWAR PERIOD (MAY 1945–DECEMBER 1949)

As noted above, processing of significant quantities of pitchblende ores began at Mallinckrodt in about 

May 1945 and accelerated with the start of Plant 6 in 1946.

During and after the war, African and Canadian ores were milled to black oxides elsewhere and at 

Mallinckrodt (DOE 1997, Mason 1958a); at Mallinckrodt, this moved from Plant 2 (apparently) to Plant 

6 after the latter began operation in 1946. Mason (1958a) gives data showing very high dust levels 

for ore milling, so the process was apparently not an entirely enclosed one. Ore milling stopped in 

mid-1949 (Mason 1958a) and after that all ore arrived mill-processed. The ore delivered to 

Mallinckrodt then had to be pre-processed only minimally before the refining process proper could 

begin. DOE (1997) states that it was done at Mallinckrodt's "Destrehan Street plant", inferentially at 

Plant 6 (see Table 1).

The continuous countercurrent ether extraction process was developed in 1945-1946 and put into 

production with the start of Plant 6 in 1946 (Fleishman-Hilliard 1967). This eliminated the necessity of 

moving the intermediate materials batchwise from one process location to another as had been 

necessary at Plant 6; rather, the materials passed from vessel to vessel or filter via piping. Also, the 

sulfur removal step was added when ores containing high levels of sulfur were used. However, the 

removal of products such as UO3 and UO2 was usually done by manual scooping and the removal of 
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waste products, such as the filtered-out solid wastes and recovery products, was usually done by 

manual methods such as scraping.

As noted earlier, Plant 6 was shut down for a time in 1949-1950 for modifications; it seems likely that 

the end of ore milling coincided with this shutdown. Mechanization improvements during this period 

decreased the amount of manual and close-in handling of ore and the other uranium forms, but there 

was still a significant amount, especially with regard to the intermediate solid forms (UO2, UO3, and 

UF4) produced as particles and chunks of material.

THE LATER POSTWAR PERIOD (1950–1958)

The Middlesex work was transferred to the Fernald plant in 1955 (DOE 1997), which thereafter was 

the sender of milled ore to Mallinckrodt. An Ore Room was added at some time in the early 1950's to 

Plant 6; the duties of the operator included opening (deheading) and cleaning ore drums (AEC 1953), 

hence this was not simply a storage area. During the period 1950-1951, Plant 6E and Plant 7 went 

into operation as previously described and a number of further process and equipment changes were 

made to reduce exposures, principally in the area of dust control and mechanization (see Table 3). 

The main reason for the provision of new plant space was to reduce exposures, but it is notable that 

in each case, production increased markedly after a new plant was added, sometimes to several 

times the original anticipated capacity (Mason 1958a). Thus the exposure-reduction changes were 

sometimes successful and sometimes not.

Some time prior to August 1954, the Ore Room and K-65 sampling operations in Plant 6 appear to 

have ended, probably due to the end of high-grade ore processing. The Plant 6 pilot plant was 

constructed in 1949-1951 and began operation at some point during that period (Fleishman-Hilliard 

1967); references to it started to appear in AEC air dust study reports by 1953 (AEC 1954c). The 

purpose of this pilot plant was said to be process and product improvement (Fleishman-Hilliard 1967). 

An AEC dust exposure report giving time-and-motion information states that a Pilot Plant 6 technician 

was cleaning old MgF2 out of a kiln, putting in new MgF2, and scooping Anaconda feed (sodium 

diuranate) into a digestion tank (AEC 1956c). Thus it is likely that the pilot plant was working on the 

development of improved methods for various parts of the refining processes. The research laboratory 

was constructed and put into operation during the same period as the pilot plant (Fleishman-Hilliard 

1967).

An alternative method of producing the metal form was developed at Plant 4 in the mid-1950's. This 

process (AEC 1956b; Fleishman Hilliard 1967) took the UF4 to the final cast (ingot/billet) form in one 

step instead of two, in order to eliminate the need for recasting and the associated contamination with 

impurities. The result was a massive single ingot, called a dingot, weighing about 3300 pounds. In 

this process, a bomb was lined at the bottom with MgF2, a mandrel (mold liner support) was inserted 

into the bomb, more MgF2 was poured into the space between the bomb shell (wall) and the mandrel, 

and the bomb was jolted to pack the liner hard between the mandrel and shell. The UF4 and Mg were 

blended and put into a drum, which was then capped and removed to a charging station. The drum 

was inverted over the station insert collar, the drum cap valve was opened, and the charge was 

allowed to flow into the bomb. During this step, the operator used a long stainless steel rod to "pole" 

the charge down into the bomb; the operator also used a mechanical rammer to tamp it down for 

maximum density. Recycled slag containing MgF2 and U was added to the top of the charge and

tamped. Finally, the exterior of the bomb was vacuumed and a steel lid was applied. The bomb was 

removed by hoist to a "Hevi-duty" furnace, where the thermite (metal-metal reduction reaction) took 

place.
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Because of the size and purity of the dingot, the postcasting processing differed from that of the derby 

and billet. The bomb was cooled in air, then transferred to a breakout enclosure where it sat over a 

downdraft-ventilated floor grill. The bomb was inverted over the grill and the contents jolted out onto 

the grill. The slag liner was broken with a mechanical sledge hammer and swept onto the grill and 

down into a hopper. A conveyor took it to a grinder, from which it was discharged into drums and 

taken for reprocessing (presumably in Plant 6E -- see below). The dingot was cleaned with a 

pneumatic chipping hammer over the grill and then removed to a machining area, where it was 

"scalped" (had the outer surface trimmed off) on the top and sides with a vertical turret lathe. If 

samples were needed, it was then taken to a saw area and sections were cut off with mechanical 

hacksaws. Finally, the dingot was put into a salt bath and heated for several hours, then put on the 

bed of a 100-ton forging press that had a mechanical manipulator for positioning. It was pressed into 

a slab for several minutes, then returned to the salt bath for about an hour for reheating. This process 

was repeated four more times, with the piece being rotated on the press between passes. The forged 

slab was then quenched and taken to storage.

Fleishman-Hilliard (1967) implies that Mallinckrodt also produced extruded billets from these dingots, 

at least on an experimental basis. Once extruded into a much longer and thinner cylinder, the billet 

was cut into shorter lengths.

Plant 7 was built not only to increase production of UF4, but also to take advantage of the new 

continuous process using a “stirred bed” reactor for producing it (Mallinckrodt 1994, Fleishman-Hilliard 

1967). At some point not long after this process went into operation, the process was modified to 

allow production of UF4 continuously from a UO3 feed instead of a UO2 feed, thus eliminating a 

transfer step (Mallinckrodt 1994, Fleishman-Hilliard 1967). At this point, UO3-to-UO2 production at 

Plant 6 seems to have ceased or at least decreased significantly.

By 1955, Mallinckrodt was using a variety of feed forms from many different sources to produce the 

various uranium product forms. The table below (AEC 1955a) illustrates this for a typical period in 

1955; figures are in tons.

January February March April

To be produced:

QM-2 500 465 535 510

TA-7 280 355 320 320

TM-5 205 240 270 260

Shipments to NLO:

QM-2 235 155 170 170

YM-5 All finished metal of the metal plant plus all shippable metal from the pilot plant

TA-7 0 100 100 45

Feed type

January 

consumption

January 31 

inventory

February 

receipts

March 

receipts

April 

receipts

Q-11 76 8 86 --- ---

MgX 26 10 40 64 30

South African --- --- 129 80 90

Portuguese 21 --- 40 --- 20

Colorado soda salt 46 12 50 50 50

Beaver Lodge 23 31 28 28 28

Vitro 9 --- --- --- ---

Colorado black 92 12 50 50 50

Canadian black 43 17 25 25 25

NLO Recycle/Scrap Plant 27 15 20 20 20

MCW 18 --- 15 15 15
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Miscellaneous --- --- --- --- ---

Dissolver 24 77 20 20 20

Sawdust 5 6 5 5 5

OTHER PROCESSES

Uranium recovery operations consisted mostly of processing solid scrap wastes, such as the portions 

of billets removed as assay samples, to recover the valuable uranium and thus to maximize the 

uranium obtained per unit quantity of ore. This was probably done in the early years by digestion of 

the scrap in nitric acid (FUSRAP 2003a). Sawdust was stored under oil until it could be processed in 

this way and converted to ingots (Fleishman-Hilliard 1967; MED 1944a; MED 1944b). Some 

processing of residues was also done (e.g., leaching of wastes) to recover uranium, but this seems to 

have been done as an integral part of the refining process itself rather than a stand-alone operation 

like the solid scrap salvaging.

In early 1954, a slag separation plant was built at Plant 6E, apparently as a six-level wing (116C) on 

Building 116, to recover most of the uranium content from the MgF2 slag produced there in the UF4-toderby operation (AEC 1959, AEC 1954f). The slag was fed into a jaw crusher on the first level, then 

the crushed slag was conveyed on a "vibro conveyor" to a bucket elevator and then to a roll mill 

hopper on the 6th level. From there the slag was gravity-fed via a roll mill feeder into a series of roll 

mills and screens on the middle levels. Reject material from the last mill screen flowed by gravity feed 

into a reject hopper and then to a reject drum on the 1st level. Discharge streams from the roll mill and 

the roll mill screens were fed into a ball mill. The discharge from the ball mill was conveyed by a 

bucket elevator to a series of air classifiers; after passing through these, the separate product and 

reject streams flowed by gravity feed to the respective drums on the first level.

In late 1955 or early 1956, operations to process slag began in Building 701 in Plant 7. This building 

was then called the Slag Separation Plant or the Slag Processing Plant. According to AEC (1956e), 

the feed material was uranium concentrate from reject MgF2 material (RMF). The RMF was charged 

from drums into a skip hoist that discharged into a hopper serving a rod mill. In the mill, the RMF was 

crushed and water was added to make a slurry. This was passed through a mechanical screen 

shaker; the part that did not pass through was recycled through the rod mill and the part that did pass 

went onto three successively smaller-sized Wolfey gravity feed tables. The tables separated the 

uranium-bearing slurry from the rest of the slurry, with the former being allowed to run off the last table 

into a screw conveyor. The slurry was then fed into a drum. This drummed material was referred to 

as "U-con". The non-uranium-bearing slurry was pumped to settling tanks and then to a rotary filter; 

the resulting cake (D-701) was discharged into dumpsters (buggies) and taken to the SLAPS waste 

storage site.

In late 1954 or early 1955, a new subplant, called Plant 7E, was established. This was part of what 

was referred to as the Minor Elements Production (ME or MEP) facility that did smaller-scale 

processing and development in Plant 7. The purpose of 7E was to process some of the AM-7 

(pitchblende) residues that had been in storage at the SLAPS site (FUSRAP Undated a; AEC 1978; 

FUSRAP Undated b; FUSRAP 1996). From the beginning of 1955 into 1957, Mallinckrodt produced a 

concentrated thorium solution for later extraction of Th-230 at the Mound site (AEC 1978; FUSRAP 

1996). (Note that the Th-230 was also called "ionium" and was referred to that way in Mallinckrodt 

records.) The residue was brought back from storage at SLAPS and stored at Plant 6. According to 

AEC (1955b), the residue was conveyed by dumpster from Plant 6 to Plant 7E, where it was digested 

in nitric acid in a tank and filtered. The resulting "liquor" was processed in a TBP (a solvent) 

contactor, forming an aqueous phase and a thorium-bearing TBP phase. The TBP phase was treated 

with hydrofluoric acid, leaving another aqueous phase, a stripped TBP phase, and thorium fluoride. 
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The thorium fluoride was treated with aluminum nitrate (Al(NO3)3) and a pentaether-ether mixture, 

forming thorium nitrate (Th(NO3)4) and impurities. The thorium nitrate was stripped from the impurities, 

yielding the solution that was sent to Mound. The various waste streams were treated in several ways 

(e.g., the aqueous fractions were treated with lime). This left various cake, slurry, and water filtrate 

forms; the first two types were sent to storage and the latter to the sewer. The main residual cake, 

called AM-10, was sent back to SLAPS storage.

Little information could be found about shipping, receiving, and storage operations and about the 

milling of the UO3, except that it was apparently done in Plant 2 until the work moved to Plant 6 

(Mason 1958a). The ore areas and warehouse areas appeared to be separate, however.

Similarly, regarding the reversion of UF4 to UO2 or UO3 (or U3O8) there was little information except for 

a short reference to the work of a panel board operator's work in an AEC Plant 7 air dust study report 

(AEC 1954g): the operator was said to vacuum "C-31 material" from a drum into a "reverter". What 

the C-31 material consisted of was unclear, but usually the "C-" designated scrap material or material 

captured in a dust collector. Later such AEC reports mention that the reverter was being used by this 

operator (AEC 1955b, AEC 1956c) but not an earlier one (AEC 1952b), so presumably this work 

continued from at least 1953 (the date of the AEC 1954g survey) to the end of Plant 7 operation.

There was no information about processing very low enrichment (VLE) uranium beyond what was 

given in Section 3.0 above, except for a reference in an AEC air dust study report for Plant 7 in which 

it was stated that the subject AEC survey covered "health and safety problems existing during 

production and processing green salt, with the added operation of processing enriched uranium"; the 

panel board operator was said to include among his observed duties "charging enriched UF4 into 

hopper" and "replacing enriched material drum and sample bottle" (AEC 1955c). Since there was no 

further mention of any of these activities in later such reports, it appears that this operation may have 

been performed for only a short period of time, e.g., to use up excess UF4 from another site. There is 

no indication that Mallinckrodt itself produced the VLE UF4. In any case, the operations described 

appear to be routine.

There was no information at all about the production of UO2F2 or the processing of (supplied) uranyl 

nitrate hexahydrate, all of which functions were mentioned by only a single reference source (see the 

summary list of work in Section 3), with no details given. There was information (AEC 1955b) 

regarding a Plant 7 experimental process begun in Plant 7 in 1956 in which MgF2 liner (MFL) was 

treated to reduce the hydrogen content for use in dingot bombs. This created fluorinated MFL (FMFL, 

perhaps also called D-30). However, this process appeared not to involve any radioactivity except 

what would likely have been present in the ambient air due to other processes. Also, FleishmanHilliard (1967) commented that the reason for adding a little UO2F2 to a bomb was to slow down the 

process themally, resulting in a better separation of slag and metal.

Mallinckrodt (1994) estimated that small-volume batch and experimental uranium-related processes 

constituted less than .3% of the total uranium produced. Some of the processes considered in this 

estimate were the conversion of slightly (very low) enrichment feed materials to metal or uranium 

nitrate liquor, extraction of Th-230 from pitchblende raffinate, production of U3O8 and UO2 using an 

experimental continuous denitration furnace, experimental extraction of uranium using tributyl 

phosphate to replace ether, production of uranium metal dingots, and recycling of slag for use as liner 

material. This supports an assumption that only a negligible part of total processing activities involved 

such short-term and usually low-volume processes.
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1.1 ORES AND OTHER FEED FORMS

In the early years, the ore arrived as a milled concentrate from Canada (Mallinckrodt 1994), consisting 

mostly of black oxide (U3O8) (Mallinckrodt 1994; Fleishman-Hilliard 1967); MED (1945) stated that the 

Eldorado mine site in Port Huron, Canada, was a source of black oxide to Mallinckrodt. These 

concentrates were produced at offsite uranium mills and were free of radium and its decay products 

(Mallinckrodt 1994). That is, in the early years, Mallinckrodt did not process the high-radiumcontaining ore of later years. Although the uranium was mostly supplied to Mallinckrodt in already 

milled form (as black oxide), some milling and pre-processing of high-grade uranium ore was 

performed at Mallinckrodt (FUSRAP 2003c).

The Vitro plant in Canonsburg, Pennsylvania produced sodium diuranate, also called soda salt, some 

of which was sent to Mallinckrodt for refining to UO2 and UO3 (DOE 1997, MED 1945a; MED 1943c); 

other soda salt may have come from Belgium and Port Hope, Ontario (MED 1943c). This appears to 

have been digested in the same fashion as the black oxide. Soda salt and other non-ore feeds were 

referred to as "soluble feed".

Other wartime ore sources were pitchblende ores from Canada (Radium City in the Great Bear Lake 

Area and Port Hope in Ontario) (DOE 1997; Eisenbud 1975). Later in the war, domestic ores were 

used also; the principal Colorado source during the war was not the carnotite ore itself, but tailings 

from vanadium mining and milling, shipped as a concentrated sludge (Eisenbud 1975).

In about 1944, AEC wanted to increase production of uranium, not only by increasing capacity but by 

using high-grade pitchblende ores. However, such ores had the disadvantage that they contained a 

high level of radium as a decay product of the uranium. This entailed significant gamma and airborne 

exposure hazards. It was because of the limitations of Plant 2 with respect to capacity and hazard that 

Plant 6 was designed and built (Mallinckrodt 1994). Another consideration was that crude 

concentrates and pitchblende ores tended to have undesirably high levels of sulfur, which 

necessitated the extra precipitation step for removal (Fleishman-Hilliard 1967). The high-level ore 

was used as a principal feed material from May 1945 until early in 1955 (AEC 1956a; AEC 1959). 

However, MED (1943c) remarked in 1943 that there appeared to be more “radiation” in the U3O8

being received by Mallinckrodt from Port Hope, suggesting that pitchblende ore concentrates may 

have been used as feed prior to May 1945. It is known that experimental processing with pitchblende 

ores was done in 1944 at Plant 1 (Mallinckrodt 1994), so this may explain the 1943 reference.

Most of the high-grade pitchblende ore processed by Mallinckrodt was obtained by AEC as a 

concentrate from the Belgian Congo in 1944 (AEC 1978), the so-called African ore. From 1943 on, a 

receiving and storage facility operated at Middlesex, New Jersey (DOE 1997). Middlesex assayed, 

crushed, riffled, and redrummed the ore as it came into the United States, then sent it to the various 

refineries, including Mallinckrodt (AEC 1949, Eisenbud 1975), including the African ore. However, 

SSS states that the African ore came through the Eldorado (Port Huron, Canada) area for processing 

before being sent on to the US refineries. Whatever the case, it was not likely that significant milling of 

this ore was done at Mallinckrodt.

After World War II, foreign uranium ore was supplied from the Belgian Congo, Canada, Australia, 

South Africa, Portugal, and other nations. Besides the foreign ore, domestic carnotite ores began to 

be mined directly, milled elsewhere, and then sent as ore concentrate (mainly U3O8) to Mallinckrodt. 

Early postwar domestic ore supply areas included Uravan, Durango, Grand Junction, and Naturita, 

Colorado, and Monticello, Utah (DOE 1997). Once the US began to stimulate domestic mining and

milling in 1948, the proportion of domestic ores used appears to have increased; thus the overall 

concentration of uranium in the ore may have decreased over time.
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Soda salt (sodium diuranate) appears to have been sent from several sites; a description of some 

work indicated that Fernald, Durango, and Anaconda diuranate were being handled by a soluble feed 

operator (AEC 1956c) and as noted Vitro also supplied soda salt until it was shut down.

1.2 RESIDUES AND EFFLUENTS

The radium-bearing pitchblende residues (wastes) were always stored separately from the nonradium-bearing residues and the barium-bearing and other types of cake (AEC 1978); thus the 

external exposure implications of handling residue waste depended on what kind of waste it was.

From 1942-1945, MED solid wastes were shipped to the Lake Ontario Ordnance Works in Lewiston, 

New York (Mallinckrodt 1994). The radium-bearing pitchblende residues were shipped by rail to the 

Middlesex, New Jersey repository for storage up to about 1946 or 1947 (AEC 1949). From 1946-

1948, all residues produced at Mallinckrodt were shipped by truck to the SLAPS waste storage site --

hence the name "airport cake", as noted earlier (AEC 1978; Mallinckrodt 1994). From 1948 to at least 

1949, some residues appear to have been shipped to Middlesex (AEC 1949). After about 1949, 

radium-bearing residues were shipped to the Lake Ontario Ordnance Works in Lewiston, New York 

(FUSRAP undated a; AEC 1949), although Mallinckrodt (1994) says that these were still sent to 

SLAPS up to 1956. Some uranium scrap was sent to the Vitro Manufacturing Company for uranium 

recovery (AEC 1949).

The gangue lead cake wastes (K-65) from pitchblende processing was stored in Building 114 until 

removed by AEC (Mallinckrodt 1994). The barium sulfate and raffinate cakes were collected in 

dumpster-type containers, loaded into AEC dump trucks, and transported on a daily basis to SLAPS 

(Mallinckrodt 1994). 

The non-recyclable materials formed a slag waste called "C-liner" or "C-liner slag", which was mainly 

used dolomite liner with less than 2% uranium content (FUSRAP Undated b; AEC 1949). It resulted 

from the separation of slag in the reduction (bomb) step in the metal plant (first in Plant 4, later Plant 

6E). This waste was created until early 1953 when the dolomite liner was replaced by a recycled 

magnesium fluoride liner (FUSRAP Undated b). There was another bomb waste called "C-special" 

(AEC 1949), which may be the same as the C-liner slag since it also originated in the bomb furnace. 

Yet another bomb waste was the recyclable magnesium fluoride slag that remained after "scalping" 

the uranium content from the magnesium fluoride slag, beginning in 1955; this was called C-701 

(FUSRAP undated a). These wastes were also sent to SLAPS.

The Sperry Cake was found to be a good source of protactinium-231. Some of it was shipped to 

Mound which processed approximately 20 tons (about eighty 55-gallon drums) of it and obtained 

approximately two grams of protactinium-231 (AEC 1978). There were also small amounts of 

raffinates present at the SLAPS waste storage site from the processing of lower-grade uranium ores 

(AEC 1978). The material obtained from processing carnotite and other low-grade domestic ores was 

kept separate from the pitchblende residue (AEC 1978).

Most “product” (UO3, UO2, and UF4) was packaged into 75-lb paper sacks and then into a barrel. 

There was usually a fine film of uranium material clinging to the sacks, so they were burned in an 

incinerator and the ashes sent to recovery (MED 1944a). Floor sweepings and spillages, swept into 

sumps; scrapings off the walls and equipment; and dust collected in collectors, usually by vacuum 

systems, were also sent to a recovery system (MED 1944a).

There were separate sets of sewers running from Plants 1 and 2, Plant 4, and Plants 6, 6E, and 7 out 

to the streets, under the streets, and out to the Mississippi River (Mallinckrodt 1994). In 1949, about 3 
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million gallons a day of liquid effluent classed as alkaline filtrate was being sent to the Mississippi 

River out a drain pipe, presumably a sewer; this was mostly cooling water, but some was process 

waste containing up to 12 pounds per day of uranium (AEC 1949). A small additional amount of acid 

waste flowed to the Mississippi via an acid sewer (AEC 1949). Mallinckrodt (1994) stated that a 1956 

description of Plant 6/6E/7 operations showed that 12,000 gallons per day of raffinate filtrate from the 

Plant 6 pitchblende extraction was being discharged to the sewer. These lines were apparently 

underground for the most part, but it is possible that some of them were aboveground where they ran 

inside the site (see the Table 1 Notes column). 
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2.0

As AEC (1950a) observed, no radiation measurements or evaluations of dust exposure were made in 

the Mallinckrodt plants in the first few years of operations because it was expected that the processing 

of uranium ores and compounds would involve little risk of radiation injury; this was because of the 

low specific activity of uranium and because of what was thought at the time to be the temporary 

nature of the work. However, when AEC’s New York Operations Office (NYOO), which oversaw the 

Mallinckrodt work, evaluated the potential hazards (which at this time included those of radiumbearing pitchblende ores), they determined them to be “considerable” (AEC 1950a). NYOO and 

Mallinckrodt began a program of workplace and personnel monitoring.

RADIOLOGICAL CHARACTERISTICS, CONDITIONS, CONSIDERATIONS, AND 

AVAILABLE DATA

AEC and Mallinckrodt had already begun to issue film badges in 1946 (AEC 1950a), with apparently a 

small-scale effort begun in late 1945. To this was added breath radon determinations in 1946 and a 

formal dust measurement program in 1948 (AEC 1950a). Urinalysis measurements appear to have 

begun in 1947 also.

Since little individual monitoring data is available prior to about 1946, some extrapolation of existing 

data to cover the unmonitored periods is necessary, as AEC itself tried to do (AEC 1950a). Also, data 

must be analyzed to allow missing dose to be calculated for individual workers where there are gaps 

in the monitored period. The sections below provide information as to the available data and other 

information that will allow this to be done.

AEC thought that as a result of improvements planned for 1949 and early 1950, there would be no 

whole-body radiation exposures greater than 300 mrep/week in Plant 6 and the dust concentrations 

would be reduced to the AEC's "preferred level" of 50 μg/m3

, or 70 dpm/m3 (AEC 1949). AEC was 

also expected that construction of a new metal plant (Plant 6E), in which UF4 would be reduced to 

metal as was currently done at Plant 4, would produce satisfactory (occupational) environmental 

conditions (AEC 1949). It was also expected that in 1951, the new UO2-to-UF4 plant (Plant 7) would 

further reduce exposures (AEC 1949). However, with the increase in production, these goals were 

not met in all cases, although there were some successes and although doses and air concentrations 

did decrease overall. The effects of the various plant changes and improvements were reflected in 

the airborne and external exposure levels, as shown in the text and tables below.

2.1 UNITS, LIMITS, AND RECOMMENDATIONS

The exposure (dose) units used by MED/AEC during most of the relevant period were milliroentgen 

(mR) for gamma doses and millirep for beta doses; the abbreviations in the film badge and other 

records were mr and mrep respectively. The rep is equal to 0.93 rad.

During the war, AEC's recommended (“tolerance”) levels of external exposure for the uranium 

processing plants were 700 mR per week to the whole body and 3500 mrep per week to the hands 

(AEC 1949; AEC 1950b); AEC (1950b) stated that this was 700 mrep per week “each of beta and 

gamma”, as accepted by the University of Rochester in processing film badges (i.e., Rochester did not 

flag reported doses as above tolerance if the weekly beta and gamma doses were each below 700 

mrep). At some point, when NYOO had assumed the job of reading the film badges, the tolerance 

level was lowered to 500 mrep per week, which Mallinckrodt continued to interpret as applying to 

either beta or gamma but not to the total (AEC 1950b). However, Mallinckrodt used a control level of 

150 mrep per week, called the "preferred level" in its 1946-1952 film badge records. The number that 

was compared to this level was the sum of the gamma dose in mR (i.e., mr in the records) and the

beta dose in mrep, as registered by the film badge, apparently with no adjustment of the mrep by the 
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factor of 0.93. MCW gave as the “tolerance cumulative dose” limits in use in 1955 as beta, 500 mrep 

per week, whole or part body; gamma, 300 mR per week, whole or part body; sum of beta and 

gamma, 500 mrep per week whole or part body; and 1500 mrep per week, extremity (MCW 1955).

In 1947, the basic dose limit was 0.1 rep/day and the relative biological effectiveness (RBE) for alphas 

was 10 (Hursh 1975); probably this was true earlier as well.

In anticipation of the lowering of the radiological dose standards (recommended limits) by the national 

expert committees such as the National Council on Radiation Protection and Measurements (NCRP), 

AEC began making changes in the plants in about 1948 to meet new "maximum permissible levels" of 

300 mR/week for whole body irradiation and 1500 mrep/week for beta radiation to the hands (AEC 

1949). In August 1949, AEC established and circulated to its contractor personnel a tolerance level of 

300 mrep per week, which was to be taken as the total gamma plus beta dose to the whole body. 

However, Mallinckrodt misunderstood that the 300 mrep limit was to be applied to the total of beta and 

gamma and interpreted it as the limit for either beta or gamma (AEC 1950b). Finally, in January 1950 

AEC made it clear to Mallinckrodt that the limit applied to the total beta plus gamma (AEC 1950b)

In mid-1950 AEC agreed to allow Mallinckrodt to interpret the 300 mrep total gamma plus beta wholebody limit as being taken as the average weekly dose over a three-month period, thus allowing the 

300 mrep to be exceeded in some weeks (AEC 1950d); this was apparently based on the fact that 

Mallinckrodt was already using a system of personnel rotation to reduce doses (AEC 1950b). 

However, AEC at the same time suggested that 150 mrep per week be taken as a recommended limit 

for most purposes and that a weekly dose of 600 mrep be exceeded only in exceptional cases. It 

should be noted that in 1953, a design contractor was stating that the design criteria for ventilation 

and dust control equipment his company had put in at Mallinckrodt and Harshaw included a maximum 

weekly exposure of 300 mR of gamma radiation, with actual design predicated on half that to allow for 

a safety factor in unusual circumstances (Miller 1953).

During the early days of wartime uranium processing, AEC/MED's acceptable levels of exposure for 

the uranium processing plants for dust in air were 500 μg/m3 for insoluble uranium salts and 150 

μg/m3 for soluble salts (AEC 1949). In 1944 MED determined that a standard was needed for 

uranium dusts and adopted the air maximum permissible concentration (MPC) level for lead, 150 

μg/m3

, as the interim standard (Hursh 1975). In 1949, a University of Rochester scientific group 

suggested an air MPC of 50 μg/m3 for soluble uranium forms based on (chemical) injury to the kidney 

and an air MPC for insoluble forms based on radiation injury to the lung (Hursh 1975). In 1953 the 

NCRP recommended in National Bureau of Standards Handbook 52 (quoted in Hursh 1975) a limit of 

73 μg/m3 for both soluble and insoluble forms; it was adopted. These were occupational standards 

that correspond to a 40-hour week (the number of hours that Hursh (1975) uses in conversions in his 

discussion of the history of standards).

By 1949, AEC had set a "preferred level" of 50 μg/m3 for uranium dust, assuming a routine exposure 

of 8 hours a day, 6 days a week (AEC 1949). This was taken to be equivalent to 70 dpm/m3 for alpha 

and is based on animal studies (Hursh 1975). For dosimetry reference, Appendix A gives the basis 

for this figure. This was later referred to as the Maximum Allowable Concentration (MAC) and was 

still in use as of 1953 (Miller 1953). In early 1955, AEC appears to have adopted a MAC of 100 

dpm/m3 for alpha, as AEC (1955c) stated in an air dust study report. In a 1958 report, AEC gave the 

limit for natural uranium, either soluble or insoluble, in air as 5 × 10-11 μCi/ml for 40 hours/wk (i.e., 

occupational) and 1.7 × 10-12 μCi/ml for continuous occupancy (AEC 1958, Table I). In 1959 the AEC 

also recommended against respirator use except in emergency situations (AEC 1949), suggesting 

that before the relevant period of Mallinckrodt work, extensive use of respirators was still tolerated as 

a means of minimizing exposure.
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However, note that in a 1958 paper by an AEC-NYOO safety official, while the wartime MAC was 

given as 500 μg/m3

, agreeing with the information above, the "present MAC" (i.e., in 1958) was said to 

be 110 dpm/m3 instead of the 70 dpm/m3 of other references (Breslin 1958). This may be because of 

the so-called "double curie" or "special curie" uranium radioactive content unit that the NCRP and 

others used prior to about 1 January 1974. The special curie was defined for the natural uranium mix 

and was taken to be 2.05 times the nominal number of disintegrations of U-238 because it included 

allowances for the U-234 and U-235 content (NCRP 1974). If the 110 dpm/m3 was based on a 48-

hour week, this would be equivalent to 65 dpm/m3

, or about 70 dpm/m3, based on a 40-hour week and 

multiplying by 2.05 to convert from the special curie to the regular curie. The NCRP MPC value that 

corresponded to the special curie was given as 6 × 10-11 "special" μCi/cm3

, which agrees with the 

value of 5 × 10-11 given in the paragraph above if the former was based on a 48-hour week. (The 

NCRP abandoned use of this unit in 1973 because it caused confusion.) It does not appear that the 

former use of this unit has caused any disparity or confusion in the interpretation of data used in this 

document, but this potential problem should be noted.

Prior to 1941, there was no standard for occupational radon exposure. In 1941, the United States 

Advisory Committee on X-Ray and Radium Protection recommended 10-11 Ci/L (curies per liter) as the 

acceptable occupational radon level, based on a 40-hour work week (Akerblom 1999). (This would be 

10 picocuries/liter in the units most used today.) This standard was published as a handbook by the 

National Bureau of Standards (NBS 1941). According to Raabe (2002), the standard was for 

ventilated rooms in which work was done with radium (e.g., luminous dial painting) and did not include 

consideration of radon daughter products. An AEC report (AEC 1949) stated explicitly that the 

acceptable radon level for "environmental" air (room or outside air breathed by occupational workers) 

in AEC-sponsored facilities was taken to be 10-10 Ci/L, i.e., ten times higher than the NBS standard; 

the report refers to this level as the " "maximum permissible concentration" " (page 14) and also as 

the "preferred level" (page 20). However, a value of 1 x 10-12 Ci/L was also mentioned as the 

tolerance level in 1950 (AEC 1950i). A ventilation design contractor stated in 1953 that the radon 

design criteria limit for work for Mallinckrodt and Harshaw was 10-8 Ci/m3 of air, or 10-11 Ci/L of air 

(Miller 1953). Thus it is claimant-favorable to assume that the standard that was applied during most 

of the period in which Mallinckrodt uranium refining took place was 10-10 Ci/L (100 pCi/L) and that this 

applied to radon only and not to the daughters.

Regarding effluents, AEC proceeded on the basis that it was acceptable for liquid and solid effluents 

to have concentrations up to one order of magnitude greater than natural background (AEC 1949). 

AEC "recommend[ed] that neighborhood air levels for these radioactive materials [containing uranium 

and radium] should not exceed 1% of the levels used within the plants" (AEC 1949). No information is 

available as to Mallinckrodt's approach to effluent control, e.g., whether it followed the AEC 

recommendation.

2.2 RADIOACTIVITY CONTENT AND HANDLING OF THE ORE, URANIUM PRODUCTS, 

AND RESIDUES

2.2.1

The origin of the ores is important in considering source terms at Mallinckrodt facilities because the 

content of uranium in the ores was greatly variable. Pitchblende ores contained high levels of radium226 and other radiologically significant daughter products, while other ores and feed materials 

typically did not. Ra-226 (in equilibrium with its daughter products) constitutes a significant gamma 

source and thus produced most of the external whole-body dose received by the Mallinckrodt 

workers, while Th-234 and Pa-234, both beta emitters, produced most of the extremity dose. In 

addition, radon and radioactive dusts were released in storage and processing, resulting in internal 

Ore
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dose due to inhalation. The concentration of radium and other daughters present in the ore, 

processed uranium, and processing residue at any given time mostly strongly depended on the 

concentration of uranium in the ore and its radium content. Thus on a per-ton-of-ore-processed basis 

(ignoring process differences), the doses received depended on where the ore originated.

Pitchblende ores from the Belgian Congo (the so-called "Congo ore" or African ore), supplied by the 

Belgium-based African Metals Company, had average concentrations of 25% uranium by weight 

(Eisenbud 1975), up to a maximum of 65-70% (DOE 1997; Dupree-Ellis et al. 2000). (Note that there 

is some confusion among the various references with regard to ore specifications, in that some give 

the percentage as applying to uranium alone and some as applying to U3O8. Since the percentage is 

by weight and the uranium is by far the larger weight constituent of U3O8, the differences are not 

significant.) Other early ore sources were pitchblende ores from Canada (Radium City in the Great 

Bear Lake Area and Port Hope in Ontario) (DOE 1997), containing uranium concentrations of about 

10% (Eisenbud 1975). Later in the war, domestic ores were used also. The wartime domestic 

supplies were actually tailings from vanadium mining and milling (Eisenbud 1975); although the 

original ore had uranium concentrations of less than 1%, the tailings were shipped as a 20% sludge 

concentrate (Eisenbud 1975).

Ores from the Belgian Congo had average concentrations contained up to 100 milligrams of Ra-226 

per ton of ore (Dupree-Ellis et al. 2000; Eisenbud 1975), possibly up to 135 milligrams per ton (AEC 

1949). Thus there could be a significant dose rate from the ore when it was in drums or when it was 

being loaded into other containers and hoppers. As previously noted, the ore was dried before use; 

this was a necessary condition for optimal processing, but the dust levels created during drying and 

later handling were high. The radon was also significant because it built up over time in containers 

and enclosed spaces. When drums, enclosed storage areas, the thaw house, etc., were opened, a 

worker could be enveloped in the escaping radon. Table 4 gives more information about the 

quantities and radiological characteristics of the ore. 

Most of the pitchblende processed by Mallinckrodt was obtained as a concentrate from the Belgian 

Congo in 1944 and was shipped to St. Louis from there in 55-gallon drums (AEC 1978). After the 

war, feed materials were usually packed into 30- or 55- gallon steel drums at the mills and shipped by 

rail in full carload lots (Mason 1958a). It can be inferred from Mason (1958a) that a rough conversion 

is about 15,000 tons of uranium ore per 100,000 drums, so that each ton represented about 7 drums 

to be dumped, sampled, and processed.

2.2.2

Once the Ra-226 was removed following the digestion step and the vessel(s) had been vented, the 

gamma dose rates were much lower and the radon (which arose from the radium) was no longer an 

issue. Radium and radon would again build up from the uranium parent, but this took more time than 

the apparent typical digestion-to-shipout time at Mallinckrodt. 

Uranium Products

The main hazard after the radium-bearing residue was removed was dust, since a fraction of the 

uranium salts and oxides tended to aerosolize when dry and when handled. Initially, somewhat crude 

precautions were taken to control dust during handling and there was extensive manual handling of 

uranium salts and oxides in the dry form (Eisenbud 1975). For example, in Plant 6 (and presumably 

in its predecessor plants 1 and 2), the UO3 dry powder was unloaded from the reaction pots into 

drums by hand-scooping (Mason 1958a), i.e., manually using handheld scoops. The UO3 was then 

moved to the furnace loading area, where it was hand-scooped into trays, which after weighing were 

placed into the furnaces to be reduced to UO2 (Mason 1958a). The trays of UO2 were then unloaded 

by hand into drums for transport to other areas or sites. The major handling improvement of 1949, the 
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installation of pneumatic unloading and conveying systems, was supposed to have eliminated all 

hand-scooping of UO2 and UO3. However, AEC inspectors repeatedly noted hand-scooping going on 

until the end of operations at the plant, often due to the failure of equipment such as the vacuum-type 

UO3 “gulpers” (AEC 1954d; AEC 1956c).

UO2 produced at Plant 6 was trucked over to Plant 4 in "small fiber containers" (AEC 1949); no 

information is given as to how this affected containment of the dust. In Plant 4, there was again 

extensive hand-scooping and other manual handling of the uranium materials (UO2, UF4, and uranium 

metal) (Mason 1958a). This was reduced by mechanization in 1948 and 1949, but even so dust 

levels were considered too high (Mason 1958a). AEC agreed to have Mallinckrodt construct Plants 

6E and 7 to replace it. These plants were even more mechanized and were said to require little (if 

any) manual handling (Mason 1958a); however, as various AEC air dust study reports indicate (e.g., 

AEC 1954g; AEC 1955b), this was not so. AEC (1955c) even reported in 1955 that a Plant 7 operator 

used a piece of cardboard in lieu of a conventional metal scoop to make up UF4, with the operator's 

(presumably gloved) fingers dipping into the material frequently.

2.2.3

A condition placed on the ores from the Belgian Congo by African Metals was that the Ra-226, the 

Ra-226 daughters, and the lead and precious metals be extracted, stored, and returned to African 

Metals (AEC 1978; AEC 1949). Thus the Mallinckrodt process included steps to extract these 

materials as a separate residue from the bulk of the ore residue; this was the radium- and leadbearing K-65 residue.

Residues (Wastes)

As much as 100 grams of Ra-226 contained in the residues was produced per month at the 

Mallinckrodt facilities (AEC 1949). This was in fact the quantity produced in December 1948 (AEC 

1949). It was reported that 200 grams had been transported at one time to Middlesex (AEC 1949), 

meaning that up to this quantity had been in storage at the Mallinckrodt site and had had to be loaded 

for transport at one time. The residues were stored in drums "for health reasons" (FUSRAP 1996). 

Some of the waste sent to the SLAPS waste storage site was hand-packed by Mallinckrodt workers 

into barrels (Eisenbud 1975). Radiation levels from the railcars and trucks exceeded those permitted 

under the regulations of the day, so shipments had to be made under special permits granted by the 

Interstate Commerce Commission (forerunner of the Department of Transportation) (AEC 1949).

As noted earlier, in 1955-1957 Mallinckrodt processed AM-7 residues (see Table 4) and produced 

3600 gallons of a concentrated thorium nitrate solution that was sent to Mound (AEC 1978; FUSRAP 

1996; US DOE 2002). AEC (1978) states that Mound purified and concentrated approximately a 

kilogram of thorium-230 from this material, but DOE (2002) states that although 500 grams was 

produced and an additional 500 g ordered, the latter was apparently never produced. The claimantfavorable assumption will be made that the entire 1 kg was produced. Thus assuming a high 

separation percentage, in the original 350 tons of AM-7 during processing there were approximately 

the one kilogram of Th-230, about 9 kilograms of total thorium, and 0.015 kilogram of uranium. This is 

consistent with the report by Figgins et al. (1962) that in the solution there were 29 ppm of Th-232 and 

3.8 ppm of Th-230, i.e., 11.6% (weight) Th-230 by isotope.

It is not clear how long it had been since the first residue sent in 1946 to the SLAPS waste storage 

site had been produced in the refining process. That is, this type of residue had been produced since 

ether extraction started in July 1942, but it is not clear whether the residues produced between 1942 

and 1946 were all stored at Mallinckrodt and then shipped to SLAPS, or whether some of the older 

residues had been shipped elsewhere. The claimant-favorable assumption will be made that the 

waste was all produced in 1942 and sat in containers for 15 years, until 1957; this will maximize the 
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radium and radon content. As Table 6 shows, although the Th-230 and Th-232 daughters would have 

had some time to build up, only the Th-232 daughters would be nearing equilibrium with the parent. 

The maximum Ra-226 content over the 2.25 possible years of processing (i.e., the 15-year maximum) 

was 0.158 mCi (ignoring decay of Ra-226) and the maximum Rn-222 content was 0.158 mCi (ignoring 

decay of Rn-222).

Regarding the protactinium processing of residues, there is no suggestion in any reference that the 

Sperry press cake was processed in any way at the St. Louis downtown site and the waste appears to 

have been shipped out to Mound directly from SLAPS. Thus, it will be considered that there are no 

radiological implications of protactinium processing associated with the Mallinckrodt downtown site.

2.3 INTERNAL DOSE CONSIDERATIONS

The primary route of internal exposure was via inhalation of airborne particulates, although radon was 

a considerable problem in some areas. The wet operations did not generate as significant a level of 

airborne particulates as the dry operations (Mason 1958a). Thus the operations that were likely to 

produce airborne particulates were those in which the uranium-containing material was dry or was 

heated, although airborne contamination could occur even when the material was not open to the 

room air because some of the equipment leaked (Caplan and Mason 1952). Material that had settled 

on the floor and other surfaces could also be resuspended in the course of operations.

Very little internal exposure data (in dose units) is available. In the early days, measurements of 

airborne concentrations were used per se as indicators of potential exposures and of the need for 

changes in processes and equipment (as suggested by AEC 1950a). Urinalysis appears to have 

been performed as a means of retroactively checking for acute exposures or for the onset of damage 

due to chronic exposure, rather than a dose measurement per se. In addition, as discussed below, 

the particle size and solubility of the various uranium forms were still being investigated 

experimentally by researchers in the field, so that the assumptions that were used at that time in 

calculating internal doses might not be acceptable based on current knowledge. Therefore in this 

technical basis document, potential internal exposure is presented in terms of airborne concentration 

data (and the derived inhaled amounts) and surrogate urinalysis results, rather than internal doses, so 

that dose reconstructors may estimate individual internal doses using modern methods. 

2.3.1

A discussion of the thinking of the time regarding particle size and solubility is included in the 

discussion below in case questions arise regarding typical operations or regarding notes in the 

urinalysis and other records.

Particle Size, Solubility, and Composition Considerations; Sampling Methods

The uranium refining operations at Mallinckrodt produced nominally insoluble uranium compounds, 

e.g., UO2, UF4, and uranium metal (Lippmann 1958, regarding solubility); no patently soluble forms, 

such as UF6, appear to have been produced. However, as some of the literature in the years between 

1958 and about 1975 showed, some supposedly insoluble particles produced in these operations 

seemed to behave like soluble particles (e.g., Lippmann 1958, Heatherton 1975, Archer et al. 1975). 

More recently, ICRP 71 (ICRP 1995a) states that studies of UF4 show behavior consistent with Type F 

in some cases and Type M in other cases, while ICRP 68 (ICRP 1995b) recommends Type M for UF4. 

Yet autopsy data from deceased workers showed a far lower concentration of uranium and thorium in 

lung and other tissues than would have been expected based on the average airborne concentrations 

inhaled. This led Eisenbud (1975) and others to postulate that the particle behavior formulation at the 

time was incorrect: they conjectured that due to the high density of the uranium and daughters, 

particles of respirable diameter behaved like lighter particles of much larger diameters.
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Eisenbud (1958) reported that "It has been shown that in these plants [the uranium processing plants] 

the mass median diameter was about 2 μ [microns]". But he also noted that while the peak for 

alveolar retention is 1-2 μ for dust of unit density (i.e., 1 g/cm3

), uranium oxide dust has a density of 9 

or 10, so that a 1-2 μ particle will behave as though it were a 3-6 μ particle of unit density. An AEC 

report on uranium mills stated that ore concentrates (the form that arrives at uranium refineries such 

as Mallinckrodt) had a "mass medium [sic] size" of 2.5 μ (AEC 1958). Some uranium refinery 

workers were said to have been exposed to UO3 dust with a mass median diameter of 2 μ; however, 

with a density for U compounds in the range of 9-10 g/cm3

, the effective aerodynamic diameter is 

much larger than 2 μ, possibly in the range of 5-6 μ (Eisenbud 1975).

Schwendiman et al. (1975) cited the results of a study by Lippmann and Harris regarding the 

application of size-selective samplers in the uranium industry. Lippmann and Harris performed a 

sampling survey of six different uranium processing plants. They used a filter pair such that the 

second filter was the respirable-particle one, with the first collecting 100% of particles ≥ 10 μm AED, 

75% of those ≥ 10 μm, 50% = 3.5 μm, 25% = 2.5 μm, and 0% of those ≤ 2 μm. They found that less 

than 15% of the airborne material collected on the respirable-particle filter. Schwendiman et al. 

(1975) noted that this agrees fairly well with the ICRP assumption of a size distribution such that 25% 

deposits in the lower respiratory passages and, for insoluble compounds, only half of the 25% (or 

12.5%) would be retained with biological half-life of 120 days: only 10% of the samples in the study 

had more than 25% collected on the second filter. Categories studied included U3O8, UO3, UO2, ore 

concentrate, and mixed scrap.

Because of the lack of specific information regarding particle sizes at the Mallinckrodt plants, the 

ICRP 66 default deposition parameters (ICRP 1994) are to be used to estimate internal doses for 

Mallinckrodt workers. However, where there are particle size issues in specific cases, the information 

above and in Table 7, which gives measured particle sizes for various uranium forms, should be of 

general application even when the listing is for depleted or enriched rather than natural uranium. 

Sanders (1975) also provides a useful general observation that foundry operations (those in which 

uranium oxides are produced) have the highest percentage of airborne particles in the lower lunginhalation range, while reduction operations (UF4 to metal) have the smallest percentage. 

Regarding the isotopic and chemical composition of the dusts, no Mallinckrodt-specific information 

appears to be available. Observations at a large uranium processing mill (probably Anaconda) 

showed that although the potential for dust creation in the dry processing steps was clear, the wet 

processing steps, such as grinding, leaching, separation, and precipitation could create aerosols by 

agitation or by transfer of solutions and slurries. Even so, in the wet processing areas airborne 

uranium was never a major problem, i.e., local hooding and exhaust ventilation were not required in 

order to keep the dust concentration below allowable levels. The physical characteristics of the 

airborne particulates in the dry and the wet areas were thought to be similar; thus it was assumed that 

nearly all of the mass of the particle was siliceous material, clay, or a mixture of the two. The mass of 

the particles was found to be less than one percent of the total mass. From specific nuclide analyses 

done over several years, they concluded that secular equilibrium existed at least through Ra-226 in 

the ore at the time of mining. Specific nuclide analyses of particles taken in various areas of the mill 

showed that the U-238:Th-230:Ra-226 ratio was 1:1:0.3 and was fairly constant regardless of the 

source of the aerosol or dust. But in the yellowcake section of the mill, where the ore had been 

processed to become mostly U3O8, the ratio was 1:0.01:0.001, indicating that these products were 

removed in the processing (Wilde 1975).

The various AEC air dust study reports state that they are following an established AEC protocol, but 

the documentation for what the protocol was is lacking. However, some information is known about 

Mallinckrodt's methods. For the data reported by MCW (1949b), dust samples were collected on 1-
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1/8" Whatman #41 filter paper disks, using a modified Fischer pump, at 0.5 ft3/min, over a period of 45 

seconds to 30 minutes depending on conditions and job time. The disks were counted on a parallel 

plate alpha counter such that statistical variations would be no more than 10% at the 0.9 confidence 

level. It was stated that a few of the low-level samples might have errors of 20%. AEC (1950a) stated 

that it was using 1-1/8” Whatman filter paper disks and a handheld air sample with a collection rate of 

15-20 lfm.

From reports of sampling at other sites and from unspecific mentions in papers and reports about 

Mallinckrodt, it appears that typical practice was for the air to be drawn through filter paper (most likely 

Whatman #41, widely used in the time period) and counted on an alpha scintillation counter (e.g., the 

laundry samples of Utnage 1958b). In 1958 at the Oak Ridge Gaseous Diffusion Plant, the Whatman 

#41 was used with fixed counting equipment, for a counting and collection efficiency of 30% (Becher 

1958). AEC in-plant air sampling was done by collection on Whatman #41 filter paper and counting of 

total alphas; a correction for self-absorption in the filter paper was applied (Eisenbud 1975). The 

results were reported as alpha dpm/m3 (Eisenbud 1975). The AEC, in sampling stack and environs 

air at various sites including Mallinckrodt, used Whatman #41 filter discs with "standard sampling 

equipment and techniques normally employed by HASL" and counted them on scintillation counters 

(Weinstein 1958). Since the methods of the time seem to have been fairly uniform (with HASL setting 

the standard), it is assumed that the AEC and Mallinckrodt measurements were taken consistent with 

these references.

The AEC's HASL staff was committed to the time-weighted average as being most representative of 

total exposure. As Glauberman and Harris (1958) put it, "HASL has found from experience that the 

multiple-sample time-weighted average exposure procedure is the most accurate....The GA [general 

area] sample normally will tend to underestimate an operator's exposure and the BZ [breathing zone] 

sample to overestimate it, but by time-weighting the average concentrations for both types of samples 

an operator's exposure may be closely evaluated...[this method] yields reasonably reproducible 

results". In this technical basis document, time-weighted averages will be preferred to static 

measurements or short-term maxima as being representative of worker exposures. These timeweighted averages are documented in a series of AEC air dust study reports (e.g., AEC 1954b) and in 

some Mallinckrodt reports (e.g., MCW 1949b).

2.3.2

Eisenbud (1975) stated that "above all other types of exposure, it was the airborne alpha-emitting dust 

that was the cause of greatest concern." This was particularly true in the sampling and crushing of 

ore and in the mechanical and manual handling of dry uranium salts and oxides (Eisenbud 1975). 

Dupree-Ellis et al. (2000) stated that daily average uranium dust concentrations of up to 100-200 

times the maximum allowable concentration of 50 μg/m3 were measured in poorly ventilated 

processing areas. An industry-wide survey showed the average concentration to be 5000 dpm/m3

, 

which an AEC-HASL official stated was "a conservative estimate of the levels that existed from 1942 

to 1948" (Harris 1958). 

Airborne Dust Levels

Mason (1958a) stated that while no regular dust sampling program was in effect at Mallinckrodt during 

1943-1947, enough samples were collected to show that concentrations were high by 1958 

standards, that concentrations of 50 to 100 times the MAC level of 70 dpm/m3 were not uncommon, 

and that some operations produced concentrations up to 1000 MAC for a few minutes at a time. 

According to the AEC, many employees were exposed to elevated dust levels for years (AEC 1949). 

In 1946-1948, respirators were said to be required "for practically all plant operations" (Mason 1958a) 

but respirators were not used consistently (see Section 5.3.3 below).
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Major improvements in dust control and ventilation were made at Mallinckrodt in 1949 under the new 

AEC health program, such as the installation of pneumatic unloading and conveying equipment in 

Plant 6 process areas that eliminated most hand-scooping and thus that mode of exposure to dust 

(Mason 1958a). However, while there was a marked reduction in dust levels, the improvement was 

not what had been hoped for in some areas, especially with respect to the handling of the UO3

(Mason 1958a). In 1953, for example, an AEC inspector reported that there were many small 

openings between the operating area an the drum storage “alleys” in the Ore Room addition; as a 

result, winds blew into the area and upset the ventilation air balance, causing dust to be blown into the 

operators’ area (AEC 1953). Even the most modern plant, Plant 7, had dust problems: in 1954 an 

AEC inspector noted that there was a fine film of UO3 on supposedly clean drums and as cans of UF4 

were transferred from a hooded enclosure to a conveyor, a green dust cloud could be seen to escape 

from under the lid of nearly every can (AEC 1954f).

Similarly, in 1948 and 1949, much of the manual handling of UO2, UF4, and uranium metal in Plant 4 

was eliminated, but dust concentrations did not get down to satisfactory levels (Mason 1958a). Even 

the building of the new Plant 6E and Plant 7 did not completely eliminate the problem: the uranium 

was never contained well enough that it ceased to create airborne levels of concern in the plant air 

and in the (multi-building) plants in general (Mason 1958a). A Mallinckrodt official remarked that it 

had been a constant battle to keep airborne levels at 1 MAC or less (Mason 1958a). An AEC safety 

official speaking of workers at all the AEC uranium-refining plants (Breslin 1958) reported that even in 

1951 approximately half of the workers were exposed to average concentrations above the MAC then 

in force (which he said was 110 dpm/m3

, although AEC (1954f) indicates that it was 70 dpm/m3 until at 

least the end of 1954). He also stated that in 1956 6% still were above the MAC; the percentages for 

those exposed to average concentrations greater than 1800 dpm/m3 were 4% and 1% respectively. 

Finally, he noted that while airborne control in these plants was largely achieved by 1955, there had 

been a retrogression resulting from a large increase in production volume.

Tables 8 through 11 give a trend overview of airborne uranium concentrations measured over the 

years at Plants 4, 6, 6E, and 7 (Mason 1958a). The concentrations are given as multiples of the 

"preferred level" at the time of measurement (1948), i.e., multiples of 70 dpm alpha per m3

. It appears 

from the reference that they represent typical concentrations rather than maxima. In AEC’s 

measurements, they did not separate the uranium and radium components of the alpha activity, i.e., 

they counted gross alpha (AEC 1950a). This was Mallinckrodt’s practice as well (MCW 1950a; MCW 

1955-2/21/55); also, beta-gamma activity analysis was not routinely done (MCW 1955). Thus Tables 

8-19 must be assumed to be gross alpha measurements, whoever took the measurements. There 

were no continuous air monitors at the Mallinckrodt site because the results were thought to be of 

doubtful value (MCW 1955).

Table 12 presents the results of airborne dust surveys made in Plant 4 by, respectively, AEC-NYOO's 

Medical Division in May 1948 and Mallinckrodt safety officials in September-October 1948. These 

results are given as time-weighted daily average levels (called DWE levels by AEC) of dust exposure 

by plant. Table 12 shows that AEC and Mallinckrodt's data were in general agreement, although 

there were some differences. In this technical basis document, AEC's data are used preferentially 

because AEC set the standard of measurement for the uranium processing sites and because AEC's 

figures for the most exposed workers are typically higher than Mallinckrodt's. Thus using AEC's 

numbers is claimant-favorable, in general. The AEC data is mostly from a series of dust studies that 

AEC did on a nearly annual basis from about 1948 on. This is presented, with a few additions from 

Mallinckrodt reports to fill gaps, in Tables 13 through 18.
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Some detailed information is available about the particular case of the laundry workers, circa 1958 

when the laundry had operated for at least ten years (Utnage 1958b), as shown in Table 18; 

information was also found in the various AEC dust study reports.

Information from other sites is helpful in deducing what would be typical at Mallinckrodt. In an AEC 

report in 1958, the breathing zone concentrations in the final ore concentrates packaging areas of 

over a dozen mills were evaluated; the concentrations ranged from 0.01 to 5.5 pCi/L, with a median of 

about 0.15 and a mean of 0.91 (AEC 1958, Table V). These figures suggest levels that might be 

encountered by Mallinckrodt workers unloading the packages (drums) at the beginning of the refining 

process. See also Section 5.3.4 below regarding resuspension of surface contamination. 

Resuspension contributions are assumed to be included in all data cited in this technical basis 

document since dust levels were typically measured while work was taking place.

The reported air concentrations generally pertain to those workers directly and continuously involved 

in uranium refining work. However, Breslin (1958) defines "auxiliary workers" as workers "not directly 

connected with production but located in or near production buildings,....[including] chemists, 

engineers, office workers, garage mechanics, outside maintenance personnel, and the like", noting 

that some of these had occasion to visit production areas in the course of their work while others did 

not. Even with the limited access, in 1948 about 13% of the auxiliary workers studied were exposed 

to average concentrations above the MAC (110 dpm/m3

 at that time) and more than 1% to 

concentrations greater than 440 dpm/m3 (Breslin 1958). By 1954 none of these auxiliary workers 

were exposed to average concentrations above the MAC (Breslin 1958), although some of the 

process workers still were. These statements are illustrated by the data given in the various AEC dust 

study reports.

Relatively high potential for dust exposure applied not only to those actually present in the dusty 

buildings, but to those working elsewhere in the plants and even outside the plants. For example, a 

Mallinckrodt safety official remarked that one reason to revise the ventilation in about 1952 had been 

that a study of plant effluents showed that "large bursts of dust found their way outside of the plant 

immediately after filter cleaning" (Harris and Mason 1953). Mason (1958a) also suggests that colocated (nearby but uninvolved) workers were exposed to elevated airborne levels. 

In later years, Mallinckrodt was supposed to sample stacks at least once a year, but it was not being 

done (MCW 1955). Weinstein (1958) reported on an air sampling study that AEC-HASL did of stack 

and environs (outside) air at various sites, including Mallinckrodt, in November 1949. They did not 

take any stack samples at Mallinckrodt, but indicated that previous data implied a probable average 

rate of emission of uranium from the Mallinckrodt stack(s) of about 0.011 g/sec, with a flow rate of 

about 20,000 cfm. About 52 tons of uranium as metal was estimated to have been discharged in the 

stack effluents since the beginning of operation. While nearly every reported (outside) concentration 

at Mallinckrodt was below maximum permissible levels, it was observed that "1000 feet would 

circumscribe the MAC" level (out from the plants), i.e., within a few hundred feet of the plant(s) the 

MAC might be exceeded. The MAC given in the version of 10 CFR 20 in force at the time was 1.7 × 

10-12 μCi/ml air for continuous exposure, or 2.5 μg/m3

, which presumably was the "public" MAC and 

not the occupational MAC. This would help explain the somewhat elevated weighted average 

concentrations even for workers who did not enter production areas.

Based on the information in the references cited in this and previous sections, Tables 20 and 21 were 

set up to help dose reconstructors interpret claimant submissions and the Mallinckrodt records. Table 

19 lists job titles obtained from AEC and Mallinckrodt reports and from film badge and urinalysis 

records; in addition, it gives (1) a code(s) corresponding to Table 20 and (2) a geometry factor set 

appropriate for each job title (for later use as described in Section 7). Table 20 gives a set of codes 
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that establishes the correspondence between a process and a type of work or functions; this is to 

assist the dose reconstructor in determining (1) the appropriate job title if claimant information is not 

clear and (2) the appropriate breathing rate to apply, as discussed in Section 7.

Tables 19 and 20 are to be used with Tables 21 through 24 to help determine the exposure to an 

individual worker when bioassay data for the worker is missing or is conflicting and when comparable 

worker bioassay data (see Section 6) is insufficient. Tables 21 through 24 were derived from Tables 

13 through 17 and other sources; the data they contain is nearly all average daily weighted air 

concentrations. All of this data is based on a natural uranium mixture, with one exception, as 

discussed below.

The process in which AM-7 residue was converted into a concentrated thorium nitrate solution, as 

explained in Section 4.5, is a special case. No information was found regarding the particle size of this 

residue, its tendency to be aerosolized, etc. However, the digestion and extraction process appears to 

be similar to the basic uranium ore digestion and extraction process and so it presents no novel 

operational or processing features. Besides the AEC-measured data given in Table 17, there is some 

data from the Mound end of the processing (Mound 1956): the maximum and average air 

concentrations in the ionium (Th-230) "high-risk" part of the Mound processing area were 48.1 x10-10

and 16.3 × 10-10 μCi/cm3 respectively. In the exhaust line of the hood in which the work was done the 

maximum and average concentrations were 384.9 × 10-10 and 38.8 × 10-10 μCi/cm3 respectively. It is 

not known whether the work at Mallinckrodt was done in a hood and since the nature of the Mound 

work was further acid digestion and extraction (DOE 2002), the Mallinckrodt exposures were 

potentially higher.

Also, it must be noted that the concentrations reported by Mound (1956) were of ionium, i.e., Th-230, 

not of total thorium. Thus the associated source term must include the Th-232 known to have been in 

the solution as well. From Section 5.2.3, there was 11.6% Th-230 isotopically in the original residue, 

hence the Th-230/Th-232 ratio will be assumed to be 0.116/0.884, or 0.131. This gives 1.61 × 10-12

μCi/cm3 of Th-232 to be added to the source term (see Table 6). However, because Th-232 behaves 

like Th-230 in the body, both are alpha emitters, and both have very long half-lives compared to the 

human lifespan, the contribution of the Th-232 to the dose from the (total) thorium source will clearly 

be negligible (the dose per unit intake for Th-232 is higher than for Th-230). Thus only the Th-230 

exposure need be considered and the source term in Table 24 for Plant 7E applies to Th-230 only.

2.3.3

An undated AEC reference that is assumed to be of 1942-1944 vintage (MED undated a) stated that 

respirator use is mandatory when required and was especially necessary for the grinding and sifting 

operation. An AEC/MED reference from May 1945 shows layouts indicating the areas of required 

respirator wear (MED 1945b). Mason (1958a) stated that MED and Mallinckrodt agreed in the early 

years of the work that production would proceed on a priority basis, with the understanding that in 

high dust areas extensive use of respirators would need to be made. Thus during 1946-1948, 

respirators were used for “practically all” plant operations. Thus it is clear that respirators were used 

from at least the late war years on. Mason (1958a) commented that the exposures received 

depended partly on the effectiveness of the respirator program (at that time).

Respirator Use

Mallinckrodt’s policy in the later years was that routine respirator use was not acceptable practice and 

that they were a temporary expedient for unusual conditions only (MCW 1955). The requirements for 

them were spelled out in standard operating procedures (MCW 1955). Even so, AEC inspectors 

noted instances where visible dust clouds were present or they measured clearly significant dust 
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levels, yet the operators were not wearing respirators ((AEC 1954b; AEC 1954c; AEC 1954e; AEC 

1954f; AEC 1954g; AEC 1955c).

A Mallinckrodt official observed in 1958 that in the plants they tried to keep below the MAC, but that if 

the concentration were greater than 1 MAC for a specific operation, they would not necessarily require 

the operator to wear a respirator (Utnage 1958c). He explained that this would depend on the 

worker's integrated exposure, taking all operations into consideration, and that if some short-term high 

alpha concentrations in the air were found, they made it a practice to have personnel wear respirators 

temporarily until the situation was corrected. He concluded that Mallinckrodt did not subscribe to the 

use of respirators as standard control equipment. Breslin (1958) states that the time-weighted 

average exposures measured by AEC-NYOO did not include corrections for respirator use and so 

should be viewed as potential exposure; however, he also asserts that in very few cases would these 

be substantial overestimates "as the use of respirators was inadequate and spotty". This is borne out 

by the various AEC dust studies that, as noted above, pointed out cases of significantly elevated dust 

levels where respirators were not being worn.

In the absence of any firm figures on respirator use and efficacy at Mallinckrodt, it will be assumed 

that respirators were not reliably used.

2.3.4

As noted earlier, radon levels could be significantly elevated in enclosed areas where material 

containing uranium daughter products was stored. Radon concentration in enclosed spaces in 

refineries produced levels of up to 10-8 to 10-7 Ci/L, which were eventually significantly lowered by 

ventilation improvements (Eisenbud 1975). After these improvements were made at Mallinckrodt in 

about early 1949, all Mallinckrodt operating areas in Plant 6 were found by AEC to be below the 

preferred level of 10-10 Ci/L, with the exception of the K-65 return oven, the ore thaw house, and the 

"wash Oliver" cell, as shown in Table 25, which covers the years 1947-1957. Most of the information 

comes from the weekly “M Z” radon reports (MCW 1955; MCW various). Additional data is shown in 

Table 26, which gives typical radon exposures for different areas at the Middlesex facility over the 

years 1944-1949 (which cover the early years of pitchblende use); these would be comparable to the 

railcar unloading and Ore Room operations at Mallinckrodt. Table 25 also gives what little information 

has been found about the radon levels in the plants other than Plant 6. However, because in the plant 

with the greatest exposure potential (i.e., Plant 6) the levels were low in all but very enclosed areas 

and because the thorium and radium were removed in the production of UO3, it can be assumed that 

the radon levels were negligible in areas where ore and pre-UO3 residues were not present.

Radon

Regarding the processing of residues to concentrate thorium, it is clear from the data in Table 27 that 

the average concentration of radon (Rn-220 and Rn-222), under process-applicable and claimantfavorable assumptions, is well below the AEC "preferred level" of 10-10 Ci/L. It is also usually less 

than the 3.0 x10-12 Ci/L that represented an "undetectable" amount for the AEC during the relevant 

period of time (as per Eisenbud 1975).

For use in dose reconstruction, the radon concentrations in units of Ci/L may need to be converted to 

units of working level months (WLM) before they can be used in calculations. One working level (WL) 

is the total amount of energy given off over a long period of time by the short-lived radon-222 

daughters in equilibrium with 100 pCi (10-10 Ci) of radon, taken to be in one liter of air. Since the 

daughters will typically not be in equilibrium with the radon if the ventilation is good, this conversion is 

not simple, but depends on the ventilation conditions. The claimant-favorable assumption will be 

made that the daughters are in equilibrium with the given concentration of radon. (Note that Applied 

Nuclear Safety (1986;1990) assumed a 50% equilibrium for post-operations radon measurements in 
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the Mallinckrodt buildings.) Thus 10-10 Ci/L of radon will be assumed to be equivalent to 1.0 WL. The 

WLM is 170 WL-hours, i.e., 1.0 WL breathed in for 170 working hours per month. The annual radon 

exposure can thus be taken to be the radon concentration in units of 10-10 Ci/L times 1.0 WL times 12 

(months) times the fraction of the annual working hours that exposure occurs. Thus a radon exposure 

of 6 x10-10 Ci/L, occurring 2 hours a day, would be equivalent to 6 × 1 WLM × 12 × .25, or 18 WLM, on 

an annual basis.

2.3.5

Although surface contamination levels per se are not indicative of airborne contamination levels or 

external dose rate, they can suggest whether or not a potential for exposure exists. MCW (1958) and 

MCW (1959) give the results of a plant-wide surface contamination and external dose rate study that 

Mallinckrodt did after some post-operation decontamination of the site. These reports indicate that 

ground areas adjacent to the production plants were heavily contaminated, with average surface 

alpha activity of 2500 dpm/100 cm2 and average beta activity of 2 mrep/hr; high spots of 35,000 dpm/ 

100 cm2 alpha and 15 mrep/hr beta were not unusual. High spots of up to 20 mR/hr gamma were 

found west of the UF4 loading dock where ore cars where cleaned prior to 1950. In Plants 6 and 7, 

the administration building (Building 112) (except for the storeroom and maintenance shop), the 

Boiler House (115), the Service Building (117) (except for the laundry and the regulated locker room), 

and the Magnesium Building (708) were not significantly contaminated. The average alpha activity on 

floors in the excepted areas was 3000 dpm/100 cm2 and the beta-gamma activity was generally at 

background levels. In Plant 4, beta activity measured at contact with surfaces in production areas 

averaged between 10 and 50 mrep/hr with the overall average activity measured at contact with 

surfaces estimated at a level of 25 mrep/hr; occasional high spots were found up to 80 mrep/hr; 

activity measured at the three-foot level in the center of production areas ranged from 1 to 5 mrep/hr. 

Gamma activity measured at contact with surfaces was 0.02-0.9 mrep/hr, with the highest readings 

being in some yard areas and the slag processing area. The overall plant average measured at the 

three-foot level in the center of production areas was 0.07-0.1 mR/hr.

Surface Contamination

Data from this Mallinckrodt study (i.e., MCW 1958 and MCW 1959, quoted by Utnage (1958a)), 

showed fixed floor surface contamination levels of greater than 10,000 alpha dpm/100 cm2 at various 

floor locations in the ore sampling area after vacuuming; it was highest at the hopper loading and 

weighing stations. In the metal reduction area (Plant 6E), the floors of the crucible disassembly areas 

and the saw areas measured 2200-3300 dpm/100 cm2 and the center of the maintenance cage 

(where there was no uranium processing) measured 1200 dpm/100 cm2

. Other stations in the 

"pitchblende (ore) area" (Plant 6) showed levels of 1000-60,000 dpm/100 cm2

; in the uranium 

products warehouse, 4500-21,500 dpm/100 cm2

; and in the metal plant (6E), 1000-9000 dpm/100 

cm2

. The UF4 production plant (Plant 7) had no reading higher than 1200 dpm/100 cm2. Since these 

figures represent fixed contamination, loose contamination removed by the vacuuming could have 

been present at far higher levels (Brobst 1958). Thus these figures, while suggestive of where the 

worst areas were, cannot be considered to be representative of operational total surface 

contamination levels. 

Although as noted above resuspension is assumed to be included in the measured airborne 

concentrations, some available information about the relationship between loose surface uraniumbearing contamination and airborne concentrations will be summarized here in case it becomes 

relevant in individual cases. A study was done at the Oak Ridge Gaseous Diffusion Plant to ascertain 

the relationship between alpha airborne activity and alpha surface contamination (Becher 1958). The 

air samples were measured with Whatman #41 filter discs and fixed counting equipment, with an 

overall counting and collection efficiency of 30%. The surface transferrable activity was measured 

with a Samson alpha survey meter, with an overall counting and efficiency factor of 20%. The data, 
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shown in Table 28, indicate that the airborne concentrations ranged from 0.36 to 5.05 dpm/m3 for 

every dpm/cm2 of surface contamination, or an air concentration of 3.6 to 50.5 dpm/m3 for every 1000 

dpm/100 cm2 of surface contamination. Other fixed surface contamination data is given in Table 29

for various surfaces and pieces of equipment in the laundry (from Table 1 of Utnage 1958b).

With surface contamination there is the potential for two modes of exposure other than inhalation. 

These are ingestion and skin doses from contamination on skin and clothing. Ingestion would most 

likely take place during eating or smoking breaks. No information is available as to the likelihood of 

ingestion during eating or the quantities that might have been taken in. However, some information is 

available on ingestion as a result of smoking. Tests done in about 1958 at the Oak Ridge Gaseous 

Diffusion Plant with UNO4, UF2, and UF4 (Bailey 1958) indicated that when loose uranium-bearing 

material was placed on the palmar surfaces of the hands, the palmar transfer of uranium from the 

hands to the cigarette amounted to about 1% of the material on the hands and that inhalation of the 

material during smoking amounted to less than 1%. They also tested absorption by placing material 

on the backs of the hands. They concluded that a maximum of 2.5 × 105 dpm on the palm of each 

hand, for a 20-cigarette-a-day smoker, would be allowable for him not to exceed the maximum 

permissible inhalation of uranium by cigarettes alone; that a maximum of 2.5 × 105 dpm on each hand 

would correspond to the maximum permissible ingestion; and that a maximum of 2.5 × 105 dpm on 

the back of each hand would correspond to the maximum permissible absorption dose. Taking the 

reciprocal of the sum of the reciprocals, they obtained a maximum of 8.6 × 104 dpm per hand for the 

total considering all three routes of exposure. They assumed a 15% geometry factor for the hands 

(based on their instruments) and concluded that the total limiting level was 13,000 cpm per hand.

The allowable amounts given in Bailey (1958) corresponded presumably to National Bureau of 

Standards Handbook 52 allowable intakes, since they cited Handbook 52 although they did not quote 

the figures. The allowable amount was assumed to be on the hands continuously for 5 days per 

week, 8 hours per day. This level or greater is likely to have been on workers' hands at times but not 

to have been on the hands constantly. Thus while a potentially significant contribution to the dose 

from hand contamination cannot be ruled out, it seems unlikely that most smokers would have had a 

sustained level of contamination of this magnitude on the hand during breaks, especially since they 

were likely to have worn gloves during most of the processes (due to heat, acidic content, etc.). 

Hence it will be assumed that ingestion can be ruled out as a major source of internal dose compared 

to inhalation.

Also, AEC (1951) observed that in a majority of the jobs evaluated at Mallinckrodt, it was found that 

the contribution to dust levels from time spent in the smoking (cigarette break) areas was greater than 

the contribution of dust from the time spent in the operations area; this was because the smoking 

areas were typically immediately adjacent to the actual work areas. This also suggests that the 

contribution to the ingestion of radioactivity by a smoker might be significant, although AEC apparently 

did not do any studies of this at the time.

Regarding clothing, Table 30 shows contamination levels and some associated dose rates from 

clothing (Utnage 1958b). The contamination measurements in cpm were taken with a Thyac betagamma meter with thin-wall tube, while the measurements in mrep/hr were taken with an unspecified 

air ionization chamber. At the time the measurements were taken in 1957, the laundry had been in 

operation for ten years and had never been decontaminated; there were nine laundry workers; and 

the laundry processed 25,000 coveralls and 25,000 "soft" items (handkerchiefs, socks, and 

underwear) each month. Clothing used in contaminated or potentially contaminated areas ("regulated 

areas") was kept separate from clothing used in non-regulated areas. However, regulated-area 

clothing was worn interchangeably by anybody, so that "a uniform contamination level [was] 

eventually obtained" (Utnage 1958b).
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At the Oak Ridge Gaseous Diffusion Plant (ORGDP), test measurements in 1957 on clothing showed 

that the highest spot reading was typically about 3.5 times the average reading (Becher 1958); this is 

probably roughly applicable to Mallinckrodt as well, although the uranium compounds at ORGDP were 

mostly soluble whereas the Mallinckrodt compounds were mostly insoluble. The ORGDP tests also 

showed 1620 alpha dpm/cm2 to be equivalent 9700 cpm/100 cm2 as measured on a Samson alpha 

meter, giving an "efficiency-geometry" factor of 6%. (Note that for the "surface transferable activity" 

on filter paper used for air sampling also reported in Becher 1958, an efficiency-geometry factor of 

20% was assumed for the Samson alpha meter.) Finally, the ORGDP measurements showed that 

about half the uranium applied to the clothing at the beginning of the test had dropped off within the 

first two hours of wearing. This suggests that uranium that gets on clothing can come back off it 

readily and that surface contamination on clothing can contribute to airborne levels via resuspension.

Railcar interiors were invariably found to be contaminated above 2500 dpm/100 cm2 after unloading 

uranium oxide, UF4, or uranium metal at uranium processing plants, even though the sites made an 

effort to decontaminate them (AEC 1949). AEC advised that strippable coatings would eventually 

need to be used (AEC 1949), but there was no evidence that this was ever done. This suggests that 

even where closed containers of uranium-bearing materials were being unloaded, it must be assumed 

that surface contamination was typically present.

2.3.6

Mallinckrodt uranium processing workers were given a pre-employment physical that included an 

initial urinalysis and a blood count and they were given an annual physical that included a routine 

urinalysis and a blood count (MCW 1955; Mason 1958a). From about the summer of 1948 on, this 

included a measurement of uranium in the urine. In addition, up to March 1954 some worker 

classifications had more frequent urinalyses, either every 4 months or every 6 months depending on 

the worker classification (MCW 1955; Mason 1958a); after this, time the frequency was no more than 

semiannual (MCW 1955). As urinalysis records indicate, some office workers appear to have been 

given urinalyses, but it is not clear whether this was done on a regular basis.

Information and Available Data Regarding Urinalyses

The radiological analysis was apparently only for uranium content (referred to as "X in urine" or 

“uranium-in-urine”). It is not clear how the urinalyses were done, but Ross et al. (1975) states that for 

all AEC contractors before 1961, estimates of lung dose were made on the basis of urinalysis and that 

this was usually done on the basis of electrodeposition and subsequent counting.

The urinalyses were performed by AEC-NYOO from about 1948, when Barnes Hospital at 

Washington University (St. Louis) began to do them. However, an AEC health official stated (AEC

1948) that it was his understanding that the analyses were being done at Barnes Hospital (at 

Washington University), but it turned out that they were being done in laboratories at the Mallinckrodt 

St. Louis site. This came to light when it was discovered, apparently in late 1947, that some urinalysis 

samples were contaminated due to contamination in the laboratory. An undated, untitled urinalysis 

listing found in dose reconstruction project files indicates that closed, blank samples were found to 

have significant levels of uranium in them, indicating contamination in the laboratory; it was suggested 

that this might explain the high levels of some of the non-blank (worker) samples. Thus at least the 

early urinalysis samples must be considered to be potentially contaminated (i.e., some of the uranium 

content may have come from the laboratory analyzing them).

Apparently Barnes Hospital resumed doing the urinalyses (MCW 1950b). However, in 1949, AEC 

compared the Mallinckrodt analyses against those for other sites handling similar material and 

concluded that the results were consistently high (MCW 1950b). They then sent Mallinckrodt some 

spiked samples and also had an independent Mallinckrodt party prepare a stock solution of known 
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concentration. The spiked samples and samples of the stock solution were sent to Barnes as regular 

samples, while Barnes standards and samples of the stock solution were sent to NYOO for analysis. 

AEC also compared Barnes methods and equipment. The conclusion was that the samples were 

indeed reading high at Barnes. Subsequent data analysis showed a gradual precipitation of uranium 

in the Barnes standards, which meant that the daily standard curves showed a gradual loss of slope 

over time, up to 30%. Also, Mallinckrodt had been called three times over the previous year to service 

the Barnes instrument because of sensitivity loss, when the problem was actually the standard. The 

maximum error in the urinalyses over the preceding 14 months was estimated to be +89%.

AEC technical personnel thought that the affected data was of doubtful value (AEC 1950j). Still, AEC 

(1950c) asked Mallinckrodt if the urinalysis data could be salvaged, i.e., if there was a consistent 

factor that could be applied to all of the subject urinalyses; MCW (1950c) thought that there was not. 

AEC also recommended that a note regarding the problem should be inserted into the medical files of 

the affected individuals (AEC 1950b), presumably to aid in the interpretation of the results. It is not 

clear whether this was done or not. AEC-NYOO stated that it was not possible for them to take over 

the urinalyses again since the number of samples to be analyzed was too high for their capacity (AEC 

1950j).

It is not clear who did the urinalyses from 1950 to 1954, although MCW (1950d) suggests that this 

was no longer being done at Barnes Hospital but at AEC-NYOO. In 1954 AEC gave Mallinckrodt 

permission to perform their own urinalyses (MCW 1954), presumably in the laboratories at the 

Mallinckrodt St. Louis site. From the Mallinckrodt Health Office monthly reports, they were analyzing 

for “X in urine” (i.e., uranium) and it appears that at times there was a significant backlog of overdue 

analyses, at least in the early 1950's. Eventually AEC-NYOO must have resumed performing the 

analyses, because a 1955 Mallinckrodt description of its health program stated that NYOO was doing 

so (MCW 1955). This report states that NYOO was analyzing about 2500 Mallinckrodt urine samples 

a year and that urine sample that were taken were split, with half going to AEC-NYOO for the 

radiological analysis and half to Barnes Hospital for the medical analysis.

Because of these questions regarding the validity of the samples and the variations in sample 

analysis methods, the Mallinckrodt urinalysis data should be used with care. However, it appears that 

the errors, if any, are in the conservative direction and thus are claimant-favorable.

Urinalysis records appear to be available, but many appear to be handwritten notes on cards. These 

are found scattered in various dose reconstruction project files. Fortunately, in about the 1970's, the 

records were entered into a computer data base for research purposes and have been used in that 

form since then by Oak Ridge Associated Universities and other research groups. The resulting file 

has more than 40,000 records (i.e., lines, with each line representing one urinalysis). A "stripped" 

version (ORAU 2003) is also available, with the names and Social Security numbers removed for 

privacy reasons.

The large stripped urinalysis file (ORAU 2003) was reviewed for comparable or surrogate worker 

cases that could be used to produce a table of intakes applicable when bioassay data for an individual 

is missing or spotty. Cases were selected on the basis of their containing a reasonably uninterrupted 

series of urinalyses and having reasonably clear notations of job title and/or area worked in. The 

selected cases were then further evaluated and a subset was extracted for each of various categories 

of identified locations and operations or position titles. These categories are given in Table 31. Where 

there were more than two applicable cases for a category, the IMBA program was then used with the 

assumptions of chronic intake and Type M form with the data from these cases to produce a categoryspecific distribution and standard deviation for the typical daily intake, as shown in Table 31; 

otherwise, the actual data (i.e., for one or two cases) was given. Since there were changes in 
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exposure potential at various times due to process improvements, engineering modifications, or the 

building of new plants, three periods were established for the determination from the cases of the 

typical daily intake, as given in Table 31. 

Table 31 does not include intakes for workers who processed wastes containing thorium (ionium); 

estimated annual intakes for these workers are given in Table 32. It appears that there were few such 

workers: (1) AEC (1955c) reports studying only six workers in their Plant 7E dust exposure study 

whereas, e.g.,119 were studied in Plant 6E, (2) few worker cases were found in the large stripped 

urinalysis file where it was clear that the worker did this type of work, and (3) the thorium worker in 

every case that was found had worked in other areas where an intake of uranium material was 

possible. It is not clear that thorium itself in urine was measured at all, rather than, say, gross 

radioactivity. Thus there likely was not any differentiation in the urine analyses between uranium (and 

its daughters) and thorium. Hence in dose reconstruction the urinalysis data from mid-1955 on for 

workers who processed thorium wastes should be generally be assumed to consist of whichever 

source set (U-234, U-235, and daughters or Th-230) gives the more conservative result.

2.3.7

Breath radon measurements began to be made in 1947, but only for workers who worked in areas 

where there was a potential for radium intake (AEC 1950a). Breath radon samples were sent to AECNYOO for analysis; in 1955, AEC-NYOO was analyzing about 500 Mallinckrodt samples a year, taken 

semiannually at a minimum but about quarterly when permitted by AEC-NYOO sample capacity 

(MCW 1955). Breath radon records are available in scattered form in reconstruction project files.

Information and Available Data Regarding Other Types of Bioassay

Breath radon samples were collected by obtaining one-liter samples of exhaled breath after two days 

of non-exposure, usually on a Monday morning (AEC 1950a); MCW 1950f). The samples were 

measured at NYOO by an “automatically recording pulse-counting device” (AEC 1950a). In early 

1950 AEC apparently became concerned about the high background that seemed to be present 

where the samples were being taken (MCW 1950f). Mallinckrodt agreed to take test samples 

elsewhere than in their usual testing area and also to take a room air sample in their normal testing 

room; these samples were then sent to AEC-NYOO. The normal testing room sample showed a 

radon content of 0.8 x 10-12 Ci/L, which AEC judged was a high background for a breath radon 

testing area (AEC 1950i). It is known that in 1950 the Mallinckrodt medical department was located 

adjacent to the change rooms, which enabled workers to take their physicals after a shower without 

getting dressed (AEC 1950k); if the breath radon samples were also taken there, that could explain 

the relatively high background radon. Mallinckrodt apparently corrected this by moving the breath 

radon sampling location to a lower-background area.

AEC considered that many of the early breath radon samples likely represented transient as well as 

fixed burden and that the background level at the point of sampling (which was generally ignored) was 

likely to have been relatively high; thus the resulting estimates they made of alpha radiation to the 

bone based on breath radon measurements would typically be higher than was actually the case 

(AEC 1950a). Up to about 1950, AEC assumed that 1 μCi/L of sample after at least two days of nonexposure represented a total radium burden in bone of approximately 0.2 μg of Ra; however, AEC 

had then decided to use an RBE of 20 for alpha to bone marrow and a skeletal weight of 7 kg as 

agreed on at the September 1949 Chalk River Conference, to give 1600 mrem/week to bone for each 

0.1 μg Ra deposited (AEC 1950a).

Whole body and lung counts appear to have been performed rarely if at all, since workers had to be 

sent to sites outside Missouri for this to be done or a mobile counter would have had to be brought to 

St. Louis. Hence there were evidently so few such counts done as to be of little use in reconstructing 
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individual doses, except possibly for those individuals actually counted. However, even individual 

whole body and lung count data appear to be unavailable

2.4 EXTERNAL DOSE CONSIDERATIONS

External doses for Mallinckrodt workers varied widely depending upon the activity they performed. 

Operations at the refinery (Plant 6) involved primarily gamma radiation, while operations at the metal 

plants (i.e., Plant 4 and later 6E) entailed primarily beta radiation (AEC 1949). 

There is little information about conditions in Plants 1 and 2 during the wartime startup: no dose rate 

measurements from 1942-1946 appear to have survived and as noted previously, film badging did not 

start until late 1945, when Plants 1 and 2 were in the process of shutting down. Doses might have 

been somewhat higher due to greater manual involvement and probably somewhat greater bodily 

proximity to sources, but on the other hand the quantities involved were much lower. It should be 

noted, for application to external exposure, the era of pitchblende use (early 1945 on) was mostly 

covered by film badge monitoring. Thus it is considered to be conservative to assume that the doses 

at Plants 1 and 2, for the same type of work, were not greatly different from those at Plant 4 and Plant 

6 around 1948.

According to MCW (1955), at least late in the life of the site, gamma surveys were done bimonthly in 

most Plant 6 processing areas and monthly at the vent ducts in the digest area. However, these 

reports do not appear to be available.

2.4.1

After high-grade pitchblende ores began to be used, refinery workers were exposed to high levels of 

energetic photons from radionuclides in equilibrium with U-238 and U-235. Ra-226, through its Pb214 and Bi-214 daughters, contributed energetic gammas to workplaces where ore was stored or 

processed. Upon removal of the uranium daughters, processed material became radiologically 

innocuous until the passage of time resulted in the ingrowth of Th-234 and Pa-234m and the 

consequent domination of the dose profile by electrons. Mallinckrodt worker dose records 

demonstrate this difference, with significant doses for mixed photons and electrons in the refinery 

operations and high electron doses with little photon dose in the metal plants. Dose reduction 

measures in plants and equipment resulted in low doses in Plants 6E and 7 compared with the mixed 

beta-gamma doses in the refinery operations.

Gamma, Beta (Electron), and Nonspecific Beta-Gamma Exposures

The gamma dose rate could be as high as 50 mR/hr near stacks of drums of Belgian Congo ore at 

25% concentration and with a radium content of about 100 mg/ton (Eisenbud 1975). Dose rates at 

points adjacent to stacks of drums of radium-bearing residues (precipitates) could run as high as 100 

mR/hr adjacent to stack of drums (~ 300 mg Ra/ton) (Eisenbud 1975, Table 2). In addition, a 1958 

AEC report on uranium mills gave dose rates of 0.8 to 8.0 mR/hr, with an average of 3.0 mR/hr, as the 

gamma dose rate at three feet from bulk ore concentrates (AEC 1958, Table XI); these dose rates are 

assumed to be for domestic ores. AEC (1948a) gave the gamma contact dose rate with the (Racontaining) Feinc filtrate sludge under equilibrium conditions as over 300 mR/hr; however, they stated, 

they had no way of knowing how close to equilibrium it was.

Some more specific information regarding gamma doses in the ore, refinery, and metal processing 

areas are shown in Tables 33-35. Dose rates from drums and railcars are shown in Table 33; from ore 

storage in Table 34, for Middlesex workers (comparable to ore storage areas at Mallinckrodt); and for 

various Plant 6 locations in Table 35, particularly for GLC (gangue lead cake or K-65, the radiumcontaining residue). It should be noted that operations that were particularly manual were the various 
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dumping, scooping, and scraping operations in which feed, UO2, UO3, UF4, and dust were handled or 

crucibles and furnaces were cleaned; the “plowing” (scraping) of the centrifuges; and the scraping of 

cake off the Feinc filter cloths (this was the pitchblende cake during the pitchblende years). Thus 

significant external dose reduction usually followed any mechanization of these processes.

Because the gamma dose arose mainly from the radium and its daughters, the gamma dose was 

significant only in those areas where the source material had not yet had the radium separated; where 

radium-bearing residues were present; or where uranium products were stored for long enough 

periods of time that the daughters built up again. This meant that the gamma doses tended to be 

highest in Plant 2 and later in Plant 6 (AEC 1949), especially around ore drums and storage areas for

the radium-bearing residue, K-65. Shielding had been designed into Plant 6 and more was added in 

1948 in some areas (AEC 1949). As noted in Section 5.2 above, there was also up to 2.6 mCi of 

radium built up in the residue that was processed in 1955-1957 to concentrate thorium, although this 

was distributed in the 350 tons that was processed into the 3600 gallons of solution sent to Mound 

(Tables 4 and 6).

Doses registered on film badges worn by people not working directly with the U and equipment, such 

as guards and office workers, was more likely from gamma exposure than from beta exposure. This 

is because they were usually at some distance away from the source (the uranium and its daughters). 

It is true that the dust was found throughout the plant to varying extents, but that would likely not 

contribute to the external dose rate much in or near buildings where there was a substantial Ra 

content in any uranium product or residue.

A 1958 AEC report on uranium mills gives 1.5 to 25 mrep/hr, with an average of 15.5 as the beta dose 

rate at three feet from bulk ore concentrates (AEC 1958, Table XI). AEC (1948a) gave the beta 

contact dose rate with the (Ra-containing) Feinc filtrate sludge (K-65) under equilibrium conditions as 

over 500 mrem/hr; however, they stated, they had no way of knowing how close to equilibrium it was. 

AEC estimated the dose to an operator’s hands from removing lids from ore drums at 200-300 

mR/day, even after a proposed body shielding window was erected (AEC 1948b).

Regarding experience at the Paducah site, Baker (1958) reported that the Th-234/Pa-234 

combination (from U-238 and U-234) produced about 1500 alpha dpm/mg U and 1500 beta dpm/mg 

U at equilibrium, producing 240 mrad/hr at the surface of U metal, 208 mrad/hr at the surface of UO3, 

and 183 mrem/hr at the surface of UF4. Further, during UO3 prep by "our suppliers" (e.g., 

Mallinckrodt), much of the beta-active material was removed, but built back up to 50-100% by the time 

it got to the UF6 production facilities (Baker 1958). This suggests that significant buildup could occur 

before the UO3 left the Mallinckrodt facilities since the storage time might be weeks and the transport 

time was likely less than a few days. Eisenbud (1975) points out that 90% of equilibrium beta activity

is restored by 90 days after vacuum casting. Eisenbud (1975) reports high dose rates, up to 1 

rad/week to the body and even more to the hands, from loading of UF4 into UF6 reaction vessels. 

This too implies that if enough time elapsed, UF4 loaded at Mallinckrodt into the bombs could also 

produce relatively high beta dose rates. Metallic uranium in equilibrium with Th-234/Pa-234 could 

produce up to 235 mrad/hr to the basal epithelium when the metal was in contact with bare skin; 

heavy gloves would significantly reduce this (Eisenbud 1975).

In addition to the beta dose rate from the uranium as natural uranium, uranium oxide, etc., there were 

two waste concentrates that produced high beta dose rates. First, when ether was used with the 

uranyl nitrate to extract the uranium, Th-234 and Pa-234 (also called UX1 and UX2 respectively) were 

left in the aqueous phase (also called the aqueous uranium tails) (Eisenbud 1975). This aqueous 

solution was filtered, resulting in a residue (cake) containing the beta emitters.
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Second, in the vacuum recasting of the uranium metal, impurities in the metal volatilized and 

condensed on the cooler portions of the furnace, creating spot deposits (AEC 1949; Eisenbud 1975). 

The impurities contained Th-234 and Pa-234, which were concentrated to a significant degree in the 

deposits (AEC 1949; Eisenbud 1975); this residue could have "up to 1000 times the beta activity of 

natural uranium" (AEC 1949). Manual contact with these deposits during charging, discharging, 

cleaning, and repair of the furnaces provided "opportunity for hand irradiation of a greater magnitude 

than whole body" (AEC 1949), possibly as much as 2-3 rads/week to exposed skin and perhaps to the 

eyes when the original ore was pitchblende at 25% average enrichment (Eisenbud 1975).

Regarding the processing of residues to concentrate thorium, Table 6 shows that with the interruption 

in the chain occasioned by the removal of the original radon (by venting) and the radium early in the 

process, the daughters had to build up again to equilibrium from the time the cake was stored through 

the maximum 15 years of storage. Consequently the strong beta emitters down the chain, such as 

Pb-214 and Ac-228 are present only in very small quantities.

Some dose rate information for exposure rates from laundry equipment and clothing appears in 

Tables 29 and 30 (this data is from the text and Table 1 of Utnage 1958b). This is mainly beta 

radiation. AEC did some clothing shielding and contamination studies using an 18” x 24” sheet of 

uranium metal in equilibrium with Th-234 and Pa-234 (AEC 1950h), with the following results. Denim 

coveralls (9-oz weight) “absorbed” an average of 22% of the beta from the source, with a standard 

deviation of 7.5%, for distances varying from 5 inches to 3 feet. Neoprene-covered cotton gloves 

shielding an average of 50%. Measurements on the inside surfaces of three cotton gloves used to 

handle uranium showed contact beta dose rates of 23-47 mrep/hr from contamination; these gloves 

had been taken at random from workers. The Mallinckrodt glove program for contact with radioactive 

material was said to be sketchy and inadequate (MCW 1955), implying that use of gloves was not 

consistent.

Measurement methods were not specified in most reports and papers. Counters and meters were 

maintained and calibrated weekly or monthly on a set schedule by Mallinckrodt technicians (MCW 

(1955-). Utnage (1958b) stated that for surface contamination measurements in the laundry (of 

clothing, equipment, and floor surfaces), a Victoreen 356 alpha survey meter and a Thyac betagamma meter with a thin-wall tube were used. There were no instrument monitors in any area, but 

film badges placed at selected locations in the process areas served as integrating area monitors 

(MCW 1955-2/21/55); there is no information as to how often they were collected and read.

2.4.2

No neutron exposure measurements are available but Dupree-Ellis et al. (2000) deemed neutron 

exposures at Mallinckrodt to be minimal. 

Neutrons

2.4.3

2.4.3.1 Film Badge Monitoring Periods

Information and Available Data Regarding Film Badges and Extremity Dosimeters

Workers were not individually monitored for external dose prior to 2 December 1945. From early 

December 1945 to March 25, 1946, at least some employees participated in a dose monitoring 

program for which total doses were reported in a memorandum from the University of Rochester 

School of Medicine and Dentistry (Rochester 1950). Subsequent to this 15-week program, film 

badges were issued to apparently all workers at Mallinckrodt's St. Louis uranium processing facilities. 

From 22 April 1946 through the end of MED/AEC work in 1957 or 1958, these were processed on a 

weekly basis as part of a routine dose monitoring program; however, MCW (1955) states that badges 
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were changed every two weeks or more often when indicated. Results were summarized quarterly 

and annually (MCW 1955). From the series of Mallinckrodt health group reports (an example of one 

is MCW 1951a), some 2000-3600 badges were read per month in the early 1950's.

2.4.3.2 Film Badge Technical and Processing Information

A letter documents the period of time when "Rochester" (i.e., the University of Rochester) processed 

the film badges (Rochester 1950). The letter states that Mallinckrodt processed the film badges 

during the period of 1 June 1948 to 1 January 1950. As is indicated by the Mallinckrodt health group 

monthly reports (e.g., MCW 1951a) and the 1955 Mallinckrodt health program description (MCW 

1955), Mallinckrodt continued to process their own badges. As the health group monthly reports and 

AEC (1950l) indicate, however, Mallinckrodt often had a significant backlog in reading the film badges.

No procedures and little other film badge specification data have been found to date. There is not 

much information about how the film badges were processed by either Rochester or Mallinckrodt. 

However, there was a series of meetings and correspondence between Mallinckrodt and AEC 

regarding whether certain readings were due to beta or soft gammas and whether the AEC and 

Mallinckrodt methods of correction for shield absorption (in the badge) were consistent (AEC 1950b; 

MCW 1950g; AEC 1950e; MCW 1950h; AEC 1950f; AEC 1950h). This issue involved the subtraction 

of the film density value under the beta “shield” from the value under the window. Section 7.3.1 of this 

technical basis document states the assumption to be made about this subtraction. However, these 

references may be consulted in case of any suspect beta readings corresponding to the 1949-1950 

time frame.

Film badges were issued as a combination security-exposure badge (numbered and with photo ID, 

per MED (1944b)) to all employees except for “office females” who presumably never entered process 

areas (MCW 1955). Since there were no potential sources of acute external exposure, the aim was to 

keep chronic exposures below tolerance levels (MCW 1955). All exposures over 50% tolerance were 

reported to supervisors (MCW 1955).

2.4.3.3 Film Badge Record Types, Arrangement, and Availability

Mallinckrodt dose records were of three types: complete records of weekly film badge results, listings 

of total doses by employee over a specified time period such as the "Mallinckrodt_1946" file (MCW 

Undated), and plant dose summaries. With the exception of records that show total dose by specified 

time period, records found to date show weekly badge processing cycles. Assignment of individual 

annual doses was based on deep-dose exposure (Dupree-Ellis et al. 2000). 

The complete records are weekly lists of employee names with beta and gamma doses. For the 

gamma doses some results are shown as "50*" and the asterisk refers to a footnote that reads 

"indicates less than" (MCW Undated). Values of 60 and 80 are sometimes asterisked in the beta 

column. Occasional values of "0" are found in the gamma column as well. Some records list 

employee names with total doses over specified time periods, with a start date and end date. 

Listings of total doses by employees over a specified time period other than a week are found in the 

dose reconstruction project file "Mallinckrodt_1946" (MCW Undated). The earliest results are of this 

form also and are recorded by total dose and number of weeks worked in the dose reconstruction 

project file "Mallinckrodt Radiation Summary APR 46 to MAR 48.” This document also supplies other 

important information for external dose reconstruction. 
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Dose summaries listed doses by plant, number and percentages of badges in dose ranges from 0-50 

mrep/week, 51-100 mrep/week, 101-150 mrep/week, etc. based on the total beta and gamma for an 

employee. Doses are not listed for employees having less than 150 mrep in a week; for the dose 

categories above 150 mrep/week, individual names are listed with gamma and beta dose results.

Complete records and/or dose summaries have not been located for all periods of MED/AEC 

operation as of this writing. This results in gaps for dose monitoring data when no information is 

available for workers in a given plant, or in some cases, for any Mallinckrodt worker. Dose 

summaries, when no complete records of weekly film badge results are also available, give no 

individual data for personnel receiving less than 150 mrep in a week. Many gaps in data are the 

result of accidents or damage in the workplace or during badge processing. These incidents are 

usually documented in the record.

2.4.3.4 Extremity Dosimeters

Because of the high extremity doses in cleaning the high-beta deposits out of the recasting furnaces, 

in 1949, film rings began to be used "by selected groups" in the metal plant (AEC 1949), but the 

records of these doses do not appear to be available.

2.4.4

Mallinckrodt uranium processing workers were given a pre-employment physical that included a chest 

x-ray (MCW 1955); Mason 1958a); they were also given an annual physical that included a chest xray (MCW 1955; Mason 1958a). That these x-rays were actually given is indicated in the series of 

Mallinckrodt Health Office and other reports (e.g., MCW 1951a; MCW 1954; MCW 1955). No 

information is available as to how much dose was received during these examinations or if all workers 

received annual x-rays. Therefore, to be claimant favorable, it is assumed that all personnel who 

worked in the plants covered by this technical basis received an annual diagnostic chest x-ray.

Information and Available Data Regarding Occupational X-Ray Examinations

2.4.5

No records appear to be available regarding skin contaminations. It seems likely that due to the 

relatively low radioactivity level of most of the uranium forms and the pervasiveness of the uraniumbearing dust, skin contaminations would have been regarded by safety officials at the time as not 

significant and thus would not have been recorded. See Section 5.3.5 regarding surface 

contamination, including clothing and smoker hand data.

Skin Contaminations and Other Radiological Incidents

No information has been found as to any incidents that may have resulted in significant 

overexposures to radiation or intakes of radioactivity. However, the following incidents were 

documented: a 1942 or 1943 explosion in the denitration process area in which agitators began to 

bind until the motors driving them finally tore loose from the concrete floor (Fleishman-Hilliard 1967); a 

1943 ether fire in Buildings 51 and 52 involving a dryer blown apart by an explosion of ether vapors 

due to burning ether (MED 1943a); a 3 July 1943 fire in a rubbish truck containing “lime” sweepings 

and other floor sweepings, possibly including metal slag (MED undated b); a 4 May 1946 explosion at 

Plant 6 involving an explosion due to seepage of ether into the nitric acid tanks due to a malfunction of 

a check valve; and a 1947 or 1948 explosion causing the rupture of a nitric acid holding tank due to 

mechanical failure of a check valve (MED 1948; Fleishman-Hilliard 1967). Also, there were indications 

of frequently occurring incidents, such as the occasional spontaneous catching on fire of uranium 

metal derbies at the derby cleaning station in Plant 6E, which would have put particulates into the air; 

this problem was spoken of as being brought under control (AEC 1952a). Another such recurring 
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incident was the plugging of floor drains, with the consequent formation of puddles of contaminated 

liquids on the floor at Plant 6 (AEC 1950k).

2.5 OTHER DATA OF DOSIMETRIC INTEREST

2.5.1

The initial April-July 1942 uranium pilot plant effort included 24 people working as a single project 

group under a project manager (Fleishman-Hilliard 1967). In 1944, there were 55 guards; 330 

workers (including guards) with a clearance for MED work, and 1500 workers on the entire site 

(presumably including non-MED workers) (MED 1944b). Regarding the total number of workers with 

potential for exposure, Fleishman-Hilliard (1967) and Mallinckrodt (1994) list the total number of 

workers as 250, the former stating that this was in 1948; AEC (1948b) lists the total number as 250 at 

Plant 6, but 400 if Plant 4 was included; AEC (1949) lists the number of workers at Plant 6 as 272 and 

the number at Plant 4 as 94. Mason (1958a) states that as of the beginning of 1948, more than 100 

of the original employees working during the period 1943-1946 were still on the payroll. AEC dust 

study reports in the 1950's give the number of each classification of workers and the number on each 

shift (e.g., AEC 1954b); some AEC reports even list the names of process and supervisory workers 

and their job classifications in an appendix. As noted above in Section 5.4.3.1, over 2000 film badges 

a month were processed in the 1950's.

Number of Workers

2.5.2

From AEC dust study reports (e.g., AEC 1954b), the following information regarding time spent is 

provided as follows:

Number of Hours Worked per Week

Length of work day, including 

breaks and locker room time 480-520 minutes (8-8.6 hours)

Lunch break 30 minutes

Smoking breaks 30-40 minutes

Clean locker room 20 minutes

Regulated locker room 15 minutes

The longer work day applied to operators and craftsmen, who presumably had to leave their work 

areas to smoke. There was a 10-15 minute variation in the work day among plants as well. The total 

smoking break time was 30 minutes for Plants 6E and 7, but 40 minutes for Plants 4 and 6 through 

about 1955; after that it was 30 minutes for all plants.

While AEC-NYOO took the weekly number of hours to be 48 (or six 8-hour days) in calculating some 

of their early time-weighted average airborne concentrations (AEC 1949), Lippmann (1958) used 40 

hours in reporting data regarding Harshaw workers. It can be assumed in the case of the Mallinckrodt 

workers that they typically worked for a full 8 hours a day, 5 days a week, or 40 hours per week, since 

that will conservatively cover both the actual 5-day and the actual 6-day cases. When using daily 

weighted average dust sampling data (e.g., in Tables 13-17 and 21-24), it is important to understand 

that break, lunch, and locker time was factored into the weighted averages reported by AEC and 

Mallinckrodt in their air dust studies.

Fleishman-Hilliard (1967) states that once the Plant 2 operations started (ether extraction), it was 

carried out 24 hours per day. It is not clear what other processes ran 24 hours per day. Guardhouses 

were manned around the clock, with three shifts per day (MED 1944b).
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2.5.3

After about 1950, film badge reports included a short note or keyword about the job or work done or 

the work area occupied by the individual during the week. After about 1948, many urinalysis sheets 

also list such notes or keywords. Various AEC reports also list job titles and functional work types. 

Those discovered to date are given in Table 19. Note that in the absence of further information, it is 

not possible at present to distinguish in these records and documents between ordinary or process 

decontamination and the decontamination that was part of decommissioning buildings and plants.

Job Types, Work Areas, and Work and Access Practices

As previously stated, to aid in classifying workers whose job titles do not appear in Table 19 and 

whose work descriptions do not make it clear which job title is appropriate for use, two other tables 

have been provided. The first is Table 5, the keyword table, which includes information from these 

notes and from operational information in other references (particularly MED 1946a; AEC 1949; and 

AEC 1978). The second is Table 20, the process-job association table.

These three tables should be used to help determine the principal occupational activity for an 

individual with missing or conflicting monitoring data.

Mallinckrodt employed an employee rotation program from about 1950 on (Fleishman-Hilliard 1967; 

AEC 1950b; MCW 1955), the point of which was to keep the weekly dose below the weekly tolerance 

level or, after about mid-1950, to keep the average weekly dose over a three-month period below the 

weekly tolerance level.

Mallinckrodt’s 1955 formal description of its health program (MCW 1955) gave the following 

information; it is not clear back to what year this information applies. Mallinckrodt maintained three 

levels of controlled areas. These were the regulated areas, which were the areas where radioactive 

materials were processed and handled; the grey areas, which were areas where any radioactive 

material and contamination was incidental to the function of the area, e.g., labs and production 

instruments departments; and clear areas, which were area where radioactive materials were not 

required and not permitted, e.g., offices and the cafeteria. Although zero contamination was not 

possible in the latter areas due to their proximity to the other areas, that was the goal of control 

efforts.

A 1944 MED security survey report (MED 1944b) gives the following information. In 1944 and 

presumably all other years as well, access into the MED areas was only through guarded entrances. 

Hourly rounds of the entire site were made during evening and early morning hours and all day 

Sundays and holidays; it took a guard 40 minutes to make a complete round of Plant 2. NonMallinckrodt truck drivers were allowed to come in, but not truck drivers’ helpers unless needed to 

unload; all trucks were escorted while within the site.

MCW (1955) also stated that work clothing “from the skin out” was provided for all persons assigned 

to regulated areas. Regulated clothing could not be worn outside regulated areas except under cover 

clothing; cover clothing was provided for brief visits to regulated areas. Because wearing 

contaminated regulated clothing on public land was undesirable, vehicular transportation was required 

for workers traveling between regulated areas (even with the cover clothing). Regulated area workers 

were issued two changes of clothing per day. Workers were required to shower before changing into 

“clear” (clean) clothing; they typically took two showers a day in 1955, but only one in the 1942-1944 

time frame (MED undated a). Both contaminated and clean clothing were laundered on the premises, 

in separate laundries.
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According to MED (1944b), smoking was not permitted in operational areas except in designated 

smoking areas or smoking rooms; smoking was permitted in offices and labs, except where ether or 

other flammable substances were handled. In 1944, the penalty for smoking in other than permitted 

areas was loss of employment. Thus it is likely that nearly all smoking was done in designated 

spaces.

2.5.4

Some quantities and dimensions of potential radiological interest (e.g., for special external dose 

calculations) are as follows. See also Table 4 and Section 4 for other amounts.

Miscellaneous Information

Bomb 10" OD × 40" long (MED 1946a)

Bomb liner (lime, etc.) Depth: 1" (MED 1946a)

Uranium billet 4.75" OD × 18" long (MED 1949b)

Ore barrel 3' high × 18" across (MED 1945a)

U metal samples

Glass tube 2" OD × 3⁄4" long (MED 1945a)

Cardboard packing box 5" sq, 1-2 lbs filled (MED 1945a)

U eggs (samples from billets) Packed eight to a box (MED 1945a)

Billet packing box, wood 5" × 5" × 13" (MED 1945a)
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3.0

Where urinalysis and other individual-specific bioassay data are available for a given period, it should 

be used to determine the individual's intake of radioactivity using the IMBA program (ACJ 2002) and 

the ICRP default parameters. 

DETERMINATION OF RADIOACTIVITY INTAKES AND INTERNAL DOSES

For analysis of urine samples, only uranium was counted, and in the air sampling, only gross alpha 

was counted. Thus it is not possible to know, e.g., how much radium was in the urine or air sample. 

Also, most workers will have some gaps in monitoring because routine bioassay did not begin until 

1948 and because there were undoubtedly some missed bioassays. The intakes over the gap 

periods will have to be determined either by comparable (surrogate) worker data or failing that, by the 

use of time-weighted daily average dust exposure data.

There was a great variability in exposure by job, by plant, and by year at the Mallinckrodt St. Louis 

site. Hence it is not feasible to calculate a matrix of intakes for all occupational types, all locations, 

and all periods for inclusion in this technical basis document. Rather, Table 31 should be used to 

determine the annual intake, prorated as necessary for the actual exposure period, for those 

individuals with missing or spotty bioassay data. Tables 21-24 should also be used, when 

appropriate, to help determine the time-weighted daily average exposure level to be applied on an 

individual basis to calculate the internal dose where urinalysis and related information is unavailable 

or spotty, especially for the period 1942-1947. 

Because urinalysis and film badge data is given in the records for the categories "AEC" and "Ledoux" 

or "Ledoux Laboratory", categories have been created for them as appropriate in Tables 21-24 and 

Table 31. However, AEC personnel presumably worked for AEC itself and the Ledoux personnel, as 

previously noted, were employed by the Ledoux Company and rendered services by contract to 

Mallinckrodt.

For workers employed in years during which no (radiological) urinalysis was done, reference should 

be made to AEC (1950a), which is a report of an attempt by AEC to estimate the cumulative 

exposures of Mallinckrodt workers then employed at Plants 4 and 6 who had been employed between 

July 1942 and October 1949 and who had more than six months of exposure to radioactive materials. 

Their estimates for the dose to the lung were based on air samples of alpha-emitting dusts (translated 

to a daily weighted average exposure level) and to the bone, on breath radon analysis (to determine 

the fixed radium burden and film badge data. Because in dose reconstruction different assumptions 

are made than AEC made and thus these calculations will have to be redone, their results are not 

repeated here. However, dose reconstructors should be aware of this report in the case of workers 

who began MED/AEC work before urinalyses were routinely done and for whom AEC’s “backcalculation” estimate may be found listed in dose records as simply an accumulated dose for the premonitoring period of operation. It should be noted that workers were not identified by name or work 

category in this report.

Also with regard to this report, it should be noted that the calculations of lung dose did not include a 

gamma contribution because AEC deemed it negligible compared to the dose from airborne particle 

inhalation; that they did not include radon because they could not estimate an average concentration 

and they assumed that airborne particle inhalation would dominate; that they assumed all the uranium 

to behave like UO2 in the lung; and that they assumed that biological equilibrium existed since the 

start of employment. AEC did include the external dose in the bone dose calculation. It should also 

be noted that AEC thought that the exposures in the unmonitored years were “at least as severe as 

they were found to be at the time of our initial studies” (in early 1947); conditions probably were not 
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more favorable and may have been “moderately” more severe. Thus they thought that their 

extrapolations could possibly be somewhat nonconservative.

Dose incurred during the decontamination and postoperations years (1959-1995) are covered in 

Section 8.0.

3.1 ASSUMPTIONS

The following assumptions should be made in estimations of radioactivity intakes by inhalation and 

the resulting internal doses when Tables 21-24 and/or Table 31 is used.

1. The number of work hours per year is 2000, i.e., 40 hours per week, 50 weeks per year. 

Adjustments can be made in individual cases if more specific information is available.

2. Urinalyses measured uranium only. For inclusion of other isotopes, the following may be 

assumed and is considered claimant-favorable.

By activity, natural uranium activity is about 48.9% U-238, 48.9% U-234, and 2.26% U-235. In 

the absence of other information, it would not be possible to tell what the degree of equilibrium 

is between the uranium isotopes and their daughters in a urine sample. It can be assumed 

conservatively be assumed that there is 100% equilibrium. Then for each pCi of uranium 

measured, about 0.49 pCi is U-234, 0.49 pCi is U-238, and 0.02 pCi is U-235. However, the 

U-238 portion can then conservatively be assumed to be U-234 as well, so there would be 

0.98 pCi of U-234 and 0.02 pCi of U-235.

Then under the assumption that the daughters of the original U-234 are in equilibtrium with the 

U-234 (but not the subsumed U-238), there are an additional 0.49 pCi of Th-230 and 0.49 pCi 

of Ra-226. Additionally; if the daughters of the U-235 assumed to in equilibrium with the U-235 

parent, then there is an additional 0.02 pCi of Ac-227. These conservative source terms are 

summarized in the table below.

For each pCi of uranium measured in urine:

0.98 pCi U-234

0.49 pCi Th-230

0.49 pCi Ra-226

0.023 pCi U-235

0.023 pCi Pa-231

3. Air samples measured gross alpha only. For inclusion of other isotopes, the following may be 

assumed and should be claimant-friendly.

For each 1,000 dpm of gross alpha measured in air:

392 dpm U-234

196 dpm Th-230

196 dpm Ra-226

7.3 dpm U-235

7.3 dpm Pa-231

4. When intakes from Table 31 or air concentrations from Tables 21-24 are used, they should be 

prorated over the listed period (e.g., a year) for the time spent in each corresponding 

occupational activity or area. For example, if bioassay data for a worker is missing for an entire 
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year and he spent 5 months of that year as an electrician and 7 months as a mechanic, his 

intake for the year should be assumed to be 5/12 of the annual intake for an electrician plus 

7/12 of the annual intake for a mechanic. The appropriate period from Table 32 should be 

used for this, depending on how many years of data are missing.

5. When air concentrations from Tables 21-24 are used, the prorating for the time spent in each 

corresponding occupational activity or area should be applied to the entire 2000 hour/year on 

the grounds of their being claimant-favorable and representatively time-averaged over all 

activities. That is, no time should be deducted for breaks, etc., since that was included in the 

time-averaging.

6. For Plants 1 and 2, the most conservative (usually the earliest) applicable data for Plants 4 

and 6 should be used, as appropriate for the worker activity.

7. The breathing rate for all workers should be taken to be 1.4 m3

/hr (as shown in Table 6 of 

ICRP 66 (ICRP 1994)). 

8. Radon exposures should be calculated from information taken from Section 5.3.4 and Tables 

25 and 27. 

9. Ingestion doses are negligible compared to inhalation doses, therefore ingestion can be 

ignored except in special cases.

3.2 ESTIMATING INTAKE BY USING SURROGATE WORKER OR ESTIMATED THORIUM 

INTAKE DATA (TABLES 31 AND 32)

Table 31, the surrogate worker annual intake table for uranium inhalation, and Table 32, the estimated 

intake table for the thorium processing operation, can be used to estimate missed dose or to 

generate doses for comparison to doses calculated from individual urinalysis and other data.

3.2.1

The steps in calculating intake by the use of the surrogate (comparable) worker data table, Table 31, 

are as follows.

Using the Surrogate Worker Intake Data (Table 31)

The assignment of job title and if needed, work area should be made on the basis of the claimant's 

submitted information, urinalysis records, film badge records (if helpful), employment records and 

other information. Where the job title or work area is not clear, Tables 5, 20, and 21 should be used 

to help make the selection. The job titles and work areas should then be tabulated by plant and by 

time period.

A surrogate worker value should be selected from the table(s) according to the appropriate worker 

classification. If there is no specific worker type or area given in employment or urinalysis records or 

other sources, then the "generic" or "mixed/miscellaneous" value should be selected, subject to the 

judgment of the dose reconstructor. For example, if worker is listed in urinalysis records as working in 

the "Bomb Step" of 6E in 4 out of 5 urinalyses records and only "6E" in the fifth record, it can be 

assumed that for gaps in this records series that he worked in the bomb step operation during the 

period covered by the fifth record; thus the "6E Bomb Step" data can be used. However, if the 5 

records include "Bomb Step", "Recast", and "6E", then the "6E Generic" data should be used.
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The appropriate datum should be selected from the table(s) for each period, as appropriate. If records 

indicate that the worker spent years in one position instead of skipping around, then the longer period 

data can be used. This will be representative and since it covers more years, can save calculation 

time. Where there are occupational changes, the shorter period data should be used. This is 

especially true for the early period (Period 1), which, as Table 31 shows, usually involved higher 

exposures for the most exposed workers than the later periods (Periods 2 and 3).

Data for the unmonitored years from 1942-1947 should be taken from Period 1 for Plants 6 and 4, 

where possible. However, for these years a calculation should also be done using the data from the 

time-weighted daily average exposure tables, as given in Section 6.3. below, for comparison, to 

ensure that the intake as inferred from Table 31 does not seriously underestimate the potential intake.

3.2.2

The estimated thorium intake data table, Table 32, is applicable only to those workers who were 

involved in processing thorium (ionium) in Plant 7E during the years 1955-1957. These would be only 

those workers who have some indication in their records (e.g., urinalysis records) that they worked in 

Plant 7E, the Minor Elements Production Facility, MEP, the ionium processing plant, or similar.

Using the Estimated Thorium Intake Data (Table 32)

The appropriate data should be selected from the table for the appropriate periods worked and 

prorated if partial years were involved. Note that the intakes for 1955 and 1957 have already been 

prorated for the partial years in the table.

3.3 ESTIMATING INTAKE BY USING TIME-WEIGHTED DAILY AVERAGE EXPOSURE 

DATA (TABLES 21-24)

The steps in calculating intakes by the use of the time-weighted daily average exposure tables, 

Tables 21-24, are as follows.

The job title selection from Table 19 and the work category selection(s) from Table 20 should be 

made on the basis of the claimant's submitted information, urinalysis records, film badge records (if 

helpful), employment records and other information. Table 5 should be used to help make the 

selection. The category(-ies) should be tabulated by plant and by time period. .

The air concentrations from Tables 21-24 should be selected to correspond to the work category(-ies), 

plants, and time periods.

Other assumptions should be made as given in Section 6.1 above. Any necessary adjustments 

should be made to allow for partial years, overtime, etc.

The intakes, in pCi, should be calculated by multiplying the appropriate air concentrations by the 

breathing rate(s) and the hours, and dividing by 2.2 dpm/pCi.

For a worker who spent time in an area where radon might present a hazard, the intake, in pCi, 

should be calculated as in Step 5, except that adjustment should be made for time spent in the area 

(i.e., the time may typically have been less than a full work day, depending on the job) and the dpm to 

pCi conversion need not be made although the μCi of the input should be converted to pCi by 

multiplying by 106

. 
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3.4 Calculation of Internal Doses for Missing Periods or for Comparison

Once the intakes in pCi have been determined for each year or subyear period as explained in 

Sections VI.A-C above, the intakes can be used in the IMBA-NIOSH program (ACJ 2002) to calculate 

the annual doses for the organ of interest. Reference should also be made to the NIOSH guide on 

internal dose reconstruction (NIOSH 2002b) and to the dose reconstruction project internal dose 

procedure (ORAUT 2003 b).

In structuring the IMBA case(s), it should be considered that the annual intakes can be assumed to be 

chronic and thus, where the intake is much the same for several years running, they can be summed 

and amortized over those several years without loss of accuracy or conservatism, for the following 

reasons.

1. The actual intakes were in fact mostly chronic.

2. Because of the typically long lapse of time between exposure and onset of cancer for most 

Mallinckrodt claimants, whether the intake corresponding to a given year occurred at the 

beginning or the end of a year or evenly over the year makes little difference.

In selecting the IMBA input parameters, the following should also be considered. As noted earlier, the 

uranium forms processed at Mallinckrodt all appear to be of the insoluble form. Human and animal 

studies have indicated that oxides of uranium can be very insoluble (see ICRP Publication 71, pg. 

299 (ICRP 1995a)). As noted earlier, ICRP 68 (ICRP 1995b) recommends Type M for UF4. 

However, to be claimant favorable, the selection of absorption type should depend on the organ of 

interest. Thus in general, Absorption Type S should be selected for respiratory tract dose calculations 

and Absorption Type M for all other organ dose calculations. ICRP 66 default parameters should be 

selected for particle deposition parameters in the IMBA program, unless other information (such as 

that presented in Table 8) for some overriding reason prompts a different choice.

The annual organ doses produced by the IMBA program (ACJ 2002) can then be entered into the 

NIOSH IREP program (NIOSH 2002c).
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4.0

4.1 GENERAL CONSIDERATIONS FOR MALLINCKRODT EXTERNAL DOSE 

RECONSTRUCTION 

DETERMINATION OF EXTERNAL DOSES

To date, only incomplete dosimetry monitoring records are available for Mallinckrodt employees. As 

alternatives for dose estimation information, area monitoring data for external radiation is sparse, and 

though much is known about the process material (source term), considerations of amounts of 

material and geometry characteristics render any dose estimation with this data subject to a great 

deal of uncertainty. Thus the approach incorporating the least uncertainty is likely to be based upon 

existing dose monitoring records. 

An external dose reconstruction guideline was formulated for Mallinckrodt claims. Generally, for a 

given claim, dates of employment are compared to the available dose monitoring information. If dose 

monitoring records are likely to be available, the dose reconstructor will request project personnel to 

search the available records and list recorded doses in the external dose reconstruction spreadsheet 

for the case. Then reconstructed dose is assigned for each weekly cycle worked during the early, 

unmonitored period. Dose reconstructions are then performed using the Mallinckrodt external dose 

reconstruction guideline in accordance with the dose reconstruction project external dose procedure 

(ORAUT 2003a) and with the NIOSH external dose reconstruction guidance (NIOSH 2002a).

Co-worker data, in Mallinckrodt cases, is likely to be a matter of identifying an appropriate worker with 

a similar work history. Use of "surrogate dose histories" is detailed in the external dose reconstruction 

guideline for Mallinckrodt, as is the performance of "summary dose evaluations" using only the 

complete records and neglecting data gaps. Summary dose evaluations are performed for cases 

where probability of causation greater than 50% may result from this quick estimate. Probability of 

causation above the criterion allows the dose reconstructor to neglect the underestimating omissions 

of missing dose data. 

Dose incurred during the decontamination and postoperations years (1959-1995) are covered in 

Section 8.0.

4.2 UNMONITORED WORKERS

Mallinckrodt dose records demonstrate that a substantial proportion of employees had film results 

recorded as 0-50. From this we infer that the dose-monitoring program was sufficiently conservative 

that even many individuals who did not receive significant occupational exposure were also 

monitored. It is further assumed that the converse is true; that individuals who were not monitored 

were unlikely to receive significant occupational exposure. To account for incidental exposure missed 

by the monitoring program, these individuals are assigned incidental dose in accordance with Section 

7.4.2.4 below. 

4.3 APPLICATION OF DOSE DATA

This section defines the use of Mallinckrodt dose monitoring data in performing dose reconstructions. 

4.3.1

One reference stated that the same film badge was in use throughout the MCW uranium division 

operation (MCW 1961). In an unpublished report this was described as follows:

Incorporation of Available Film Badge Dose Monitoring Data
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The A.M. Samples stainless steel badge holder with open-window and cadmium filters 

permitting beta and gamma differentiation and measurement. DuPont Type 552 

dosimeter film was used in the badge. The film was processed by techniques 

calibrated and standardized with film exposed to standard gamma and beta radiation 

sources. Gamma standards were obtained by exposing film to a platinum 

encapsulated radium needle. Beta standards were obtained using an aged, natural 

uranium block as a source. 

The foregoing comment does not describe calibration using a phantom, and it is likely that openair calibrations were performed. Therefore the recommendation is that Mallinckrodt recorded 

doses be converted using dose conversion factors for Roentgen-to-Hp(10) dose for photons 

from Appendix B of the NIOSH External Dose Reconstruction Implementation Guideline (NIOSH 

2002a). 

Work underway for this project at the present time describes the standardization of the film badges 

used by AEC contractors. A similar badge in use at Idaho National Engineering Laboratory (presently 

INEEL) is listed as having an open-window density of 0 mg cm-2 and a density behind a 1-mm Cd 

shield of 1000 mg cm-2. This work, to be published as a complex-wide technical basis for external 

dosimetry, states that MED/AEC/DOE beta/photon dosimeters are generally expected to reasonably 

measure the Hp(10) dose under most workplace radiation fields, and is claimant-favorable with the 

single exception of response in workplaces with low-energy photons, such as plutonium facilities. For 

this reason, no modification is proposed to recorded deep doses, once converted to organ doses 

using the Roentgen-to-Hp(10) dose conversion factor.

Examination of the results for ‘gamma’ and ‘beta’ in the Mallinckrodt dose records gives the 

impression that the beta doses are derived by subtracting the optical density for the shielded portion 

of the film from that of the unshielded part. This is borne out in a series of memoranda between the 

AEC and Mallinckrodt (AEC 1950e, AEC 1950g, MCW 1950g), where the method of subtracting the 

two quantities is discussed. For Mallinckrodt, it is assumed that ‘the beta readings are obtained by 

subtracting the density under the shield from the density under the window and assuming the 

difference in density is due to beta radiation’ (MCW 1950g). For the purposes of dose reconstruction, 

the ‘beta’ readings in the Mallinckrodt dose records are assumed to be equal to the shallow dose, 

Hp(0.07). A claimant-favorable dose conversion factor of one is assumed for application of shallow 

dose to the skin, testes, and breast.

4.3.2

Geometry of external radiation exposures varied significantly by job assignment. When job 

assignments are known in sufficient detail from the individual work history, exact geometries may be 

applied by specific job title from Table 19. The geometry factors in this table are developed from the 

time-and-motion studies performed as part of the AEC dust exposure monitoring activities. For dose 

reconstruction, these may be applied using the "custom geometry" option in the external dose 

reconstruction spreadsheet calculational tool. 

External Exposure Geometries

When such detailed individual information is not present in the employee information, assumptions 

must be made based on the information available. More general categories are available to apply to 

these cases. These incorporate mixtures of the anterior-to-posterior (AP) and rotational (ROT) 

geometries. 

Exposure geometries for three job categories in uranium facilities are listed in the external dose 

reconstruction implementation guideline (NIOSH 2002a): "general laborer," "machinist," and 
Effective Date: 10/24/2003 Revision No. 00 Document No. ORAUT-TKBS-0005 Page 57 of 124

"supervisor". Though these categories are useful when little is known about a given uranium facility, 

the details of the Mallinckrodt facilities are better understood. Based on the research to date, the 

following general categories are suggested for use when detail is insufficient to allow use of Table 19. 

Corresponds to the "supervisory" category in the IG. Should be used for supervisors, 

foremen, and mechanics assigned to multiple buildings.

50% AP/50% ROT

Corresponds to the "general laborer" category in the IG. Should be used for personnel who 

are not assigned to specific process equipment.

25% AP/75% ROT

Corresponds to the "machinist" category in the IG. Is applied to most process workers.

75% AP/25% ROT

Is used for Mallinckrodt workers who performed process work and who likely received a 

majority of their dose in the frontal geometry in a short period of time.

90% AP/10% ROT

Is used to assign dose to non-process workers.

100% ROT

The most significant adaptation of the categories above from the categories and proportions in Table 

4.2 of the IG is that the ROT geometry has been substituted for the ISO geometry. There are several 

reasons for this change. A survey completed subsequent to the operation of the Destrehan Street 

plants (MCW 1959) states in the summary that “In general most of the radioactive contamination is 

located in floors, [and] walls below the six-foot level...Walls, steelwork, and metal platforming...above 

the six-foot level are usually only slightly contaminated”. It is true that a survey completed after 

operations ceased would not perfectly reflect operating conditions, especially since the roofs of 

processing buildings were found to show heavy contamination (MCW 1959). However, review of the 

process and available floor plans conducted for preparation of this document show little overhead 

process piping such as would be found in a more recently designed facility such as a commercial 

nuclear power plant. Further, the early process shows much direct manual handling of the radioactive 

material. Finally there is the consideration that for the workers receiving the highest doses, higherdose activities often imparted much of the dose in a short period of time while directly handling the 

material or equipment. Isometric-geometry exposures are included in the detailed geometries in 

Table 14 for activities likely to result in overhead dose. Error in these assumptions occurs on the side 

of claimant-favorability by adoption of the ROT as opposed to the ISO geometry in most cases. 

4.3.2.1 Photons in the 50% AP/50% ROT Category

Dose conversion factors in this category represent an even split between the AP and the ROT 

exposure geometries. Job titles to which this geometry should be applied are supervisors, foremen, 

and mechanics/maintenance personnel without specific equipment or area assignments.

4.3.2.2 Photons in the 25% AP/75% ROT Category

Dose conversion factors in the tables represent a 25% AP exposure and a 75% ROT exposure. 

Examples of personnel to whom this general geometry should be applied are warehouse workers, 

general laborers (personnel not assigned to specific process equipment) in process plants, engineers, 

fork truck drivers, and instrument technicians.
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4.3.2.3 Photons in the 75% AP/75% ROT Category

Exposure geometries for this category of workers reflects a 75% AP/25% ROT geometry. This 

geometry is appropriate for many process workers. 

4.3.2.4 Photons in the 90% AP/10% ROT Category (Process Workers)

Time-and-motion studies document that certain workers received most of their external exposure in a 

short period of time performing dose-intensive activities. This geometry should be applied to process 

workers in the higher-dose categories. 

4.3.2.5 Photons: Non-Process Workers

These workers are expected to have entered process areas only incidentally and rarely approached 

process equipment as part of their assigned duties. Examples of these types of workers are clerks, 

other office workers, and dispensary personnel. 

4.3.3

Metal plant workers are exposed to natural uranium separated from radium and its progeny, the 

source of high-energy photons in uranium ore. Workers in the refinery were exposed to uranium in 

many states, from minimally processed ore through the various stages of uranium separation. Other 

workers, such as laboratory workers and guards, were exposed to uranium in varying states also.

Photon Energy Ranges

Photon doses for all workers should be assumed to be evenly divided between the 30 – 250 keV and 

> 250 keV energy ranges. 

4.4 RECONSTRUCTED EXTERNAL DOSE 

Some considerations for the reconstruction of external dose in this technical basis document are 

based on the methodology discussed in Watson et al. (1994). That study utilized external doses for 

workers at the Oak Ridge National Laboratory (ORNL) and the Y-12 Facility in Oak Ridge, Tennessee 

to evaluate the accuracy of estimates resulting from the use of the NEARBY procedure, developed by 

D.J. Strom, as cited in the reference. The procedure is not treated in detail here, due to the ready 

availability of the reference. The results are significant for this document, however. A statistical test 

for goodness of fit between estimated doses and actual doses showed the first step (of the 10 ordered 

steps) of the NEARBY procedure to result in the smallest difference between estimated and actual 

doses for ORNL. The correlation between estimated and actual dose went down, in general, with 

increasing step, with exceptions. For the uranium facility, Y-12, the use of department median or 

mean doses produced as good a fit as use of the NEARBY procedure. Though exact application of 

the NEARBY procedure is not possible in this case, the methodology is followed to the extent possible 

with existing dose monitoring data.

4.4.1

The lack of early external monitoring data for Mallinckrodt likely reflects the novelty of the uranium 

processing industry, the provisional nature of early uranium activities at Mallinckrodt and the 

assumption that airborne exposure was the primary hazard. The implementation of a more 

comprehensive health and safety program in the early post-war period led to questions about external 

doses that previously had gone unmeasured. This resulted in the publication of the AEC report “An 

Estimating Dose, Unmonitored Period 1942-1945: Workers With Subsequent Dose 

Monitoring Records
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Estimate of Cumulative Multiple Exposures to Radioactive Materials, Mallinckrodt Chemical Works 

Plants 4 and 6, July 1942 to October 1949” (AEC 1950a). This study develops cumulative dose data, 

but unfortunately does not apply it to individual workers. Doses applied to individual workers for dose 

reconstructions should rely upon recorded doses for actual workers during the monitored period to 

project doses for the unmonitored period. However, dose reconstructors should be aware of this 

report in the case of workers who began MED/AEC work before film badges were routinely worn and 

for whom AEC’s “back-calculation” estimate may be found listed in dose records as simply an 

accumulated dose for the pre-monitoring period of operation.

Because in dose reconstruction different assumptions are made than AEC made and thus these 

calculations will have to be redone, their results are not repeated here. The cumulative exposure 

estimates covered workers then employed at Plants 4 and 6 who had been employed in MED/AEC 

work at Mallinckrodt between July 1942 and October 1949 and who had more than six months of 

exposure to radioactive materials. AEC’s estimates for the dose to the skin were based on film badge 

data and for the dose to the bone on breath radon analysis (to determine the fixed radium burden) 

and film badge data. Calculations of lung dose did not include a gamma contribution because AEC 

deemed it negligible compared to the dose from airborne particle inhalation. It should also be noted 

that AEC thought that the exposures in the unmonitored years were “at least as severe as they were 

found to be at the time of our initial studies” (in early 1947); conditions probably were not more 

favorable and may have been “moderately” more severe. Thus they thought that their extrapolations 

could possibly be somewhat nonconservative.

For workers whose covered employment took place during 1942-1945 and have dose results from the 

early monitored period, external dose may be estimated from the total dose listed in the period in the 

tabulated total doses in the Mallinckrodt Radiation Summary (MCW undated). In this document, only 

total doses for all weeks worked are listed. For these workers, the average weekly dose is computed 

by dividing the listed total for gamma and beta each by the number of weeks worked. This average 

weekly gamma and beta dose is then applied to each week worked during the unmonitored period. 

This conforms as closely as possible to Step 8 of the NEARBY procedure, but the lack of data 

prevents the exact following of NEARBY. Additional sources of uncertainty for this estimate are that 

the correlation of estimated-to-actual doses decreases with the number of the step; the statistics of 

applying Step 8 to several years prior to the year for which dose is copied is untested in the reference; 

and finally, the limitations pointed out at the beginning of this section add to uncertainty also. 

However, this method seems less inherently uncertain than the application of a simple median from 

the MCW (undated) data.

This approach appears to be generally consistent with the approach of Dupree-Ellis et al. (2000), who 

stated that in an Oak Ridge Associated Universities study of Mallinckrodt workers, "for the 20.8% of 

working years in which doses were not monitored [i.e., 1942-1944], an algorithm was used to assign 

doses". This author also states that use of 1946 data ‘should provide valid exposure estimates for this 

early period’ (Dupree-Ellis et al. 1998). In situations where data for 1946 is incomplete or unavailable, 

the dose reconstructor should move on to subsequent periods for monitoring data which could 

introduce additional uncertainty into the dose estimate. 

Based on this method, for workers with recorded doses for 1946 (or the nearest time period with 

recorded doses), the dose reconstructor should assign the same doses for years 1942-1946 when the 

record indicates the same or similar work assignments. When the record indicates that the work 

assignment changed, co-worker data should be found that corresponds with the likely work 

assignment and this dose assigned for the appropriate period.
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4.4.2

It is assumed that workers without subsequent dose monitoring records either terminated prior to the 

beginning of dose monitoring or performed work that did not meet later Mallinckrodt criteria for 

monitoring. For the former, a surrogate dose history is to be formulated based on workers with similar 

job titles during the period of external radiation dose monitoring. The latter may have had low- or nodose jobs during the 1942-1945 unmonitored period, in which case assignment of a lower dose based 

on the median dose from the early monitored period is a claimant-favorable measure. 

Estimating Dose, Unmonitored Period 1942-1945: Workers Without Subsequent 

Dose Monitoring Records

1. Workers assumed to have terminated prior to the start of external radiation monitoring

Doses to these individuals should be assigned based on surrogate dose history estimated 

from recorded doses of co-workers from 1946, or the closest subsequent period. Dose 

reconstructors must compare information available from the DOE record and the computeraided telephone interview (CATI) to the reference documents in order to identify workers with 

a similar work history, whose recorded doses should then be applied to the worker for whom 

no doses were found. Care must be exercised to identify the appropriate work history to use 

as a surrogate for the subject employee, but adequate information is likely in the documents to 

formulate a reasonably accurate surrogate dose history when the case file contains adequate 

work history information for the employee.

. 

2. Workers who were outside the uranium division operation during the monitored period

For these individuals with no identified records, and work assignments subsequent to the 

unmonitored period that would probably not result in significant exposure, application of the 

average doses from the early monitored period provides a claimant-favorable estimate that 

likely addresses any incidental dose that may have been received from 1942-1945. These are 

presented in Table 36. 

. 

The values in Table 36 were generated from the average doses received by Mallinckrodt "pilot plant" 

workers during the earliest known period of film badge monitoring, the 15-week program described 

above (Rochester 1950). It incorporates significant uncertainty due to the following features of the 

data. Firstly, only total doses were listed for each worker for the number of weeks monitored (n = 1-

15). Second, no detail is supplied in the reference as to what activities the workers were engaged in, 

other than the fact that the location was listed as "pilot plant." The memo, however, states that the 

doses are from "prior to the operation of the current plant", and is dated around the time that Plant 6 

became operational; also, it is known that Plant 4 was called "the pilot plant" after Plant 6 was built 

and before a pilot plant was established at Plant 6. Thus, though it is unknown whether the listed 

doses were received at Plant 1, Plant 2, Plant 4, or a combination of plants, the results likely reflect 

doses received in the early operations prior to the improved control measures presumably 

implemented in the construction of Plant 6. Further, to what extent early "benchtop" operations may 

have resulted in doses that differ from the estimates below is not known. Finally, as clear production 

levels have not come to light for the early period, no attempt has been made to scale the exposures to 

reflect the quantity of material processed. These factors result in dose estimates that are highly 

uncertain, but represent the best information at hand.

The methodology used to create Table 36 was compared with the techniques discussed in Watson et 

al. (1994). Whether it would be a better fit to use the department median or mean, due to the fact that 

Mallinckrodt was a uranium facility like Y-12, or to use the NEARBY procedure is unfortunately 

academic in the case of workers without monitoring records subsequent to the unmonitored period, as 
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insufficient data is available even to approximate the use of NEARBY. Nor is it known how Plants 1 

and 2 may have been like Y-12. Additionally, the Mallinckrodt facilities in use from 1942-1945 likely 

changed significantly over this period, so the dose estimates are applied to plants and equipment with 

unknown stability in function and arrangement, unlike those used in the study. However, the 

approach described below resembles the use of a department mean or median dose as closely as 

possi

From total doses in Rochester (1950), the dose distribution of the average weekly dose for the 32 

workers considered was evaluated and the values in Table 36 were prepared. Distribution of the data 

and values for the median and geometric standard deviation were calculated using LOGNORMTM and 

CrystalBall©. 

ble with the available data.

The median dose is applied to each cycle for which dose is reconstructed during the unmonitored 

period. For deep dose, the conversion to organ dose for the relevant worker category is 

accomplished by the use of the energy proportions specified in Section 7.3.3, the geometry 

proportions listed in Table 19 or Section 7.3.2 (as appropriate), and the energy- and geometry-specific 

dose conversion factors given in the NIOSH External Dose Reconstruction Implementation Guide 

(NIOSH 2002a). Shallow dose is assigned as electrons of energy greater than 15 keV (with an 

implicit dose conversion factor of unity). Integrating the triangular distribution of the energy 

uncertainty and the lognormal distribution in the table is best done with an appropriate computer 

application such as CrystalBall©. 

4.4.2.1 Dose During the 15-Week Monitoring Period

Dose from this period for listed workers should be estimated from the average dose per week 

calculated from the listed total dose and the number of weeks the employee participated in the study. 

Other claimants (not listed in the dose results) are assigned doses from Table 36 in the same manner 

described above.

4.4.2.2 Gaps in Data

Periods when monitoring records are missing should be filled in accordance with the guidance of 

NIOSH (NIOSH 2002a) and with the dose reconstruction project external dose reconstruction 

procedure (ORAUT 2003a). This may involve extrapolation, interpolation, or both. 

4.4.2.3 Non-Specific Summary Results

Monitored workers may have received up to 150 mrep in a week without the dose being recorded, if 

dose summaries are the source of dose information. There is some probability that the dose received 

was actually zero: dose summaries consistently show a significant number of badges in the 0-50 mrep 

dose range. For the purpose of dose reconstruction, the most claimant-favorable assumption is that 

any monitored employee assigned to a given plant but not specifically listed in the dose records for 

the plant for that week received a dose of 150 mrep that week. Division of the total between gamma 

and beta components is based upon an average ratio derived from the weeks with specific dose 

monitoring entries for that employee. 

4.4.2.4 Unmonitored Workers

As stated above, most workers who received significant occupational exposure were likely to have 

been monitored. To account for the possibility of an unmonitored individual receiving incidental 

exposure to photons, dose will be assigned for unmonitored Mallinckrodt employees. 
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For the purposes of dose reconstruction, the minimum level of detection was determined to be 50 

mrep/week from the fact that Mallinckrodt individual dose monitoring records list many entries in the 

gamma column as "50*"and the asterisk refers to a footnote that reads "indicates less than." 

Individuals without monitoring data are to be assigned "incidental dose" for each weekly cycle worked. 

This is to be assigned as photon dose equal to the LOD (limit of detection) divided by 2 (LOD/2), or 25 

mrep, for each weekly cycle worked. This 25 mrep is assumed to be in the form of photons of energy 

between 30 and 250 keV, and the exposure geometry is assumed to be 100% ROT. 

Use of the LOD/2 method results in "slight positive bias" for monitored workers, as stated in the 

NIOSH external dose reconstruction guidance (NIOSH 2002a). Additional claimant-favorability arises 

from the fact that as unmonitored workers they are less likely to receive significant occupational dose 

(see above), and so true doses are closer to zero than "less-than-detectable" or "missed doses" for 

monitored workers. This overestimating assumption is a claimant-favorable way to account for 

individuals who may have been incidentally exposed. 

4.5 X-RAY DOSE

4.5.1

Employees of Mallinckrodt (cf. Section 5.45) received an annual occupationally related diagnostic xray (MCW 1955; Mason 1958a). The annual chest x-rays were taken at the Barnes Hospital (the 

Washington University School of Medicine) (AEC 1950k; MCW 1955) and the x-ray records remained 

the property of the hospital (AEC 1950k). There is no evidence so far in the Mallinckrodt documents 

to indicate when the annual chest x-rays began, although Fleishman-Hilliard suggests that this was 

done from the start since Mallinckrodt insisted at the outset that the Washington University School of 

Medicine be engaged to do the physical examinations. A claimant-favorable assumption would be that 

chest x-rays were performed annually from 1942-1958. Since the radiographs were made at a 

hospital, and hospital procedures routinely used both PA and lateral chest views much more 

commonly than non-hospital facilities, the annual dose from chest x-rays for Mallinckrodt workers 

should consist of the dose from both views. No evidence so far indicates that photofluorographic chest 

films were performed.

Mallinckrodt-Specific Information

Since no actual x-ray output measurements or x-ray technique factors are available for the Barnes 

Hospital in Mallinckrodt records, default values for entrance kerma appropriate for this time period will 

be used in the calculation of organ dose conversion factors for use in dose reconstruction.

4.5.2

Information to be used in dose reconstruction for the early years for which no specific information is 

available is provided in the Savannah River Site technical basis document for dose reconstructions

(ORAUT 2003c). This information was used to produce Table 37, the organ doses for a generic 

occupational diagnostic x-ray examination. The important points supporting this table are reiterated 

below for the convenience of the dose reconstructor, but ORAUT (2003c) should be referred to for 

complete information.

General Information

A source-to-image distance (SID) of 72 inches (183 cm) was standard for the time for both the PA and 

lateral views. The x-ray machines from this time period (1942-1958) were single phase and typically 

no air gap was used between the patient and the film. It is assumed that the x-ray equipment was 

operated at 80 kVp, had at least 1.5 mm Al total filtration (see Table 3.1 of NCRP 102 (NCRP 1989) 

and that the HVL was approximately 2.5 mm Al equivalent (see Table B.2 of NCRP 102), which are 

typical machine parameters for chest x-rays performed in this time period.
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Specific organ doses for the PA and lateral chest x-rays calculated on the basis of the dose 

conversion factors found in ICRP Publication 34 (ICRP 1982) are given in Table 37. Doses for organs 

not listed in ICRP 34 but specified in the IREP code were determined by analogy with anatomical 

location as indicated below. 

Analogues for IREP organs not included in ICRP 34

Anatomical 

location

ICRP 34 

reference organ

IREP organ 

analogues

Thorax Lung Thymus

Esophagus

Stomach

Bone surface

Remainder organs

Abdomen Ovaries Liver/gall bladder

Urinary/bladder

Colon/rectum

Uterus

Head and neck Thyroid Eye/brain

It is assumed that the x-ray beam was not collimated, even though these x-rays were performed at a 

hospital and collimation practices were generally better in hospitals than in non-hospital facilities. 

Therefore, organs not normally in the primary beam for a PA and lateral chest were included in the 

primary beam by using ICRP 34 organ dose conversion factors for procedures where those organs 

would normally be included in the primary beam.

For any individual entrance skin exposure (ESE) or derived organ dose, an uncertainty of ± 30 % at 

the 99 % confidence level may be assumed; for further conservatism it may be appropriate to assume 

that errors are all positive and thus that only the + 30% should be used.

4.6 OTHER DOSE CONTRIBUTIONS 

4.6.1

Given the relatively small number of claims and the unknown proportion requiring calculation of 

extremity dose, this subject is not treated in this TBD. Extremity dose estimates, when necessary, will 

be formulated on a case-by-case basis. 

Extremity Dose

4.6.2

Submersion dose is likely to be significant only for the skin, testes, and breast, and is not used when 

testes dose is used to estimate dose to the prostate. As dose reconstructions are based upon the 

partial film badge dose monitoring records, submersion doses are not separately calculated. 

Submersion Dose

4.6.3

As discussed above, Mallinckrodt dose records contain ‘beta’ values obtained by subtracting the 

optical density of the film behind the Cd shield from that behind the open window. These recorded 

values are assumed to be equivalent to Hp(0.07). Work presently underway may result in default 

assumptions regarding appropriate electron organ dose conversion factors to apply in uranium 

facilities, but at the present time, for the purposes of dose reconstruction, shallow dose is assigned 

from the film badge data using a claimant-favorable dose conversion factor of 1. 

Shallow Dose
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As noted earlier, work conducted in the early days of the AEC Health and Safety Division dealt with 

attenuation of beta sources by workplace clothing (AEC 1950h). Attenuation of beta radiation was 

determined for coverall cloth to be an average of 21.8% (with a standard deviation of 7.5%) for 

distances between 6 and 36 inches from a sheet of uranium metal in equilibrium with its two major 

beta-emitting progeny, Th-234 and Pa-234m. Since this study was performed using an appropriate 

radiation source and the exact material used in the work coveralls, the results may be applied directly 

to the problem of dose reconstruction. The variability in the mean attenuation ‘appeared’ to be related 

to distance from the source, though no pattern could be identified. For this reason, the value 

proposed for use in dose reconstruction is a constant value of 20% attenuation in shallow dose at 

cancer sites concealed under clothing. 
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5.0 DETERMINATION OF EXPOSURES DUE TO RESIDUAL CONTAMINATION 

REMAINING FROM MED/AEC OPERATIONS

RESERVED
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ATTACHMENT A

TABLES

LIST OF TABLES

Table

1 Plants and buildings used at the St. Louis Downtown Site for AEC/MED 

uranium processing work ..........................................................................................................

Page

79

2 Summary chronology of Mallinckrodt site operations ................................................................ 82

3 Principal changes made in sites, processes, and equipment..................................................... 83

4 Types and quantities of material produced in association with Mallinckrodt 

uranium refining ........................................................................................................................ 84

5 Functional and process keywords and codes............................................................................ 88

6 Thorium and daughter content of the AM-7 residue .................................................................. 91

7 Airborne uranium particle size in process areas........................................................................ 92

8 Uranium dust concentrations by process, Plant 4, 1943–1952.................................................. 92

9 Uranium dust concentrations by operation, Plant 6, 1946–1957................................................ 92

10 Uranium dust concentrations by process, Plant 6E, 1950–1958................................................ 93

11 Uranium dust concentrations by process, Plant 7, 1952–1958.................................................. 93

12 Airborne uranium dust concentrations in Plant 4 Areas, 1948: AEC versus 

Mallinckrodt measured data ...................................................................................................... 94

13 Plant 4 measured daily weighted average exposure concentrations ......................................... 95

14 Plant 6 measured daily weighted average exposure concentrations ......................................... 96

15 Plant 6E measured daily weighted average exposure concentrations ....................................... 98

16 Plant 7 measured daily weighted average exposure concentrations ......................................... 99

17 Plant 7E measured daily weighted average exposure concentrations ....................................... 99

18 Average and highest airborne dust concentrations in the laundry ........................................... 100

19 Job titles and classifications, with function class codes and geometry factors......................... 101

20 Relationship between major processes and job functions ....................................................... 106

21 Uranium dust exposure levels, Plant 4.................................................................................... 107

22 Uranium dust daily weighted average exposure levels, Plant 6............................................... 108

23 Uranium dust daily weighted average exposure levels, Plant 6E............................................. 109

24 Uranium dust daily weighted average exposure levels, Plant 7 (including the 

Slag Separation Plant) and Plant 7E....................................................................................... 110

25 Measured radon concentrations at various indoor and outdoor areas, in units of 

1 x 10-10 Ci/L ........................................................................................................................... 111

26 Middlesex ore storage worker radon exposures, January 1944–January 1949 ....................... 113

27 Potential radon exposures from thorium processing................................................................ 113

28 Ratio of alpha airborne activity to surface contamination......................................................... 113

29 Contamination levels on various surfaces and pieces of equipment in the 

laundry.................................................................................................................................... 114

30 Contamination levels and associated dose rates from work clothing....................................... 114

31 Surrogate (comparable) worker inhalation intake values ......................................................... 115

32 Estimate of annual intake of thorium dust during Plant 7E operation....................................... 117

33 Exposure rates (mR/hr) from drums of K-65 and Q-11 residue in a railcar .............................. 117

34 Middlesex ore storage worker gamma-beta doses, 1944–1949 .............................................. 118

35 Measured dose rates at various positions in Plant 6 ............................................................... 118

36 Weekly external dose values, April 1942–December 1945...................................................... 119

37 Organ doses (rem) for a beam quality (HVL) of 2.5 mm Al and 80 kVp (1942-

1958) ...................................................................................................................................... 119
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38 Decontamination and decommissioning years, 1959-1961: maximum annual 

doses ...................................................................................................................................... 120

39 Postoperations years, 1962-1995: maximum annual doses .................................................... 121

40 Summary of postoperational years exposure data as measured by two survey 

groups..................................................................................................................................... 122
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Table 1. Plants and buildings used at the St. Louis Downtown Site for AEC/MED uranium processing work. Plant Building Process, function, or area Notes

1 25 Lab quality control, R&D Project offices

1 Alley south of 25 Pitchblende R&D extraction

1 K1E Temporary pitchblende ore pilot plant

1 A General plant maintenance (?)

1 X Locker room

2 38B(?) Personnel change house

2 40 Temporary storage of residues Temporary storage of residues

2 45, 45A Warehouse Raw, in-process, and finished materials

2 47 Warehouse Raw, in-process, and finished materials

2 50 Feed material and product storage, UF4 experiments, other Ore concentrates, UO3 product; tanks of stored process liquids

2 51 U3O8 feed preparation (nitric acid dissolution), recovery, other

2 51A Denitration of UNH to UO3, reduction of UO3 to UO2

2 51X Temporary canopy enclosure outside for extraction of pitchblende 

liquor

 

2 52 Ether extraction to produce uranyl nitrate hexahydrate (UNH)

2 52A Miscellaneous(?) re-extraction Pilot plant for countercurrent ether extraction

2 52X Canopy, temporary structure

2 55 Shotgun lab -- temporary structure

4 400 Production of UF4, U metal; slag handling, laboratory, change house; 

metals pilot plant

UF4 product, U recast

4 400, 401 1st floor: casting pilot plant, slag pilot plant, dingot works In casting pilot plant: lead man office and saw room; in slag pilot plant, roll mill, chip 

burning area, and Hoffman cleanup area; in dingot works, blender, saw area, bomb air 

cooler, bomb cooling room, breakout grate, furnace tank pit, and furnace residue pit

4 400 2nd floor: bomb step area; casting area, laboratory area, ceramic pilot 

plant

In bomb step area: blending and bomb areas, KB-2 area, bomb tramrail

4 401, 401A (?) storage, metals and UF4 pilot plant

4 402 Warehouse center

4 403 Machine shop

4 404 Storage room

4 405 Part of the lathe and forging area in 408?

4 406 (A&B) Magnesium storage

4 407? ?

4 408 1st floor lathe & forging areas; storage of slag, dolomite, KOH, NH, 

HF, etc.

Pilot Lab; furnace; forge press; manipulator; lathe area; slag crushing area; salt bath and 

quench tanks; slag storage

4 400 Yards Machine shop (403) yard; guard house and guard tower; incinerator There were also various production dust collectors in these yards outside 406B and 407

6 100 Electrical substation Shed attached to the west end of the building was used for storage of uranium materials.

6 101 Research laboratory, decontamination(?) room, receiving offices Also production offices, shipping & receiving, and decontamination (DX) facilities. Lab had 

hoods, the DX facility a hood.

6 102 Analytical Laboratory (main chemical lab), Control Lab, Metal Lab, 

Sample Prep Lab; possibly an additional R&D lab; lab offices

The Control Lab had a hood. There was a dust collector on the roof.

6 103 Refrigeration, spectrographic laboratory

6 104 Main refinery building: ore to UO3 to UO2 Entire building housed wet processes. Barium sulfate cake, raffinate cake were residue 

products

6 104 Ore Room digest area, feed makeup area, M-20 area, C-3 area Found to be the most contaminated areas in 1958 due to use when pitchblende was main 

feed; beta, gamma especially high in Ore Room and M-20 areas (latter up to 12 mr/hr) 

6 104 Raffinate area, wet pilot plant, ether pilot plant, pot room, pot room 

addition, QM-2 packaging, NA recovery area, MGX area

Housed the continuous process equipment

6 104A, AA Main refinery: ore handling and milling Pitchblende ore (104A)

6 104B Main refinery: pilot plant area

6 105 Main refinery: the “ Ether House” Extraction

6 106 Nitric acid recovery
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Table 1 (continued)

Plant Building Process, function, or area Notes

6 106A Nitric acid recovery

6 107 Nitric acid recovery, tank farm pump house

6 108 Shotgun sample preparation lab Superseded by lab in Building 102?

6 109 Acid unloading station Includes 109, 109A, 109B

6 110 Main warehouse for tagged goods, pitchblende area, ..., UO3 product Pitchblende ore, ore concentrate, UO3 and UO2 product. Was the receiving warehouse for 

pitchblende ore arriving by rail. Found to be moderately and uniformly contaminated 

(including gamma) in 1958 due to trackage, especially on loading platforms.

6 110A Main warehouse, part used as Ledoux Lab

6 110B ....repair

6 111 Sample preparation, North; Ledoux Lab, South; maintenance shops 

in the south end

The Ledoux Lab included the main lab room, the oven room, the weighing room, and the 

dry box room; in these areas were a muffle furnace and dry boxes. It was found to be 

contaminated in 1958; the main(?) room had a walk-in hood behind which the floor was 

heavily contaminated.

6 112 Administration, maintenance shop, instrument shop (including a parts 

room)

This building was 50% offices, uncontaminated, and 50% storeroom and shops, both 

handling some contaminated materials.

6 113 Paint shed

6 114 Scale house, temporary storage of residues For temporary storage of residues, including radium-bearing cake (K-65) in drums. Had 

interior sampling bays and an exterior conveyor. Found to be heavily contaminated in 

1958, including some gamma.

6 115 Boiler house and steam plant

6 119 Steam plant (?), maintenance storage shed

6 120 & 121 U metal dissolver (120) with digest and recovery area (pitchblende); 

pickling building (121)

120 had a sump; 121 had a derby conveyor and pickler. Found to be heavily 

contaminated in 1958.

6 122 Slag recovery pilot plant Found to have loose contamination

6 127 .... & dissociator Ammonia cracking?

6 101 Yard Loading docks Used by Shipping & Receiving.

6 104 Yard Found to be contaminated in 1958. Gamma background from M-20 cell block.

6 105 Yard Outdoor tanks Concrete, asphalt, gravel all found to be contaminated in 1958. Tanks, sump in the M-70 

pit. Hole by 105 door due to cave-in caused by sump leakage.

6 106 Yard Found to be contaminated in 1958.

6 NW Yard Storage yard Found to be contaminated in 1958, including some high gamma spots.

6 108 Yard Laboratory site A laboratory on this site was demolished in 1955, except for the concrete floor. Materials 

handled contained radium.

6 110 Yard Boxcar cleaning site Boxcars that contained pitchblende ore containers were cleaned on the gravel-soil part of 

this yard; there was a 1.5 mR/hr hot spot between the rails and a 5 mR/hr hot spot under 

the shipping dock in 1958. A sewer ran under or over the yard.

6 111 Yard Found to be contaminated in 1958. A sewer ran through the yard.

6 112 Yard Concrete between 112 and 117 The main gate opened into this yard, as did the dispensary doors and a change room. 

Some contamination due to trackage

6 115 Yard Concrete adjacent to the boiler house Dust collector and incinerator created heavy contamination in yard. An ash silo and the 

Hoffman drumming station were located here.

6 116 Yard Storage area Storage of feed materials. Some contamination, including gamma.

6E 116 (including 

116-1 and 116-

2)

UF4-to-metal facility (116-1?) with various operating areas and a 

maintenance shop, residue recovery area, warehouses (116-1 and 

116-2), graphite machining, foremen’s office (116-2?), smoking area

Used for manufacturing UF4 to metal: reduction furnace area (~ 18 furnaces); casting 

furnace area (with multiple furnaces, 4 hacksaws, Kinney pumps, a crushing station, toilet 

area); breakout area (with sump, slag conveyor, chipping station, furnace rebuilding 

station, shop); jolter area (with jolter platform); filling area; center aisle area (with ingot 

mold station?), generator room, and ingot storage area (with both boxed and finished 

ingot storage, “pickled derby” shipping enclosure, ingot room with ingot table). 

Maintenance shop. “F” machine area (2 machines). UF4 slag residues handled in residue 

recovery area, which included sump, filter, and recovery pit. Warehouse (blender hopper 

room, mold outgassing room(?), mold furnace room, loading platforms).
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6E 116B Electrical substation

Table 1 (Continued)

Plant Building Process, function, or area Notes

6E 116C Slag recycle building Magnesium fluoride slag handling and grinding. Had conveyor in (to?) packaging room; 

ball mill; rolling mill; elevator to C-3, hopper. Light dust contamination found in 1958.

6E 117 (including 

117-1 and 117-2)

Service building: clean locker room (117-1?), regulated locker room 

(117-2?), laundry (117-2), security office (117-1), cafeteria (117-2), 

other offices and support services

The laundry and the regulated locker room were the only areas found in 1958 to be 

contaminated. In laundry: lobby, laundry storage room, and small and large sewers in the 

laundry pit. Locker rooms: turnstiles. Clean locker room: clean clothes room.

7 700 Warehouse, safety office, electrical and carpenter shops, slug 

machining (fabrication) plant

The slug machining area had lathes and an inspection area. 700 was found in 1958 to be 

lightly contaminated except for the moderately contaminated slug machining area.

7 701 Slag recovery plant (aka Slag Separation Plant or Slag Processing 

Plant) 

Found in 1958 to be heavily contaminated with loose material, especially with high beta, 

due to substantial amounts of "aged uranium". It had "Wilfey" tables, pumps, tanks, a ball 

mill, drum washer w/elevator and sump, and a filter.

7 703 Hydrofluoric acid vaults, HF tank farm

7 704 HF feed and recovery, HF offgas treatment 704 had sumps, tanks, and a scrubber. 704-707 were all attached and were all found in 

1958 to be moderately contaminated.

7 705 Main processing area for manufacturing UO2, UF4: packaging station 

area, maintenance shop, reactor area

Reactor area had UF4 product hoppers, hydraulic pumps, and access platforms. There 

were a UO3 feed station, a Hapman UO3 conveyor, blenders, a UO2(3?) packaging 

station(s), and a screw storage area. Localized contamination was found in 1958 around 

operating stations; green salt was caked on the roof. See also 704.

7 706 Warehouse for U materials (UO2, UO3, UF4) Had shipping platform. See also 704.

7 707 Ammonia cracking building (manufacturing H2 and N2 from NH3) See also 704.

7 708 Magnesium storage building

7 709 HF refrigeration equipment and pump house Contained refrigeration equipment for the GS(?) pit

7 710 NH3 tank farm and pump house, ammonia storage

7 711 Storage shed

7 

(7E) 

712 Minor elements production facility This had a change room, a "cold" (nonradioactive) lab, a "hot" lab (with hoods and a 

sump), and a production room (with pumper-decanters, mixer-settlers, settling tank, 

packaging station, sumps). It was used to process residues to obtain an ionium (Th-230) 

concentrate. Found in 1958 to have high levels of contamination.

7 SW Yard Storage of drums, feed materials, and contaminated equipment Moderate contamination was found in 1958, but it was highest where spills had occurred 

or equipment was stored.

7 700-701 Yard Secondary gates in 700 yard; conveyor in 701 yard Found in 1958 to have visible fixed contamination, higher near 701.

7 703(?), 704-707, 

and 711 Yards

Storage of contaminated equipment in all yards; Th-230 liquor 

drainage area and boxcar cleaning area in 711 yard

Screws and tubes were stored in the 704-707 yards. Yards were primarily concrete 

around Plant 7, except gravel around 703. Highest contamination (both alpha and 

gamma) found in these yards in 1958 was around 711: in 711 yard, Th-230 liquor was 

drained to the area by the RR tracks.

7 706 Yard UF4, scrap (??)

7 712 Yard Mostly open storage, probably of contaminated scrap Had lean-tos and an open storage bin. There was also a tank farm on a concrete pad at 

the south end of 712. The concrete in this yard was found to be heavily contaminated in 

1958 from the activity from boxcars and contaminated scrap.

Additional Notes:

The information in this table is drawn from FUSRAP 2003a ; MCW 1958; MCW 1959; and ORNL 1981.

Illegible areas in the references are indicated by dots (.....).

Information regarding contamination levels found in 1958 and 1959 is from MCW 1958 and MCW 1959, which are reports on the results of Mallinckrodt's postoperation survey prior to the start of 

intensive decontamination.

Buildings that were remained after the general demolishment of 1959-1961 are shown in Table 42.

Plants 1 and 2 were known collectively as “Main Plant”; Buildings 50, 51, 51A, 52, 52A, and 55 were known collectively as Building 51; and all of the Destrehan site was collectively referred to as 

Plant 6 at times.

1 
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1 Table 2. Summary chronology of Mallinckrodt site operations. 

April 1942 Plant 2 was used to develop a batch process using ether to extract uranium as UO3 from milled ore and 

then to convert the UO3 to UO2. Plant 1 was used for developmental work.

October 1942 Plant 4 was converted for use in the UO2→UF4→U metal process. The ore→UO2 operations continued 

in Plant 2, while miscellaneous activities related to R&D work continued in Plant 1.

April 1943 Production of UF4 from UO2 began in Plant 4.

1944 Experimental extraction of uranium using pitchblende ores began in Plant 1.

1945 At some point in 1945, uranium operations at Plant 1 ceased. Plant 2 was apparently still used for some 

metallurgical-type work. Pitchblende ore began to arrive at the site in greater than research-level 

quantities in about May 1945.

1946 Plant 6 began operation in early 1946, with all ore→UO2 production operations shifted there. Uranium 

operations at Plant 2 ceased in early 1946, the work (including UO3 milling) apparently shifting to Plant 6. 

Only Plants 4 and 6 were in operation.

1949–1950 In 1949-1950, major improvements were made in dust control at Plants 4 and 6, with the latter shut down 

during part of 1949-1950 for this. Ore milling at Plant 6 stopped in 1950.

October 1950 Plant 6E operations began. The UF4→U metal work shifted there from Plant 4, with the UO2→UF4 work 

remaining at Plant 4. Plant 4 was also modified for metallurgical-R&D work and became known as the 

Pilot Plant; some metal production (derbies, dingots) continued to take place there for experimental 

purposes.

1951 Plant 7 operations began in the first half of 1951. At that time, some UF4 production work continued until 

perhaps 1952 at Plant 4, while the UO3-to-UO2 production at Plant 6 seems to have ended completely. 

Instead, UO3 was sent to Plant 7 to be converted in a continuous process to UF4. Some recovery and 

storage operations also shifted to Plant 7.

1952-1953 At some point, the continuous UO3-to-UF4 process began in Plant 7, after which time Plant 6 made only 

UO3.

1954 The Ore Room and K-65 sampling operations in Plant 6 appear to have ended by about August 1954, 

possibly with the shipment of the last of the pitchblende ore (which would have been processed into at 

least 1955). It is not clear when the various Plant 6 pilot plant(s) began, but a 1954 start appears 

reasonable. Also, the Plant 6E Slag Separation Plant started in the first half of 1954. Some reversion of 

UF4 to UO2 and UO3 was done in Plant 7.

1955 In 1955, thorium extraction from AM-7 residue began in Plant 7E and slag processing began in the Slag 

Separation Plant (Bldg. 701, part of Plant 7). Also, processing of residues to extract thorium began in 

early 1955 and the processing of a small amount of "enriched uranium" was done at Plant 7 early in 

1955. Predigestion ore grinding ceased.

1955 or 1956 In late 1955 or early 1956, dingots began to be produced in Plant 4, with derby production only 

intermittent; both were for experimental purposes.

Late 1956 All operations at Plant 4 ceased..

1957–1958 In 1957, all regular site operations ceased, except for some Plant 7 activities that continued until July 

1958. Other postprocessing and shutdown-related activities may have continued into 1958.

 2 

3 
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1 Table 3. Principal changes made in sites, processes, and equipment (e.g., AEC 1949, Miller 1953).

Year Plants New method or form Purpose(s)

1942-

1943

4 Converted Plant 4 U metal production from UF4; UF4 production from 

UO2

1946 6 Began operation of Plant 6 Increase production, reduce all types of exposures 

from processing of pitchblende ores

1948 4, 6 Start of formal health program Track and reduce exposures

1948 6 Used ventilated ore railcars Reduce radon exposures

1949 4 Added ventilation to bomb step and recasting 

furnaces

Reduce deposition of volatiles and thus reduce 

beta exposures; reduce airborne dust levels

1949 4 Redesigned UO2 and UF4 handling methods Reduce airborne dust levels, eliminate hand 

scooping

1949 6 Added ore room shielding Reduce gamma exposures

1949 6 Added remote control for filters Reduce gamma exposures

1949 6 Added K-65 centrifuge shielding Reduce gamma exposures

1949 6 Added shield tanks Reduce gamma exposures

1949 6 Added C-3 cell block shielding Reduce gamma exposures

1949 6 Revised ore house weighing process Reduce gamma exposures

1949 6 Redesigned ore room dust control (drum 

weighing and deheading)

Reduce airborne dust levels

1949 6 Revised UO2 handling (pneumatic unloading and 

conveying systems)

Reduce airborne dust levels, eliminate hand 

scooping

1949 6 Revised UO3 handling (pneumatic unloading and 

conveying systems)

Reduce airborne dust levels; eliminate hand 

scooping

1950 6E Began operation of Plant 6E Increase production, reduce all types of exposures

1951? 6, 6E? Centralized exhaust ventilation Reduce airborne dust levels

1951 7 Began operation of Plant 7 Increase production, reduce all types of exposures

1953 6 Mechanical conveyor facilities provided in the 

Ore Room Addition for handling ore drum lids

Decrease airborne dust levels

1953 6 Increase in capacity of pneumatic gulping system 

in the Pot Room

Lower breathing zone and general area airborne 

dust levels

1953 6 New Ledoux sampling labs with better ventilation Lower airborne dust levels

1953? 4 Enclosure around upper part of casting furnace Reduce dust levels and beta radiation levels when 

furnace lid is removed

1953 4 Enclosure around blender and charging pit Reduce dust levels in general area

1955 6E All-purpose dust hood in the area adjacent to the 

smoking and maintenance areas

Reduce dust levels in breaking out ruptured derby 

furnace shells, rebuilding recast furnaces, 

dumping waste into drums, and dumping out drum 

to sort
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1 Table 4. Types and quantities of material produced in association with Mallinckrodt uranium refining. Material Process or operation Content and form notes Amount

ORES AND OTHER FEEDS

All ores Eldorado processed all the Canadian and part 

of the Congo pitchblende; Vitro, all the 

vanadium tailings and some Congo ore 

(Eisenbud 1975). Some milled ore also came 

from Linde. Mallinckrodt would have received 

most of the Congo only during and just after 

the war; Canadian pitchblende and domestic 

ores were used after that. Pitchblende ores 

were apparently used exclusively until early 

1955.

Mallinckrodt processed up to 50,000 tons of ore 

from 1942-1957 (DOE 1996). Typical amounts in 

1945: Eldorado black oxide, 60,000-160,000 

lbs/month, average 80,000 lbs/month; Vitro black 

oxide, 30,000 lbs weekly; Vitro soda salt, 20,000 

lbs per 10 days (MED 1945a)

Belgian Congo ore 

(Q-11)

Most of the pitchblende processed by 

Mallinckrodt was obtained as a concentrate 

from the Belgian Congo in 1944 and shipped to 

St. Louis in 55-gal drums (AEC 1978).

Pitchblende ore, up to 65% (DOE 1997, MED 

1949) or 70% (Dupree-Ellis et al. 2000) U3O8 by 

weight; up to 100 mg Ra/ton (Dupree-Ellis et al. 

2000); averaged 135 mg Ra/ton (AEC 1949); 

0.185 ppm equilibrium Ra for Q-11 (60%) ores 

(AEC 1949); 0.1 Ci/ton (total?), up to 70% U, 

average U concentration >25%, about 100 mg/ton 

ore for 25% U (Eisenbud 1975)

3400 tons U produced during wartime (through 

1944) (Eisenbud 1975)

Canadian 

pitchblende ores 

(Great Bear Lake, 

Port Hope)

Ores at perhaps 10% U (Eisenbud 1975); ores and 

U3O8 concentrate

850 tons U produced from Canadian ore in 

wartime (through 1944) (Eisenbud 1975)

Domestic ore and 

tailings: Uravan, 

Durango, Grand 

Junction, Naturita 

(Col); Monticello 

(Utah)

During wartime (through 1946?) vanadium tailings 

were used, not fresh ore (Eisenbud 1975); <1% U 

(Eisenbud 1975); shipped as a 20% ore sludge 

(Eisenbud 1975). The US stimulated domestic 

production from 1948 on; ores and lower-grade 

concentrate (DOE 1997). Colorado ores were 

carnotite type (Eisenbud 1975); N. American ore 

contained less than 1% U3O8 (AEC 1978)

850 tons U produced from the vanadium tailings 

(through 1944) (Eisenbud 1975). It is not clear how 

much if any of this was used by Mallinckrodt, 

except such as came as soda salt, etc. 

U3O8

(milled ore or black 

oxide)

Ore usually arrived at Mallinckrodt in milled or 

concentrated form, as black oxide. However, 

DOE 1997 stated that Mallinckrodt produced 

black oxide, presumably at Plant 6.

Originally, (wooden?) beer barrels were used to 

transport the Congo ore from Eldorado, with 

wooden bracing in the railcars. This was 

unsatisfactory so metal containers (barrels) were 

used. The metal containers weighed about 100 lbs 

each when full. (MED 1945a)

Normal in-process inventory circa 1945 was about 

one month's production (MED 1946a).

Sodium diuranate

(soda salt)

Packed in fiber containers (MED 1945a) Vitro converted U ores to sodium diuranate (DOE 

1997); some apparently also came from 

Anaconda, Durango, and Fernald (AEC 1956c). 

Fiber containers weighed about 75 lbs each when 

full (MED 1945a).

 
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Table 4 (Continued)

Material Process or Operation Content and Form Notes Amount

REFINING PRODUCTS

UO3

(orange oxide)

Feed digested in nitric acid; precipitation of RaPb w/ sulfuric acid (pitchblende ores); filtration to 

remove acid-insolubles; sulfate removal w/ Ba 

salt; centrifuging of solution; boiling of “liquor”; 

double extraction of U w/ diethyl ether; water 

wash to remove uranyl nitrate from ether; 

dewatering in Sperry press; boiling of molten salt 

to “hex liquor”; decomposition in pots to form 

UO3; UO3 “gulped” out of pot using vacuum 

system, packed in fiber containers for shipment.

Digestion took 4-8 hours (MED 1946a). Various 

solid and liquid wastes were produced, including 

most of the residues listed below. 2.5-gal fiber 

containers weighed about 75 lbs each when full 

(MED 1945a). 

Sent to Clinton Engineer Works: 30,000+ lbs 

monthly prior to 15 DOE 1944 but 15,000 lbs 

weekly after that. (MED 1945a) 

UO2(brown oxide) UO3 was transferred from fiber containers into 

stainless steel drums, then weighed out on 

monel trays; reduced with cracked ammonia in 

batch electric (muffle) furnace to form UO2

(MED 1949a); scooped from trays into fiber 

containers for transfer elsewhere (MED 

1945a). Packed in fiber containers for transfer 

elsewhere.

This step took about 7 hours (I). 2.5-gal fiber 

containers weighed about 75 lbs each when full 

(MED 1945a). There was 349 lbs on a 4-tray 

charge (MED 1944a).

By DOE 1942 Mallinckrodt was producing a ton a 

day (DOE 1996). Per MED (1944a), Mallinckrodt 

used 32,000 lbs weekly. Per MED (1949a), 

Mallinckrodt produced 2/3 of the US total; 64% of 

what it made stayed at Mallinckrodt, 20% went to 

Harshaw, and 16% went to Linde. Per MED 

(1946a), in 1944-45, 20,000 lbs monthly went to 

Linde; 10,000 lbs/week went to Harshaw from 

Sept-Oct 1944, 28,000 lbs in DOE, and 13,000 

lbs/week after that.

UF4

(green salt)

UO2 placed on graphite or nickel trays in 

graphite or nickel boxes in the hydrofluorination 

reactor (furnace); HF gas passed over it to 

form UF4; UF4 removed from furnace and put 

through pulverizer; UF4 packed into fiber 

containers (MED 1945) or 5-gal containers for 

transfer to Plant 4 or 6E or another site (AEC 

1949).

Fiber containers weighed about 75 lbs each when 

full (MED 1945a). In 1944, one control sample 

was taken per charge; there were 107 runs per 

week; 535 lbs per week was sent to a recovery 

tank; and there were 135 lbs per drum and 3375 

lbs per lot (MED 1944a).

Mallinckrodt was the major producer of UF4; up to 

1949, some UF4 came from ElectroMet (DOE 

1997). Normal in-process inventory circa 1945 was 

2 days’ production (MED 1946a). In 1944, 37,000 

lbs weekly was produced, of which 11,000 lbs was 

used at Mallinckrodt to make metal (MED 1944a). 

In 1945, 8000 lbs/week was being sent to Harshaw 

and 20000 lbs/week to Iowa State (MED 1945). 

U metal in derby 

form

Reduction with magnesium in furnace to U 

metal (slag + derby); slag chipped off to leave 

derby

In 1944, there was 135 lbs of UF4 per bomb, along 

with 55 lbs of liner and 24 lb of magnesium; a 

biscuit weighed 92 lbs and was associated with 

about 122 lb of slag, of which 10 lb was metal; 

about 80 lbs of sawdust a week was produced; 15 

and 2 lbs of samples were sent (weekly?) to plant 

and outside labs respectively (MED 1944a).

In 1944, there were 7500 lb of biscuit produced 

weekly (MED 1944-TTTTT). Normal in-process 

inventory circa 1945 was one week’s production 

(MED 1946a).

U metal in billet form Derby was vacuum recast to form the billet Normal in-process inventory circa 1945 was one 

week’s production (MED 1946a). In 1945, billets 

were being shipped out every other week to 

Hanford in a carload lot of about 30,000 lbs, from a 

weekly production of 13,000-15,000 lbs (MED 

1945a).

U metal in dingot 

form

A dingot was a single massive ingot needing 

no recasting. The dingot-making operation was 

most similar to the regular derby-making 

operation. After the chipping step, the dingot 

was pressed into a slab. (AEC 1956b)

In late 1955 or early 1956, this replaced the derbybillet operation, except for occasional experimental 

derby production in Plant 4, per AEC (1956b). But 

AEC 1956d reports that in mid-1956 (all?) billet 

recasting was being done in 6E, using new 

graphite molds. A dingot weighed about 3300 lbs 

(AEC 1956b).

 
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Table 4 (Continued)

Material Process or Operation Content and Form Notes Amount

Radioactive metal 

samples

“Small” samples were sent to Clinton Engineer 

Works (Y-12) packed in glass tubes and 

packed into cardboard boxes. “Eggs” were sent 

to the Chicago Area Engineer (MED) packed 8 

to a box in wooden cardboard boxes. (MED 

1945a)

A “small sample” was sent daily to Y-12; eggs 

were sent in 2-3 lots daily, 63 lbs to a lot. (MED 

1945a)

"Tubealloy" Early synonym for uranium (Fleishman-Hilliard 

1967), but in what form is unclear. 

Manufactured by Mallinckrodt and shipped to 

the Chicago Area Engineer (of MED).

150 lbs shipped daily to Chicago (MED 1945a)

RECYCLED AND RECOVERED MATERIALS

Organic solution of 

Th(NO3)4

AM-7 residue was processed via a nitric acid 

strip to concentrate Th-230 in Plant 7E; 

solution was sent to Mound, residue (AM-9) 

returned to storage.

Concentrate had about 1 kg of Th-230, 0.7% 

alphas from Th-227 & daughters, <0.03% from Th228.

350 tons of AM-7 processed for Th- 230 (FUSRAP 

Undated a, AEC 1959); 3600 gal (13,630 l) of Th230 solution sent to Mound (DOE 2002).

U slag (derby) Derby slag was scalped or cut off derby and 

separated into a MgF2 part and a C-liner part; 

the MgF2 part was sent to Vitro for recovery; 

the C-liner part was stored as waste. 

Eventually both parts were processed at 

Mallinckrodt.

Some of the C-liner slag was apparently 

reprocessed to recover U from about 1953 on 

(AEC 1954f). In 1955, an interim pilot plant at Plant 

7 was built to scalp off most of the U-bearing 

segment of the MgF2 part.

U slag (recast) Recast slag was recovered as residue from the 

recast furnace.

Recast furnace slag was highly concentrated in 

UX1-UX2 (beta emitters) (AEC 1949, Eisenbud 

1975) .

U slag (dingot) Dingot slag was broken off and swept down 

through a floor grill, collected on a conveyor, 

put through a grinding series, drummed, and 

sent to the Slag Building (701) for 

reprocessing. (MCW 1949c)

U scrap Miscellaneous material, including some 

residues, ash from incinerating the UO3 fiber 

containers, and metal; some oxide and nitrate 

scrap was sent from the Chicago Area 

Engineer (MED); some scrap was sent to Du 

Pont. C-2 scrap was packed in 50-gal whiskey 

barrels; C-1, C-3, C-4, C-5, and D-2 scrap was 

packed into 5-gal containers with a steel clamp 

top. (MED 1945a)

Scrap from Chicago Area Engineer, 1945: 1500 lbs 

oxide type per 2 months, 1500 lbs nitrate type per 

4-5 months. Scrap shipped to Du Pont as follows. 

C-1, C-3, C-5, D-2: 80,000-90,000 lbs total per 4-6 

weeks; C-2: 80,000-90,000 lbs per 5-6 weeks; C-4: 

100,000-120,000 lbs per 4 months. (MED 1945a)

K-65 One source suggests that this residue was 

"reworked to recover additional uranium 

values" (i.e., reprocessed?) before transfer to 

Lake Ontario.

Radium-bearing residue

Sawdust and fiber 

containers

Fiber containers were incinerated and 

processed to recover uranium; sawdust was 

apparently processed similarly.
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Table 4 (Continued)

Material Process or Operation Content and Form Notes Amount

RESIDUES AND OTHER WASTES

Pitchblende raffinate 

(AM-7)

"Airport cake"; produced as part of the 

digestion process

0.2% U (U, pg 48)); 29 ppm Th-232, 3.8 ppm Th230 (11.6 isot %) (Figgins and Kirby 1962). 

Pitchblende residues:Th-232/Th-230 ~ 8.

33000 lbs/day (AEC 1949); 74K tons total, 113 

tons U stored through about 1953

De-thoriated 

pitchblende raffinate 

(AM-9)

Residue after processing AM-7 for ionium (Th230)

Domestic ore 

raffinate (AM-10)

"Airport cake" (different from AM-7); produced 

as part of the digestion process

Carnotite residues: Th-232/Th-230 ~ 15-20 32.5K tons total, 48 tons U stored through about 

1953 (AEC 1959)

Pb-Ra precipitate

(K-65)

"Lead cake"; appears to be a subset of AM-7; a 

Ra-bearing residue produced as part of the 

digestion process (first precipitation)

750-900 mg Ra/ton (AEC 1949); 750 mg/ton Ra 

and 0.2% U (AEC 1949); up to 300 Ra mg/ton 

(Eisenbud 1975)

8000-12000 lbs/day (AEC 1949); K-65 part of the 

AM-7 (AEC 1959)??

Ba sulfate cake

(AJ-4)

Produced as part of the digestion process 1 mg Ra/ton, 0.1% U (AEC 1949) 6800 lbs/day total AJ-4 (AEC 1949); stored: 1.5K 

tons total unleached (wit 22 tons U), 8.7K tons total 

leached (with 7 tons U) (FUSRAP Undated a), i.e., 

10.2K tons total (with 29 tons U (AEC 1959) 

Sperry cake Produced in Sperry press from aqueous tails 

from the ether extraction step; some later sent 

to Mound for extraction of Pa

Good source of Pa-231 (2 g/20 tons); per Salutsky 

(1956), this was a cake @ 50% solids, 1.6 g/cm3

density, 0.1-0.3 ppm Pa-231

20 tons (AEC 1959)

Vitro residues (C-6) Sent from Vitro (?) for storage at the airport 290 tons total, 1.9 tons U (AEC 1959)

Bomb furnace 

residue (C-Special)

This was from the Mallinckrodt and ElectroMet 

bomb furnaces and appears to be the same as 

the "C-liner slag" below. (AEC 1949) 

U-containing sands, 

precipitates (V-10)

Captured from the Japanese 60 tons total, 0.2 tons U (AEC 1959)

Dolomite liner (Cliner or C-liner slag)Slag material, mainly dolomite, remaining after 

the derby slag was separated from the derby 

and the top (MgF2) part of the slag was 

detached. 

This remainder slag was produced until early 1953, 

when dolomite was replaced by recycle 

magnesium fluoride. Some was reprocessed to 

recover U from about 1953 on (AEC 1954e). U 

content: <2%.

7800 tons, 122 tons U (1959)

Interim Residue 

Plant Tailings

(C-701)

Resulted from scalping the U content from the 

Mg fluoride slag from 1955 on; 701 apparently 

refers to Bldg 701

 7K tons total, 144 tons U (AEC 1959)

30- and 55-gallon 

drums

Empty drums, stored as contaminated waste 55000 (AEC 1959)

Metal and alloy 

scrap

Stored as contaminated waste 3500 tons (AEC 1959)

Aqueous tails From the ether extraction step 

NOTE: See the text for process details. See the keywords table (Table 5) for other code numbers and terms.
Effective Date: 10/24/2003 Revision No. 00 Document No. ORAUT-TKBS-0005 Page 88 of 124

Table 5. Functional and process keywords and codes. Plant Keyword Note

7 7E 7 East (Building 712), where the processing of AM-7 to extract thorium was done

7 7W 7 Warehouse? 7 West?

6 Acid Acid addition: in ore digestion (nitric) or Pb-Ra precipitation (sulfuric)

6 AEC Atomic Energy Commission

6 Airport The former airport site later used for waste storage by the AEC

4,6,6E,7 Area M Area mechanic

6 Assist LO Assisting the lead operator, assistant lead operator

4 Ballard Vertical turret lathe (manufactured by Ballard) used to scalp the dingot

6 Barium Barium salt addition

6 Bird centrifuge Solid-bowl centrifuge (manufactured by Bird) used to separate liquids and solids or 

heavier and lighter liquids in the digestion process

 Black oxide U3O8

4, 6 Blender Apparently ore in Plant 6; UF4 + Mg in Plants 4, 6E; slag in Plant 6E

6 Boildown A step between digestion and ether extraction

4 Bomb Container for Mg-UF4 in the metal reduction process

4, 6E Bottom Lower (furnace or F machine)

4 Box Crucible holder (external assembly)

6E Breakout Removal of the derbies from the bombs

 Brown Brown oxide, i.e., UO2

---- Brushing Same as chipping

6E Burnout, burn Crucible burnout (heating)

6 C- When followed by a number, denotes material collected by a given dust collector or 

filter press or a type of scrap

6 C-3 (C-3A, etc.) C-3 cell block operations (e.g., packing, centrifuging) related to dust or cake collection

C-liner Refers to the slag left after the bomb is opened and the derby removed

4, 6E Cage Scrap area

6E Cap Putting the top on the crucible assembly

6 Centrifuge, CEN, Cent, Furn Cntr Also abbreviated CNF?

--- Change Room Locker room, where clothes were issued by the porters

6E Charge Room, Charging Room Area where bomb was loaded ("charged")

4 Chipping Removal of the slag from the derbies using a manual or power hammer

--- CL Cleanup? Cell?

--- Cleanup Generic for area or item cleanup: see associated process keyword (e.g., TA-7)

6 Cloth Cloth (actual or metal) used to filter solids from liquid streams

6 CM Cloth man?

4, 6 Continuous furnace, Cont furn Unclear which furnace (probably the recast furnace) (6 should probably be 6E)

--- CROM Cloth Room?

Croppings Chips or pieces taken off derbies, billets, and dingots

6E Crucible, Cruc Crucible and/or assembly (if 6 shown, probably a typo -- should be 6E)

6 D- When followed by a number, denotes a given dust collector or filter press OR the 

material collected there or a type of scrap

7 D-30 Dust collector for the FMFL product (dust is the product)

6 DA Dissociated (cracked) ammonia

6,6E, 7 Decontamination Generic

Deheading Removing the lid of a drum, usually of ore

--- DEN Probably same as DEV

--- DEV Development, as in PRO (product) DEV and MET (metallurgical) DEV

6 Digest, DIG Digestion process (ore dissolution by acid)

4 Dingot Extra-large ingot produced instead of derbies and billets

6 Diuranate Sodium diuranate (sodium salt), also called soluble feed

DR DX?

--- DX Probably decontamination

Egg Piece of U metal in the shape of an egg, produced for external assays

6 Ether, Ether House, EH Ether storage, handling for U solvent extraction process

--- Exp, Expansion, Exp Eng, Exp Off Expansion -- apparently plant expansion or modification work (Expansion Office)

--- Expeditor, construction expeditor Possibly related to Expansion (q.v.); apparently a coordinator or engineer

4, 6 Extra, Extra Man Extra (floater) for a process or area (4, 6); possibly "extraction" (6)

6 Extraction Extraction of U in ether extraction process

6 FE Feed? Feinc?

6 Feed Ore or other feed material; also could indicate main stream of process (i.e., not 

residue)

6 Feinc String discharge rotary vacuum filter (manufactured by Feinc) used to separate solids 

from liquids

6 FH Feed hopper?
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Table 5 (Continued)

Plant Keyword Note

6 Filter press Press type of filter for separating solids from liquids and leaving a dewatered cake

6E F Machine UF4-to-Derby bomb filling ("F") machine: the top is used for the UF4, Mg mixing, the 

bottom for filling the bomb

7 FMFL Fluorinated MgF2 liner, used as a low-hydrogen liner in dingot bombs

4, 6 FR, Furnace, Furn Furnace Room

--- FTO Fork truck operator

6 GLC Gangue Lead Cake, i.e., the Pb-Ra cake

Green Green salt, i.e., UF4

6 Gulping Vacuum-sucking UO3 out of the pot it was produced in into drums

H-20 ??? (H2O, i.e., water?)

7 Hapmann Conveyor for loose material; typically discharges into a feed hopper

6 Hex, hex liquor Uranyl nitrate hexahydrate in an aqueous solution

6E Hoffman Dust collector (manufactured by Hoffman)

All House Generic for dedicate additive handling and storage building or area

6 Instrument, Instru, Instr Instrument Shop

4, 6E Jolter, jolting Jolter-filling machine in UF4-to-U metal production, also the operator; the operation

6 K-65 K-65 (Ra-Pb) residues (in the ore-to-UO3 process)

4 Label Apparently labeling product containers

6 Laboratory, Lab, LAB Unclear which lab is meant (Analytical, Shotgun, Ledoux, Research, etc.)

6 Laundry One each for contaminated and uncontaminated clothing; Mallinckrodt had own 

contaminated laundry from at least 1948 on

6 Leach Oliver Cell Leaching cell in the extraction process

6 Ledoux (LeDoux) Laboratory, LL The raffinate and U assay laboratory in Plant 6

LG Lead gangue? See GLC

--- Lime In the derby production process, lime or dolomite (i.e., calcium or dolomitic quicklime)

6 Liquor Extracted liquid concentrate, usually after removal of undesirable materials as solids

4, 6 Loading, Load, Ld Generic: see associated process keyword, e.g., TA-7

--- M Apparently generic for maintenance

6 M-20 Process ???

--- MFG Manufacturing: generic for production processes, as opposed to development

--- MGH Same as MGX?

6 MGX, MgX, MGX Process Unclear: Milling & Grinding × (uranium)? Manufacturing × (i.e., uranium products)?

4 Mag Room Magnesium [Storage] Room

6, 6E Maintenance, M, MNT Maintenance, presumably process maintenance

--- ME (1) Mechanic, mechanical work (often with location designation: e.g., ME 

Powerhouse) (2) Minor Elements (see MEP)

7 MEP Minor Elements Production Facility (Building 712, where, e.g., AM-7 was processed to 

extract thorium)

7 MFL MgF2 liner (for use in dingot bombs)

--- Mtns Maintenance

--- Mtrs Typo for maintenance (Mtns)?

4 Muffle Batch electric furnace used to reduce UO3 to UO2

6 NA, NAH Nitric Acid House

6 Neutral Neutralizing uranium solution (in acidification step?)

6 Niagara, Nia Pressure leaf filter (manufactured by Niagara) used to separate solids from liquids

6 NOK(?) HOK? H20-uranyl-nitrate-hexahydrate is called "OK hex liquor" (I) -- ???

6 Ntns Typo for Mtns?

4 Office, OFF Generic for office

6 Oliver Filter press (manufactured by Oliver)

Orange Orange oxide, i.e., UO3

6 Ore, Ore Room, OR Ore processing, storage before use; also handling of residues

4, 6, 7 Pack, packing, packaging Appears to be generic for packaging of uranium products

--- PEW Powerhouse

--- PH Probably Powerhouse

6E PLO Process Lead Operator

4 Pole, poling Pushing the charge down into the dingot bomb with a steel rod

6 Pot, Pot Room Denitration pot room (producing UO3)

6E Pouring Part of the recast process, possibly the extraction of the stop to allow the molten metal 

to flow into the mold

6E Pickling Soaking derbies in acid to remove surface impurities (scale, oxides)

4, 6 Pilot Plant, PP Plant 4's name (for metallurgical research and UF4 work), after metal production 

moved to 6E; also, a slag recovery pilot plant in 6E and an unspecified pilot plant at 6

4 Pl.7 Plant 7 (apparently Plant 4 person detailed to Plant 7)
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Table 5 (Continued)

Plant Keyword Note

6 Powerhouse Power-generating or distribution area; may be same as Boiler House

6 Press Filter press

6 Production Engineering, Prod Eng, 

Prod Off Eng

Generic for production/process engineering work

6 Production Office, PO The MCW [Main] Production Office

6 Raffinate, Raff Residues (mainly the cakes)

4,7 Reactor Reaction vessel/heater in which UO3 is converted to UO2 with cracked ammonia

4 Recast, Recasting, Recast Furnace Where derbies are recast into billets

4 Receiving & Shipping Generic? The main Receiving & Shipping area was in Plant 6 

6 Recovery, REC Nitric oxide recovery (6); other types of recovery at other plants

6 REF Refinery (i.e., Plant 6)

Regulated Refers to radiologically controlled areas, e.g., the regulated locker room versus the 

general (clean) locker room 

6 Reverter (riverter?) Unclear. Appears to be a type of pump or processor used in the recycling (reverting) 

of dust collector material, probably to convert UF4 back into UO2 or UO3

7 RMF Recycle magnesium fluoride (from bomb liner residue)

6E Rockwell Rockwell furnace (UF4-Mg reduction to derby metal)

6 Rover Extra pair of hands?

6E Ruemelin Exhaust hood (from the name of its manufacturer)

6E RUNOUT Could be a typo of "Burnout"

4 Salt bath Bath in which a dingot is put to heat it

--- SAM, SAMPLING PLAN Late 1950's: apparently general sampling but may have been a special project

--- Sample Room Any of several sample rooms, e.g., the K-65 Sample Room

6 Sampler, Spl Generic for sampling or work in a sample room

6E Saw Sawing off a portion of the metal billet as a sample

6E Scalping Mechanically slicing off slag from the external surfaces of a billet

6 Scrap Usually, scrap left over after cleaning the derbies; miscellaneous scrap

7 Slag Processing Plant Same as Slag Separation Plant

[4] Slag Shed Storage of the chipped-off bits of slag?

6E Slye Dust collector (manufactured by Slye)

6 Shotgun Laboratory Shotgun sample preparation laboratory

6 Shotgun Samp Prep, SSP Shotgun sample preparation (lab)

6 Soda salt Sodium diuranate (Na2U2O7, from Vitro or another site)

6 Soluble feed Usually sodium diuranate (see Soda salt)

Spec Anal Lab Spectrographic Analysis Laboratory

6 Sperry Filter press (manufactured by Sperry); also, the cake produced

6 Storeroom, SR Generic; the main storeroom was in Plant 6

6 Stripper Unclear; probably piece of equipment rather than operator name

6 T-3 Typo for C-3?

6 Thawhouse Building where ore was thawed in winter before undrumming

6 THP THP ether?

--- Tinner Worker with tin and sheet metal (presumably in the tin shop)

4, 6E TM Top man, Top-off man, etc.

4, 6E Top Refers to the upper or top recasting furnace or the top F machine

4, 6E, 7 Topping Refers to the operation of adding material to ensure proper fill

6 UNH Uranyl nitrate hexahydrate

6E Utility Utility man or "floater" who filled in as needed

All Warehouse, WH Generic: various warehouses were used, with the main one in Plant 6 and later in 

Plant 7

6 Wash Oliver Cell Wash cell in the extraction process

4 Weigh, check weigh Generic: feed and products were weighed at various points

6? Weighmaster Apparently the "official" certifier of weights of uranium products and possibly ore

6E Wilfey Shaker table (manufactured by Wilfey) used to separate high-U slurry from low-U 

material

Codes:

Pitchblende ore Q-11 (high-grade); AQ-4

Black oxide (U3O8) 162, 172, or 182; Chemical A (not of spec grade), K-35, C-55, or K-82; GY-3

Sodium salt ( Na2U2O7) Chemical S

Magnesium uranate MgX

Calcium uranate CX

Miscellaneous feed F

UNH (U nitrate hexahydrate) SC-5

Orange oxide (UO3) QM-2; 264 or 272

Brown oxide (UO2) LF-9; 306
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Brown oxide (UO2) YB-1 (when produced from UO3 in the experimental continuous reactor)

Table 5 (Continued)

Codes: (Continued)

Green salt (UF4) TA-7

Green salt (UF4) TZ-7R (when produced from UO2 in the experimental continuous reactor)

Derby (rough U metal form) KB-2

Billet (finished U metal form) YM-5

Uranium metal (generic) Tubealloy

Magnesium metal M

Magnetically fused dolomite OZ

Uranium metal scrap ST

“Radium element” D

Plants 1 and 2 Project 89

Plant 4 green salt operation Project 90

Plant 4 derby production Project 91

Plant 4 recast operations Project 92

The Mallinckrodt St. Louis site Location E

Table 6. Thorium and daughter content of the AM-7 residue ("airport cake").

Material

Concentration, 

ppm

Mass, 

grams

Specific activity, 

Ci/g*

Total curies in 

ore Notes

Ore --- 3.18 × 108 --- --- 350 tons processed (see Table 4)

Th-230 3.8 1,208 0.0202 24.4 Concentration: Figgins and Kirby 1962

Th-232 29 9,222 1.09 × 10-07 0.00176 Concentration: Figgins and Kirby 1962

*From Shleien 1992

Isotope Half-life

Activity after 15 

years, Ci

Percentage of 

original Notes

Th-230 Chain

Th-230 77,000 years 2.44 × 101 ~100%

Ra-226 1,600 years 1.58 x 10-1 0.65% Ignoring Ra-226 decay

Rn-222 3.82 days 1.58 x 10-1 0.65% Ignoring Rn-222 decay

Po-218 3.05 minutes 1.58 x 10-1 0.65% Ignoring Po-218 decay

Pb-214 26.8 minutes

Bi-214 19.9 minutes

Po-214 164 microsecs

Pb-210 22.3 years

Bi-210 5.01 days

Po-210 138 days

Th-232 Chain

Th-232 1.41 × 1010 years 1.01 × 10-3 ~100%

Ra-228 5.75 years 8.46 x 10-4 84% Ignoring Ra-228 decay

Ac-228 6.13 hours 8.46 x 10-4 84% Ignoring Ac-228 decay

Th-228 1.91 years

Ra-224 3.66 days

Rn-220 55.6 seconds

Po-216 0.15 seconds

Pb-212 10.6 hours

Bi-212 60.1 minutes

Notes

Fifteen years is assumed to be the maximum decay time (1942-1957).

As the second part of the table shows, in 15 years secular equilibrium has not been reached for either the Th-230 chain or the Th-232 chain, 

although Th-232 is almost there. Thus only the activities of the first few members are shown.
Effective Date: 10/24/2003 Revision No. 00 Document No. ORAUT-TKBS-0005 Page 92 of 124

Table 7. Airborne uranium particle size in process areas

(Sanders 1975, Table 2).

Area and U type

Mass median 

diameter, μm

% Activity from

<7 μm particles

Foundry

Depleted U 2.8 ± 2.7 ┐

├─

┘

Enriched U 3.3 ± 2.2 88

Enriched U 2.1 ± 2.0

Machining

Depleted U, 

milling dry 3.0 ± 2.3 ---

Extruding

Depleted U 3.2 ± 2.7 ---

Table 8. Uranium dust concentrations by process, Plant 4, 1943–1952 (Mason 1958a, Table 2).

Year

Concentration, multiple of 70 Alpha dpm/m3 Concentration, alpha dpm/m3

UO2

handling

UF4

production

KB-2

production

YM-5

production

UO2

handling

UF4

production

KB-2

production

YM-5

production

1943 30 34 17 36 2,100 2,380 1,190 2,520

1944 30 34 17 36 2,100 2,380 1,190 2,520

1945 30 34 17 36 2,100 2,380 1,190 2,520

1946 30 34 17 36 2,100 2,380 1,190 2,520

1947 30 34 17 36 2,100 2,380 1,190 2,520

1948 30 34 17 36 2,100 2,380 1,190 2,520

1949 6 4 4 11 420 280 280 770

1950 4 2 3 11 280 140 210 770

1951 4 3 280 210

1952 4 3 280 210

Note: The last two processes were moved to Plant 6E in 1951, the first two to Plant 7 in 1953.

Table 9. Uranium dust concentrations by operation, Plant 6, 1946–1957 (Mason 1958a, Table 1).

UO3 production UO2 production

Year Warehousing Ore grinding Feed digest Milling Pot room Packaging Load Unload Packaging

1946 3 190 6 180 111 76 45 161

1947 3 195 6 180 111 76 45 161

1948 3 195 6 180 111 76 45 161

1949 180 111

1 5 1 0 60 10 20 10 5

1950 * 11

0.4 5 1 5 10 10 5 5

1951 0.5 5 2 2 1 6 3 5

1952 0.5 3 5 3 2 6 3 5

1953 0.9 2 0.7 3 2 ** ** **

1954 0.3 2 0.6 2 2

1955 0.3 * 0.8 2 2

1956 0.3 0.4 3 4

1957 0.3 0.8 3 1

Units are multiples of 70 alpha dpm/m3. 

Notes

The above operations were done in Building 51 (Plant 2) up to 1946, then moved to Plant 6.

Various significant improvements in dust control were made in Plant 6 in 1949-1950.

Milling of UO3 was discontinued in 1950 (*).

UO2 production work was transferred to Plant 7 in October 1952 (**); the rest remained at Plant 6.

All work was transferred to Weldon Spring in March 1957 and all remaining plants were closed.
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Table 10. Uranium dust concentrations by process, Plant 6E, 1950–1958 (Mason 

1958a, Table 4).

Year KB-2 production YM-5 production

Average High Source of high Average High Source of high

1950 0.1 0.3 Charging 1.0 1.8 Crucible assembly

1951 0.3 0.8 Residue 1.1 1.8 Top furnace

1952 0.4 1.0 Residue 1.2 2.3 Burnout

1953 0.3 1.2 Charging 0.5 0.7 Top furnace

1954 0.5 3.0 Residue 0.9 1.7 Bottom furnace

1955 1.6 4.0 Residue 0.6 1.5 Bottom furnace

1956 0.4 1.6 Capping 0.5 0.6 Bottom furnace

1957 0.4 1.5 Burnout 0.7 1.2 Bottom furnace

1958 0.8 2.1 Breakout 1.2 2.1 Bottom furnace

Units are multiples of 70 alpha dpm/m3. 

Notes

KB-2 is the derby form and YM-5 the billet form of uranium metal.

Table 11. Uranium dust concentrations by process, Plant 7, 

1952–1958 (Mason 1958a, Table 3).

Year Average High Source of high

1952 0.5 1.6 UO2 dumping

1953 0.4 1.7 Furnace operation,TA-7 packaging

1954 0.5 7.0 Sampling and cleanup

1955 0.3 1.1 UO2 dumping

1956 0.4 0.8 UO2 dumping

1957 0.3 0.8 UO2 dumping

1958 0.5 1.4 TA-7 packaging

Units are multiples of 70 alpha dpm/m3. 

Notes

Operations were moved to Weldon Spring in 1958.
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Table 12. Airborne uranium dust concentrations in Plant 4 Areas, 1948: AEC versus Mallinckrodt 

measured data (AEC 1949, Table 3).

Activity

Multiple of

"Preferred Level" ug/m3 Adjusted to dpm/m3

AEC Mallinckrodt AEC Mallinckrodt AEC Mallinckrodt

LF-9 Loading

Operator A 47.7 29.7 2.39E+03 1.49E+03 1.99E+03 1.24E+03

Operator B 47.7 30.7 2.39E+03 1.54E+03 1.99E+03 1.28E+03

Furnace Tending 

Operator A 7.5 5.7 3.75E+02 2.85E+02 3.13E+02 2.38E+02

Operator B 9.1 6.1 4.55E+02 3.05E+02 3.79E+02 2.54E+02

TA-7 Unloading

Operator A

- Manual 186 66.8 9.30E+03 3.34E+03 7.75E+03 2.78E+03

- Semi-mechanized 16.4 8.20E+02 6.83E+02

Operator B

- Manual 186 57.2 9.30E+03 2.86E+03 7.75E+03 2.38E+03

- Semi-mechanized 20.7 1.04E+03 8.63E+02

TA-7 Mixing & Packing

Operator A 63 13.1 3.15E+03 6.55E+02 2.63E+03 5.46E+02

Operator B 57 24.6 2.85E+03 1.23E+03 2.38E+03 1.03E+03

Bomb Charging 51 42.4 2.55E+03 2.12E+03 2.13E+03 1.77E+03

Topping 10.4 32.4 5.20E+02 1.62E+03 4.33E+02 1.35E+03

Jolting 51 7.5 2.55E+03 3.75E+02 2.13E+03 3.13E+02

Charge Firing 13.3 13.8 6.65E+02 6.90E+02 5.54E+02 5.75E+02

Derby Unloading

Operator A 5 9.1 2.50E+02 4.55E+02 2.08E+02 3.79E+02

Operator B 5 18.7 2.50E+02 9.35E+02 2.08E+02 7.79E+02

Chipping 26.3 11.5 1.32E+03 5.75E+02 1.10E+03 4.79E+02

Slag Handling 1.6 2.2 8.00E+01 1.10E+02 6.67E+01 9.17E+01

Top Furnace Tending

Operator A 61 36.5 3.05E+03 1.83E+03 2.54E+03 1.52E+03

Operator B 61 23.2 3.05E+03 1.16E+03 2.54E+03 9.67E+02

Bottom Furn Tending 73 59 3.65E+03 2.95E+03 3.04E+03 2.46E+03

Sawing 15.8 5.4 7.90E+02 2.70E+02 6.58E+02 2.25E+02

Cage Handling 2.7 52 1.35E+02 2.60E+03 1.13E+02 2.17E+03

Office 0.6 4.1 3.00E+01 2.05E+02 2.50E+01 1.71E+02

Shipping & Receiving 1.6 6.8 8.00E+01 3.40E+02 6.67E+01 2.83E+02

Mechanics 5 10.7 2.50E+02 5.35E+02 2.08E+02 4.46E+02

Carpenter 2 4.6 1.00E+02 2.30E+02 8.33E+01 1.92E+02

Porters 0.8 2.9 4.00E+01 1.45E+02 3.33E+01 1.21E+02

Guards 0.4 1.4 2.00E+01 7.00E+01 1.67E+01 5.83E+01

Notes

At the time of this report, AEC's "preferred level" for U dust in air, 50 μg/m3

, was based on an assumption of exposure 8 

hrs/day, 6 days/week (AEC 1949). Columns 2 and 3 are based on this.

Columns 4 and 5 represent the values in Columns 2 and 3 respectively multiplied by 50 μg/m3 and adjusted to 40 hours per 

week.
Effective Date: 10/24/2003 Revision No. 00 Document No. ORAUT-TKBS-0005 Page 95 of 124

Table 13. Plant 4 measured daily weighted average exposure concentrations. Weighted average concentration, alpha dpm/m3

 

Occupation 5/56 11/53 3/53 6/50 10/49 9/48 5/48 AEC 1949 AEC 1951

Magnesium operator 35 70

Lime blender 35 70

Slag man 70 105 210 140

Derby unloader 175 245 1,260 280

Bomb topper 210 280 2,310 840

Charge firing 140 350 980 910

Derby chipper 140 350 910 1,890

Jolter 70 140 490 3,500

Bomb charger 210 490 3,010 3,640

Green lead man 70 140

Cleanup man 140 140

Furnace tender 70 70 350 560

Furnace box puller 35 140 560 630

TA-7 Pilot Plant 980 175

Brown loader 280 350 2,240 3,360

Green packer 245 210 1,750 3,990 7,210

Green miller and mixer 70 140 980 4,690

Green unloader 210 490 1,540 13,020

Plant superintendent 7.3

Technical supervisor 6.6

Engineers 7.3 9.8 14

Chief chemist 5.9

Vacuum fusion chemist 39

Vacuum fusion technician 59

Microscopist 18.4

Chemist 10

Chemical technician 10 4.6 7

Foreman 22.5 6.7 12 35 70 175

Shift foremen 12.4 56 98 175

Lead operator 25 8.2 19 119 63 ---

Dingot/bomb, slag grinding oper 85 33 64 X X X X

Furnace and saw man 17.5 X X X X

Casting furnace operator 10.8 110 480 5110

Furnace operator (UF4-derby?) 91 70 570

HF operator 91 70 570

UO3 & Brown packer 217 322 2,730 4200 2,730

Green packing operator 196 315 7,210 4,000; 13,000

Asst green packing operator 112 133 2,800

Residue 27.4

Ceramic 14.8

Vertical lathe 28.5

Forge press lead operator 22

Forge press salt bath man 21.5

Forge press manipulator (oper'r) 22.6

Forge press operator 21.9

Clerk 5 42

Guard 7.1 28

Porter 40 2.7 5.8 56

Area mechanic --- 22 15 84 112 350 350

Notes

Data from the surveys of 6/50, 10/49, 9/48, and 5/48 is from AEC 1950m; data from the surveys of 3/53 and 11/53 is from AEC 1954b; and data from the survey 

of 5/56 is from AEC 1956b. Other data is from the references given in the column heading.
Effective Date: 10/24/2003 Revision No. 00 Document No. ORAUT-TKBS-0005 Page 96 of 124

Table 14. Plant 6 measured daily weighted average exposure concentrations. Weighted average concentration, alpha dpm/m3

Occupation 5/56 5/54 10/53 1/53 1/52 1950 1949 1948* 1948**

Digest area lead operator 6 60 36 62 140 84 686

Digest operator 7.3 37 41 52 370 77 77

U-con man #1 7.3

U-con man #2 14

Feinc operator 6.2 96 38 110 175 154 980

Barium operator 38 130 144 126 1.8

Feed operator 40.8 23 100 150 110 126 910

C-3 wash filter operator 79 32 48 120 116 497

C-3 adjustments operator 22 420 120

C-3 centrifuge operator 42 630 52 140 567

Ore Room operator 140 170 370 392 350 13,720

Extraction area lead operator 34 5.4 4

Ether House operator 11 40 46

Ether House lead operator 66 154

Sump recovery operator 8.5 100 76 126 273

Raffinate operator 216 11 8 170 68 154 273

Reduction area (furn room) lead oper'r 22 25 28 69 54 147 686

QM-2 (Orange) packager 268 1,961 120 130 130

LF-9 (Brown) packager 364 38,990 2,730

Nitric acid recovery operator 20 9.6 19 44 35 99 46

Pot Room operator 234 113 45 190 100 336 770 7,770

Metal dissolver #1 204

Metal dissolver #2 21

MGX operator 29 68 52 94

Utility operator 88 129 94 97

Furnace operator 12 33 55 96 150 1,400 24,780

Miller (Mill Room) X X 12,600

Pilot Plant group leader 7.5 6.9 3.1 105 245

Pilot Plant lead operator 7.7 8.8 6.1 77 116 105 245

Pilot Plant technician 1,940 9.2 6 77 116 105 245

Production superintendent 7.7 8.8 56 25

"Experimental continuous furnace"? 8,540 122

Asst. production superintendent 18 21 26

General foreman 14 18 30 50

Foreman 17 21 29 58 52 161

Technical supervisor 18 21 25 33 52 161

Production Office clerk 9.1 12 18 17 27 161

Production Office secretary 3.4 3.4 27 161

Shift foreman 19 25 27 81 96

Cloth & Training Grp Lead Operator 23 25

Cloth operator 18 19 92 245 665

Trainers 231 2,520

Decontamination man 17 22 19 60 99

Decontamination man 3.5 2.7 2.8 29

Receiving clerk 5.2 19 4.5 10 99 28

Cleanup man 22

Production Research Lab personnel 3.7 2 5 13 30 12 84

Ledoux Lab asst technician (raffinate) 15.2 8.1 39

Ledoux Lab technician (raffinate) 12.9 8.1 39 140 420 91 189

Ledoux Lab technician (K-65) 21 7.5 27 440 1,900 1,400 2,100

Ledoux Lab technician (MgF2) 21 7.5 27

Shotgun Lab analyst 24.1 10 27 23 25 24 238

Laboratory personnel 42 2.9 30 23 21

 MCW Laboratory west section 21 245

 MCW Laboratory east section 13 245

Powder sample technician 56.5 217? 3,150

Metal room sampler 420

Outside sampling man 22.5

Sample Room supervisor 41 245 448

Laboratory Office personnel 42 2 5.6

Truck operator 16 19 20 63 75

Truck operator 20 19 20 63 75

Warehouse foreman & Asst Foreman 4.2 2.9 6.2 17 70 161
Effective Date: 10/24/2003 Revision No. 00 Document No. ORAUT-TKBS-0005 Page 97 of 124

Table 14 (Continued)

Weighted Average Concentration, alpha dpm/m3

Occupation 5/56 5/54 10/53 1/53 1/52 1950 1949 1948* 1948**

Warehouse man -- K-65 sampler 350 270 230 84 196

Warehouse man 5.8 10 20 38 46 84 196

Boiler House operator 9.3 7.3 7.5 8.9 2 36 44

Laundry operator 6.2 19 11 19 4.5 13

Porter 3.9 17 14 39

General cleanup 39

Change room 48

Lunch room 5.6

Clothes issue man 18 19 9.4 92

Chief guard 1.7 14 16 14 1.8

Security Office 6.3

Guard 10 13 15 22 1.8 32

Health Office - personnel (office) 1.6 6.7 15 14 0 11 14

Health Office - personnel 8.1 11 15 14 0

Health Office personnel (plant 

monitor/health surveyor) 10 15 15 14 0 46

Health Office person'l (plant monitor) 15 16 15 14 0

Medic 1.3 3.5 6.3

Nurse 3.5 6.3 42 99

Dispensary & Safety 56 175

Instrument Shop technician 12 33 17 40 60 51 252

Instrument Shop machinist 5.5 44 17 27 60 51 252

Maintenance/mechanical supervisor 140 13 10 42 38 50

Maintenance Office clerk 6.5 12 7.7 39

Area mechanic 24 29 28

Ore & Furnace Room AM 2.7 189

Digest & feed AM 1.9 133

Raffinate and C-3 AM 2.3 161

Ether & NA House AM 1.1 77

Welders, pipefitters, etc. 98 126

Carpenters 66

Stock Room (Storeroom) foreman 3.7 14 22 13 33 21

Stock Room clerk 2.6 9 34 15 33 21

AEC Office personnel 2.2 1.8 6.7 0 Non-det 33

AEC Engineer 19 9.9 31 7 

MCW Office personnel 1.5 2 2.9 0.7 0 50

MCW engineer 4.2 4.5 5.4 10 7 15

MCW Office messenger 15 14 15 40

MCW Office maintenance 7.5 12 20 10

MCW Office construction expeditor 9.6 29

Overall average weighted exposure 41 24 25 56 63

Notes

The first set of 1948 data (Column 9) is mostly Mallinckrodt values from MCW 1949a (repeated in AEC 1949) and MCW 1950i, but AEC values from 

AEC 1949 are used where available (they are almost always higher). The second set of 1948 data is from AEC 1951 (it is the daily weighted average 

for packaging UO2 at Plant 4, for comparison). The 1949 data is from MCW 1950i ; the 1950 data is from MCW 1950i and AEC 1953; the 1/52, 1/53, 

and 10/53 sets of survey data are from AEC 1954c; the 5/54 survey data is from AEC 1954d; and the 5/56 survey data is from AEC 1956c.
Effective Date: 10/24/2003 Revision No. 00 Document No. ORAUT-TKBS-0005 Page 98 of 124

Table 15. Plant 6E measured daily weighted average exposure concentrations. 

Weighted average concentration, alpha dpm/m3

Occupation 7/56 3/55 6/54 11/53 4/53 10/52

Lime blender/Slag blender 5.7 11 13 8.9 19 8.9

Jolter 15 79 18 8.9 17 25

Utility operator 50 38 33 32 56 37

Top/Upper "F" machine operator 23 100 46 12 85 25

Bottom/Lower "F" machine operator 29 52 24 13 17 23

Top(ping)-off man 24 113 17 13 17 28

Reduction furnace operator 7.2 20 7.4 6.8 17 14

Breakout operator/man 23 42 28 23 42 25

Residue man 24 300 115 210 80 66

Reduction (KB-2) lead operator 21 36 26 24 36 45

Furnace loaders 23 16 15

Crucible loader 60 19 49 43 130 28

Burnout man 39 31 23 26 26 160

Crucible assembler 42 32 31 58 43 81

Upper/Top furnace operator/man 26 28 30 23 16 21

Bottom furnace operator/man 41 107 118 34 28 68

Saw operator/man 13 425 34 49 17 30

4th or cage saw man 38 21 66 94 17

Cage/cage grinding man 349 35 20 55 166 55

Billet grinder 668 425 34

Brushing man 47

Brushing man/chipper 19 2,110

Recast furnace (YM-5) lead operator 71 30 47 21 19 49

Production machinist 7 11 7.5 10 17 380

Mechanic 73 36 23

Millwright 14 36 23

Maintenance man 27 36 23 53 26 45

Porter 15 23 12 19 8.8 42

Production clerk 9.5 14 7.8 18 13 13

Technical superintendent 4.4 24 15

Technical/Chemical engineer 5.7 24 15 11 16 27

Shift foreman 12 31 23 17 17 44

Foreman/General foreman 6.1 27 19 12 13 20

Lift truck driver 8 22 17 14 15 30

Electrician 35 14 43 34

Decontamination man 15 34 24 21 21

Slag building operator 18 224 110

Average of all personnel 44 113 30 33 43 55
Effective Date: 10/24/2003 Revision No. 00 Document No. ORAUT-TKBS-0005 Page 99 of 124

Table 16. Plant 7 measured daily weighted average exposure concentrations. 

Weighted average concentration, alpha dpm/m3

Occupation 7/56 3/55 6/54 11/53 4-5/53 10/52 9/51

Utility operator/man 17 18 23 77 14 28 9

Area mechanic 12 12 6 11 63 14 37

Welder 14 13 7 10 14 14

Porter 32 25 13 6 14 14

Lift truck operator/driver 41 26 14 6 14 14

HF operator 5.3 5 5 4 14 14 8

TA-7 hoisting operator 17 11 13 11 14 14

Furnace operator 14 16 10 120 14 21 25

Sampler and cleanup man 9.1 8 8 530 42 28

36' Level operator 19 6 5 20 14 14

Panel board operator 16 30 16 7 21 21

TA-7 packager 24 49 68 150 56 28 242

QM-2 dumper/hoister 56 74 48 42 63 112 107

Magnesium Room operator 4.5

Foreman 15 16 7 6 14 21 28

Assistant foreman 8 21 22

Asst plant superintendent 32 70 7

Technical supervisor 18 15 6

Engineer 70 7 13 7 14

Lead operator 30 17 9 12 14 21 38

Clerk/Record Clerk 20 32 8 6 14 21

Decontamination/-ator 18 17 9 10 14 21

Safety inspector 9 12 16 28

Fire marshal 9 12 16

Safety clerk 14 9 17

Average for all personnel 19 19 13 57 22 21

5/53 data includes screw-pulling operation

Table 17. Plant 7E measured daily weighted average exposure 

concentrations. 

Occupation

Weighted average concentration, 

alpha dpm/m3, 3/55

Ionium plant operator 0.06

Ionium plant lead operator 0.1

Overall average 0.07

Note that the ionium plant was the thorium processing plant, in operation from 

1955-1957.
Effective Date: 10/24/2003 Revision No. 00 Document No. ORAUT-TKBS-0005 Page 100 of 124

Table 18. Average and highest airborne dust 

concentrations in the laundry (alpha dpm/m3

) 

(Utnage 1958b, Table 2).

Concentration

Average Worst

Operation

Load washer with coveralls 560 820

Load dryer with coveralls 50 140

Unload dryer with coveralls 20 70

Press white coveralls 40 55

Repair white coveralls 60 90

Sorting to wash 20 870

Weighted average by job

Washer operator 50

Presser 40

Repairman 40

Sorting and handling 30 140

Average general air in laundry 20 110

Notes

"Worst" measurements for sorting and handling were taken 

with no ventilation; "Worst" measurements for the general air 

case were taken with ventilation turned off for 4 hours.

"Average" measurements were taken with ventilation on. The 

worst "average" air in the general area case was found in the 

vicinity of handling and loading into the washer.
Effective Date: 10/24/2003 Revision No. 00 Document No. ORAUT-TKBS-0005 Page 101 of 124

Table 19. Job titles and classifications, with function class codes and geometry factors. 

Plant Job Title or Classification Class Code Notes

Geometry 

Classification, %

 AP ROT ISO

7 36' Level operator D2 Subclassification of 7 Furnace operator (rotating jobs): mainly the UO2-to-UF4 reactor area --- --- ---

6E 4th saw man F Mostly same as Saw man 90 10

6 AEC engineer Z1 Spent time in Plants 4, 6, 6E, and 7 but office was in 6 10 90

6 AEC Office personnel Z2 Spent time only in AEC office in Plant 6 100

4,6,6E,7 Area mechanic P May have worked in all buildings, not dedicated to one 50 50

6E Asst foreman N3 See Shift foreman 50 50

7 Asst plant superintendent N1 May have been in charge of only Plant 7 or of the whole site 10 90

6 Asst production superintendent N2 Assumed to have spent time in Plants 6, 6E, and 7 production areas as well as Plant 6 offices, as 

the production superintendent did

50 50

6 Asst warehouse foreman S2 Worked in the U products warehouse, apparently 25 75

6 Barium operator B4 Worked in the barium salt addition phase of digestion 50 25 25

6E Billet grinder F7 Cleaned and finished billets after recasting and before shipping 90 10

4,6E? Blender E1, G1 See Lime blender 50 50

6 Boiler House operator Q1 Presumably worked to provide steam for the boildown processes 100

4 Bomb charger E1 Loaded MgF2, UO2 into bomb and sealed it; may have blended them 75 25

4 Bomb makeup operator E1, G1 Blended UF4, Mg; charged bomb; dumped out dust collectors 75 25

4 Bomb makeup/deslagging (slag 

grinding) operator

E1, G1, I1 Performed combined functions of Bomb makeup operator and Slag grinding operator 75 25

6E Bottom "F" machine operator E3 Charged bomb (using F machine): UF4-Mg mixture, Mg 75 25

4,6E Bottom furnace operator/man F5 Removed assembly from recast furnace, removed mold, put in new assembly 75 25

6E Breakout operator/man E7 Removed (broke out) derby from bomb 75 25

6 Brown furnace operator/unloader/ 

packager

See UO2 operator/unloader/packager 75 25

6E Brushing man E8 Same as Chipper 75 25

6E Burnout man F5 Removed broken crucible, knocking lid off 80 20

6 C-3 adjustments operator B5 Cleaned wash precipitate filter press; other duties 50 25 25

6 C-3 centrifuge operator B6 Operated and "plowed off" Bird(?) centrifuge (digestion process) 50 25 25

6 C-3 wash filter operator B8 Operated and cleaned the wash precipitate filter press 50 25 25

6E Cage grinding man F7 Mostly same as Saw man 90 10

4,6E Cage operator/man F7 Cleaned and finished billets after recasting; handled scrap in cage 90 10

6E Cage saw man F7 Mostly same as Saw man 90 10

4 Carpenter Q2 May have worked in all buildings, not dedicated to one 25 50 25

4 Casting operator/man F2 Operated the billet casting furnace 75 25

6 C. Eng O1 See Chemical engineer 25 75

4 Ceramic (technician?) L4 Split time: ceramic lab (Ceramic Pilot Plant), Production Research Lab 50 50

4 Charger E1 Same as Bomb charger 75 25

4 Charge firing (man) E2 Furnace operator for bomb furnace 75 25

6E Chemical engineer O1 Worked in production aspects; had production area access 25 75

4 Chemical technician L3 Did chemical analyses in Analytical Lab; possibly some assays 90 10

4 Chemist L3 Did chemical analyses in Analytical Lab; possibly some assays 90 10

4 Chief chemist L3 Did chemical analyses in Analytical Lab; possibly some assays 90 10

6 Chief guard Y1 Spent time in Plants 4, 6, 6E, and 7 but base was in Plant 6 100

4,6E Chipper E8 Deslagged and cleaned derbies (and dingots?) 75 25

4,6 Cleanup man B12, E9 Miscellaneous cleanup activities 50 50

4,6E,7 Clerk X1 Plant 7 clerk worked in Plant 6E as well as inventorying in the Plant 7 production area; Plant 4 clerk 

may have been same person(s)

100

6 Cloth operator/man B3 Cut and replaced filter cloth for Feinc and similar filters 50 25 25

6 Cloth & Training Group lead operator B3 Apparently dual position: coordinated cloth ops, headed Training Grp 10 90

6 Clothes issue man U2 Issued work and protective clothing in the locker rooms 100
Effective Date: 10/24/2003 Revision No. 00 Document No. ORAUT-TKBS-0005 Page 102 of 124

Table 19 (Continued)

Plant Job Title or Classification Class Code Notes

Geometry 

Classification, %

6E Crucible assembler F1 Assembled crucible and mold assembly in billet production process 75 25

6E Crucible loader F3 Loaded the crucible assembly into the recast furnace 75 25

6,6E,7 Decontamination man T Did decontamination in all plants, on boxcars, in Plant 6 Decon Room 50 25 25

7 Decontaminator T Same as Decontamination man 50 25 25

6E Derby chipper E8 Same as Chipper 75 25

4 Derby unloader E6 Unloaded bomb from furnace 75 25

6 Digest area lead operator B1 Worked on the digestion process (up to extraction) 50 25 25

6 Digest operator B1 Worked on the digestion process (up to extraction) 50 25 25

4 Dingot operator G2 Prepared, loaded, removed dingot bomb; probably operated furnace 50 25 25

6 Dispensary (personnel) W Medical-pharmaceutical personnel; no prod area access assumed 100

6E Electrician P2 May have worked in all buildings, not dedicated to one 25 50 25

4,6E,7 Engineer O1 Assumed to be the process engineer assigned to individual plant; may have worked in all; had 

production area access

25 75

6 Ether House operator Q1 Worked in the ether house (operating tanks, valves, etc.) 50 50

6 Experim'rntal Continuous Furnace D Pilot Plant project? May have been pilot for eventual Plant 7 use 25 50 25

6 Extraction area lead operator B7 Worked on the extraction process (up through UO3 production?) 50 25 25

6E Extra man * Presumably a "floating" worker (cf. Utility man) 50 50

6E F (machine) charger E3 Loaded MgF2, UO2 into bomb and sealed it; may have blended them 75 25

6E F machine operator E3 Charged bomb (using F machine): UF4-Mg mixture, Mg 75 25

6 Feed operator B1, B2 Loaded black oxide and other feeds for digestion; washed out feed Niagara; may have handled 

other aspects of the ore-to-UO3 process

50 25 25

6 Feinc operator B2, B5 Operated Feinc filter; cleaned out cake; washed out feed Niagara 50 25 25

7 Filter operator I3 Operated solids-removal(?) filters in Slag Separation Plant (Bldg 701) 50 50

7 Fire Marshall V1 Did inspections in Plants 4, 6, 6E, and 7, but base was in Plant 6 90 10

4,6,6E,7 Foreman N3 Apparently dedicated to individual plant, but may have worked in more 50 50

4 Forge press lead operator G5 Operated the forge press in dingot finishing 75 25

4 Forge press manipulator G5 Same as Forge press operator 75 25

4 Forge press operator G5 Operated the forge press in dingot finishing 75 25

4 Forge press salt bath man G4 Operated the salt bath segment of dingot finishing 75 25

4,6 Fork truck operator/driver Q2 Same as lift truck driver 25 75

4 Furnace and saw man E5, F7 Divided duties: see Furnace operator and Saw man 75 25

4,6E Furnace loader F3 Loaded bomb or crucible assembly into furnace, depending on plant 75 25

4,6E Furnace operator F4 Operated recasting (billet) furnace 75 25

6 Furnace operator C2 Operated the UO3-to-UO2 (Rockwell) furnace 75 25

7 Furnace operator D2, H2, D5 Job (Plant 7) rotated 36' Level, Panel Board, and Sampler & Cleanup tasks: see individual job titles 75 10 15

4 Furnace puller F4 Unloaded UO2-to-UF4 furnace? 75 10 15

4 Furnace tender E5 Tended UO2-to-UF4 furnace 75 25

6E Furnace unloader E6 Unloaded bombs from furnace, cleaned out residue(?) 75 25

6,6E,7 General foreman N3 UF4 production (for Plant 6E); or may have spent time in all plants 50 50

6E Graphite shop personnel R1 Assumed to be doing only clean work manufacturing graphite molds 100

7 Green packager D6 Packaged UF4 (green salt) 75 25

4,6 Guard Y1 Spent time in Plants 4, 6, 6E, and 7 but base was in 6 75 25

6 Health Office -- office personnet V5 Assumed to spend all time in Health Office 100

6 Health Office -- other personnel V2 Spent time in Plants 4, 6, 6E, and 7 but base in 6; assume no production area access 100

6 Health Office -- plant monitor V3 Spent time in Plants 4, 6, 6E, 7 but base in 6; production area access 50 50

6 Health Office -- health surveyor V4 Spent time in Plants 4, 6, 6E, 7 but base in 6; production area access 50 50

4,7 HF operator B6, J1, Q1 Spent some time at Bird centrifuge (7), in recovery areas (7), as well as in providing HF for the 

hydrofluorination process (UO2-to-UF4)(4,7)

10 90

7 Hoisting (slag) operator I1 Worked in the Slag Separation Plant (Bldg 701) 50 50

6 Instrument Shop machinist P3 Worked only in Instrument Shop; some equipment contaminated 10 90
Effective Date: 10/24/2003 Revision No. 00 Document No. ORAUT-TKBS-0005 Page 103 of 124

Table 19 (Continued)

Plant Job Title or Classification Class Code Notes

Geometry 

Classification, %

6 Instrument Shop technician P2 Spent time in Plants 4, 6, 6E and 7 but base in 6 (Instrument Shop); production area access 25 75

7E Ionium plant lead operator K Th-230 processing (mostly operating chemical tanks and filters) 75 25

7E Ionium plant operator K Th-230 processing (mostly operating chemical tanks and filters) 75 25

4,6E Jolter E2 "Jolted" (air hammer?) bomb liner, other material to compress, remove 75 25

6 K-65 sampler S2 See Warehouse man: K-65 sampler 90 10

4,6E KB-2 lead operator E Handled all aspects of derby (KB-2 production) 75 15 10

6 Laboratory Office personnel L3 Worked in the Analytical Lab (Building 102) 100

6 Laboratory personnel L3 Worked in the Analytical Lab (Building 102) 90 10

6 Laundry lead operator U1 Handled operations at the laundry (incl contaminated items) 10 90

6 Laundry operator U1 Laundered work and protective clothing (incl contaminated items) 10 90

4,6,6E,7 Lead Operator * Generic: see specific titles; otherwise, (default) geometry factors at right may be used 75 25

6 Ledoux Lab asst tech (raffinate) L1 Worked in the Ledoux Lab pulverizing, prepping, assaying raffinate 90 10

6 Ledoux Lab technician (raffinate) L1 Worked in the Ledoux Lab pulverizing, prepping, assaying raffinate 90 10

6 Ledoux Lab technician (K-65) L1 Worked in the Ledoux Lab assaying K-65 residue 90 10

6 Ledoux Lab technician (MgF2) L1 Worked in the Ledoux Lab assaying MgF2, incl recycle MgF2 90 10

4 LF-9 (furnace) loader D1 See Furnace loader, Furnace operator for Plant 4 (UO2-to-UF4) 75 25

6 LF-9 unloader/packager See UO2 operator/unloader/packager 75 25

6E,7 Lift truck driver (operator) Q2 Drove lift truck, handling miscellaneous materials, possibly at all plants 25 75

4,6E Lime blender E1 Charge blender for derby production 50 50

4 Magnesium operator Q1 Handled magnesium storage and disbursal; limited exposure potential? 75 25

7 Magnesium Room operator Q1 Handled magnesium storage and disbursal; limited exposure potential? 100

6E Maintenance man P2 May have worked in all buildings; assume production area access 50 50

6 Maintenance Office clerk R3 Worked only in Maintenance Office 100

6 Maintenance supervisor N3 Spent time in Plants 4, 6, 6E and 7 but base was in 6 (same person?) 25 75

6 MCW engineer O1 Spent time in Plants 4, 6, 6E and 7 but base was in 6 10 90

6 MCW Lab (personnel) L3? This lab was not the Ledoux, Shotgun, or Research Labs, which were listed separately in the 

records; perhaps the Analytical Lab

25 75 

6 MCW Office construction expeditor X2 Spent time in Plants 4, 6, 6E, and 7, but probably only in office and construction areas 100

6 MCW Office maintenance X2 Spent time in Plants 4, 6, 6E, and 7, but apparently only in office areas 100

6 MCW Office messenger X2 Spent time in Plants 4, 6, 6E, and 7, but probably only in office areas 100

6 MCW Office personnel X2 Spent time only in MCW offices in Plant 6 (Bldg 112?) 100

4,6,6E Mechanic P1 See Area mechanic 50 50

6 Mechanic supervisor N3 See Maintenance supervisor 50 50

6 Medic W Worked in dispensary presumably, but may have gone into production areas on occasion 100

6 Metal dissolver (#1, #2) I6 Dissolved scrap U metal in acid for recycling as feed? Also spent up to half time in Pot Room 75 25

6 Metal room sampler I7 Sampled scrap and other metal for U or content? 75 25

6 MGX operator B Unclear -- may be Milling & Grinding uranium (operator) since U=X 25 50 25

4 Microscopist L5 Worked full-time in the Microscopy Room, probably on U specimens 100

4, 6 Miller D4 Performed the UF4 pulverizing (4)? Milled UO3 (orange oxide)(6) 50 50

6E Millwright N3 May have worked in all buildings, not dedicated to one 25 50 25

6 Nitric acid recovery operator P2 Worked in the nitric acid recovery area (several buildings) 50 25 25

6 Nurse W Worked in dispensary presumably, but may have gone into production areas on occasion 100

4,6E Office employees X2 Assume they are Plants 4 and 6E production(?) office-only personnel 100

6 Ore Room operator A1, A2, H4 Handled ore, residue: storage, lidding, delidding, cleaning drums 70 20 10

6 Outside sampling man H2, H4 Sampled dust collectors; sampled drums (incl. opening and sealing) 50 50

7 Panel board operator H2 Subclassification of 7 Furnace operator (rotating jobs): vacuuming C-3 into reverter, sampling D-30 

material, replacing D-30 drums, work at panel board area including reverter operation

--- --- ---

6 Pilot Plant engineer M2 Participated in experimental extraction processes and like activities 25 75

6 Pilot Plant group leader M2 Participated in experimental extraction processes and like activities 75 25

6 Pilot Plant lead operator M2 Participated in experimental extraction processes and like activities 75 25
Effective Date: 10/24/2003 Revision No. 00 Document No. ORAUT-TKBS-0005 Page 104 of 124

Table 19 (Continued)

Plant Job Title or Classification Class Code Notes

Geometry 

Classification, %

6 Pilot Plant technician M2 Unclear what activities this Pilot Plant was carrying on 75 25

6 Pipefitter N3 May have worked in all buildings, not dedicated to one 50 50

4 Plant superintendent N1 May have been in charge of only Plant 4 or of the whole site 10 90

4,6,6E,7 Porter R2 Worked in a janitorial capacity, likely in all buildings or rotating among them; entered only office 

areas, not production areas

100

6 Pot Room operator B9 Worked in the denitration (UO2-to-UF4) pot room 50 25 25

6 Powder sample technician L5 Assume mainly did sampling and worked with small samples 75 25

6E Production clerk X1 Kept records and did inventories; assumed to spend some time in production area 100

6E Production machinist R1 Spent all time in graphite shop machining molds and the like 100

6 Production Office clerk X1 Kept records and did inventories; assumed to spend some time in production area 100

6 Production Office secretary X2 Assumed spent all time in office 100

6 Production Research Lab personnel L4 Worked in the Production Research Lab in Building 90 10

6 Production superintendent N2 Spent time in Plants 6, 6E, and 7 production areas and Plant 6 offices 50 50

7 QM-2 dumper C1 Loaded UO3 into trays for the UO3-to-UF4 conversion 75 15 10

7 QM-2 hoister C1 Same as QM-2 dumper 75 15 10

6 QM-2 loader See UO3 loader 75 25

6 QM-2 (orange) packager B11 "Gulped" (vacuum-extracted) UO3 out of the pot and into drums; weighing and making up (weight) in 

drums, emptying dust collectors

75 15 10

6 Raffinate operator B5, B9 Handled the various residues from the filters (drummed, sampled); spent up to half time in Pot Room 75 25

6E Recast furnace lead operator F4 Operated the billet casting furnace 50 50

6 Receiving clerk S1 Recorded and inventoried incoming shipments of ore and U products 50 50

7 Record clerk (A, B) X2 Kept records of receipts, production, and shipments; some entry into production areas and outside 

areas for inventory and like purposes

100

6 Reduction (area) lead operator C2 Operated UO2 (Rockwell) furnace; assumed to load UO3, unload UO2 75 25

6E Reduction (area) lead operator E5 Operated the bomb (UF4-to-U metal) furnace 50 25 25

6E Reduction furnace operator E5 Operated the bomb (UF4-to-U metal) furnace 50 25 25

4,6E Residue man H2, H3 Changed derby chip drums and Hoffman and Mikro dust collectors; cleaned residue from the plate 

and frame press; picked out KB-2

75 25

7 Safety clerk V6 Worked full-time in the Safety Office 100

7 Safety inspector V5 Did inspections in Plants 4, 6, 6E, and 7 (production area access) 10 80 10

4 Salt bath man G4 Same as Forge press salt bath man 75 25

7 Sampler and cleanup man D5 Subclassification of 7 Furnace operator (rotating jobs): sampling UF4, cleanup of furnace platforms --- --- ---

6 Sample Room supervisor A2, A4? Supervisor operations in the Sample Room (probably in Bldg 111) 75 25

4,6E Saw operator/man F7 Removed billet from quench tank; ground, sawed, and weighed it 90 10

6 Security office (personnel) Y2 Assumed to be other than guards (e.g., clerical) 100

6,6E Shift foreman N3 Assumed to be generic, although may have covered several plants; default geometry factors at right 

may be used

50 50

4 Shipping & Receiving (personnel) S1 Handled the receipt and shipment of U products 50 50

6 Shotgun Lab analyst L2 Worked in Shotgun (sample assay) Lab, first in Bldg 55 in Plant 2, then Bldg 108 in Plant 6, then 

Bldg 102(?) of Plant 6

90 10

6E Slag blender I5 Blended C-liner and other slag from Plants 4 and 6 for use in bombs 75 25

6E Slag building operator I3 Ground C-liner and other slag, sorted it via shaker tables, drummed it 75 25

4 Slag man, slag grinding operator E8 Mostly the same as Chipper 75 25

6 Soluble feed operator B1 Loaded soda salt ("diuranate") at appropriate point in digestion process; see also Feed operator 75 25

6 Stockroom clerk S1 Stockroom possibly in Bldg 112. Spent time in Receiving (Bldg 101) 100

6 Stockroom foreman S1 Stockroom possibly in Bldg 112. Spent time in Receiving (Bldg 101) 100

6 Sump recovery operator J2 Worked on U recovery from sump fluids 25 50 25

6E Supervisor O2 Probably same as Technical supervisor 50 50

7 TA-7 hoisting operator (hoister) D1 Hoisted and loaded UO2 into reactor for conversion to UF4 (TA-7) 25 50 25
Effective Date: 10/24/2003 Revision No. 00 Document No. ORAUT-TKBS-0005 Page 105 of 124

Table 19 (Continued)

Plant Job Title or Classification Class Code Notes

Geometry 

Classification, %

4,7 TA-7 packager D6 Packaged UF4 (green salt) 75 25

4 TA-7 Pilot Plant (personnel) D Assumed to be technicians and operators producing UF4 (green salt) 50 50

4 TA-7 unloader (operator) D3 Unloaded UF4 (green salt) from hydrofluorination reactor 50 50

7 Tables operator I3 Worked in the Slag Separation Plant (701) operating shaker tables 25 75

6E Technical engineer O1 Same as Chemical engineer or (Process) Engineer 25 75

6E Technical superintendent O2 May have spent time in all the plants 25 75

4,6,6E,7 Technical supervisor O2 Unclear what duties were; probable frequent production area access 50 50

4 Top cleaner F2 Cleaned top furnace in billet (YM-5) production; not Top seat man 75 25

6E Top "F" machine operator E3 Charged bomb (using F machine): UF4-Mg mixture, Mg 75 25

6E Top furnace operator/man F4, F5 Operated recasting (billet) furnace, removed crucible parts 50 50

6E Top-off operator/man E3 "Topped off" bomb with slag, bolted on lid 75 25

4 Topper E3 Same as Top-off operator? 75 25

4 Top seat man E3 Involved in billet production, furnace area; same as Top-off operator? 75 25

6 Trainer B2, B3 Duties not clear. Assumed to train operators, especially in filter work 10 90

6 Truck operator/driver Q3 Spent time in Plants 6, 6E, and 7 areas; radiation most likely from the back in hauling, the front in 

loading: * 50 AP, 25 PA, 25 ROT

*

6 UO2 (furnace) operator Operated the UO3-to-UO2 (Rockwell) furnace 75 25

6 UO2 unloader/packager Unloaded UO2 from the Rockwell furnace, packaged it 75 25

4 UO3 & brown packer Loaded UO2 into Rockwell furnace; unclear regarding UO2 75 25

6 UO3 loader Loader UO3 onto trays and into the Rockwell furnace (may have been collateral duty of the UO2

furnace operator)

75 25 

6E Upper furnace man F4 Same as top furnace man 50 50

6 U-con man (#1, #2) I3 Handled U-containing slurry from Slag Sep Plant that went to Plant 6 50 25 25

7 Utility operator/man * Worked on various production jobs 50 25 25

6,6E Utility operator * Average of all production jobs for the respective plant 75 25

4 Vacuum fusion chemist L5 Worked in office and vacuum fusion area; did some hands-on U work 90 10

4 Vacuum fusion technician L5 Worked in office and vacuum fusion area; did some hands-on U work 90 10

4 Vertical lathe (operator) G6 Scalped dingot after casting and before forge-pressing 75 25

6 Warehouse foreman S3 Handled storage of ore and U products 25 75

6 Warehouse man: K-65 sampler S2 New job, 1953: sampled K-65 residue, plus typical warehouse duties 90 10

6 Warehouse man (other) S3 Handled storage of ore and U products 25 75

6 Weighmaster A2, S4? Duties not clear. Assumed to perform or approve ore, K-65 weighings 25 75

6,7 Welder N3 May have worked in all buildings, not dedicated to one 10 90

6E YM-5 lead operator F Worked on billet (YM-5) casting 80 20

*Note

Since these classifications are mostly based on later records (after the UO3-to-UF4 direct process was established at Plant 7), UO2 (brown oxide, LF-9) unloading and packaging classifications and UO3 (orange 

oxide, QM-2) milling and loading classifications have been added for Plant 6.

1 
Effective Date: 10/24/2003 Revision No. 00 Document No. ORAUT-TKBS-0005 Page 106 of 124

1 Table 20. Relationship between major processes and job functions. General

code

Operational/functional

area Item code

1 2 3 4 5 6 7 8 9 10 11 12

A Ore handling Thaw, delid, 

clean

Weigh, sample Milling

B Digestion to UO3 Feed loading 

and operations

Feinc/Niagara 

operations

Cloth Barium 

addition

Residue 

removal

Centrifuge Extraction Postextraction 

filter operation

Pot 

Room

UO3

unloading

UO3

packaging

Cleanup

C UO3 to UO2 UO3: loading Furnace ops UO2: unloading UO2:weighing, 

sample, pkg’g

D UO2-UF4, UO3-UF4 UO2 loading Furnace 

(reactor) opers 

UF4 unloading UF4:milling, 

pulverizing

UF4:weigh, 

sample

UF4 

packaging

UO3 loading 

(UO3-UF4) 

E UF4 to Derbies UF4+Mg mixing Charge jolting Bomb:chrg, 

top, seal

Bomb loading Furnace ops Bomb 

unloading

Derby 

removal

Chipping, grind, 

pickling

Cleanup 

F Derbies to Billets Crucible 

assembly

Furnace 

preparation

Crucible 

loading

Furnace ops Crucible 

unloading

Billet 

removal

Billet: crop, 

grind, saw

Billet: package, 

weigh

G UF4 to Dingots UF4+Mg mixing Mold handling, 

furnace oper

Chipping Salt bathing Forging Scalping 

H Residue/Waste handling Cake ops (exc 

removall)

Dust: collect, 

sample

Slag, chip 

handling

Store, sample

I U recovery operations Slag loading Slag grinding Slag 

processing

Slag filter 

operations 

Slag blending Metal 

dissolving

Metal Room 

sampling

J Other recovery ops Nitr ox recovery Sump recovery

K Th processing All phases

L Laboratory ops Ledoux Shotgun Analytical Prod Resrch Other

M Process/metallurgical devel Pilot Plant 4 Pilot Plant 6

N Process mgt & supervision Infrequent PAA Frequent PAA Routine PAA

O Tech. supervision & supp Engineer Tech. supervisor

P Process, area maintenance Mechanic Other PAA craft, 

technicians

Contaminated 

equip shop

 

Q Process support Supply facilities Fork/Lift driver Truck driver PAA constr.

R Nonprocess maintenance, 

support

Graphite Shop Clean maint/

support

Office only 

S Ship’g & Rec'g/Warehouse Ship & Rec Warehs - –K-65 WH - Other Weighing

T Decontamination

U Laundry/Clothes handling Laundry Clothes issuing

V Safety & Health Fire marshal Health-other Plant monitor Health, med Safety insp Office only

W Medical

X Office personnel PAA Clerk Office only

Y Security Guard/chief Office only Note: PAA stands for production area access, indicating potential for exposure. Z AEC Engineer Office only

* Generic -- ID process
Effective Date: 10/24/2003 Revision No. 00 Document No. ORAUT-TKBS-0005 Page 107 of 124

Table 21. Uranium dust daily weighted average exposure levels, Plant 4. Airborne dust exposures, alpha dpm/m3

Job title

Oct 

1942 - 

1946 1947

1948 - 

1949 1950 1951

1952 - 

1953

1954 - 

1955 1956 1957

Mechanic/Area mechanic 350 350 350 84 15 15 22 22 X

Blender/Bomb charger/ Charger/Bomb makeup/ Dingot 

operator/Slag grinding operator

3,010 3,010 3,010 210 64 64 33 85 X

Cage operator/man 190 190 190 190 X X X X X

Carpenter/Other craft 140 140 140 84 15 15 22 22 X

Casting operator/Furnace operator/Top seat man 5,100 910 980 140 480 480 110 11 X

Lab: ceramics/microscopy X? X? X? X? 18 18 18 18 X

Charge firing (man) 931 910 980 140 X X X X X

Chemist/Chief chemist/Chemical technician 40 40 40 40 7 7 10 10 X

Chipper/Cleanup man/Saw operator/man 1,890 1,890 910 140 140 140 140 140 140

Derby unloader 350 280 1,260 175 X X X X X

Foreman/Shift foreman/Engineer/Technical supervisor 175 175 175 56 12 12 10 23 X

Forge press operator/lead operator/manipulator* X X X X X 23 23 23 X 

Furnace and saw man X X X X 18 18 18 18 X

Furnace loader (UF4-derby) 3,360 3,360 2,240 280 X X X X X

Furnace tender 560 560 350 70 X X X X X 

Guard/Chief guard 28 28 28 28 28 5.8 2.7 7.1 X

HF (fluorination) operator 570 570 570 70 X X X X X

Jolter 3,500 3,500 490 70 X X X X X

KB-2/YM-5/Dingot lead operator/Furnace puller 931 630 560 35 19 19 8.2 19 X

UO2/LF-9 loader/packer 4,200 3,360 2,240 280 X X X X X

Lime blender/Magnesium operator 70 70 70 35 X X X X X

Miller-mixer (UF4/TA-7)/Top cleaner 4,690 4,690 980 70 X X X X X

Office: Plant superintendent/Clerk/Other 42 42 42 42 42 5.8 2.7 7.3 X

Porter 112 112 112 56 56 5.8 2.7 40 X

Residue man/Salt bath man/Vertical lathe operator X X X X X X 28 28 X 

Shipping & Receiving 126 126 126 126 15 15 22 22 X

Slag man/Slag grinding operator 140 140 210 70 X X X X X

TA-7 packager 7,210 7,210 7,210 245 X X X X X

TA-7 unloader (operator) 13,000 13,000 1,540 210 X X X X X

Topper 840 840 2,310 210 X X X X X

Vacuum fusion*: chemist/technician X? X? X? X? X? 59 59 59 X

Notes

An “X” indicates that the job title did not exist during the indicated period. An “X?” indicates that it is uncertain if the job title existed, i.e., it is not 

certain when that job began.

* Work started in 1953
Effective Date: 10/24/2003 Revision No. 00 Document No. ORAUT-TKBS-0005 Page 108 of 124

Table 22. Uranium dust daily weighted average exposure levels, Plant 6. Airborne dust exposures, alpha dpm/m3

Job title

1946 -

1948 1949 1950 1951 1952 1953 1954

1955 -

1957

AEC engineer X? X? X? 7 31 9.9 19 19

Barium operator 1.8 1.8 126 144 130 38 X X

C-3 centrifuge/wash filter/adjustments operator 567 567 140 140 420 630 79 79

Cleanup man/utility operator X X X? 97 97 94 129 88

Cloth operator 665 665 245 245 92 19 18 18

Cloth & Training Group lead operator/trainer 2,520 2,520 231 231 23 25 18 18

Clothes issue man 92 92 92 92 92 9.4 19 18

Crafts: Carpenter/Pipefitter/Welder 126 126 98 28 28 28 29 24

Decontamination man/U-con man* 99 99 99 99 60 19 22 17

Cloth/Digest/Reduction operator, Outside sampling* 686 686 245 370 92 41 60 22

Dispensary: Nurse/Medic/Other (personnel) 175 175 56 99 42 6.3 3.5 1.3

Boiler/Ether House/Extraction /Nitric acid recovery 

operator

46 46 99 50 44 19 11 34

"Experimental Continuous Furnace": Pilot Plant 

project**

X? 8,540 8,540 X? X X X X

Feinc/Feed/Soluble feed operator 980 980 154 175 150 100 96 41

Foreman/General foreman/Shift foreman/Technical 

supervisor

161 161 161 96 81 30 25 19

Furnace operator 24,780 24,780 1,400 150 96 55 33 12

Guard/Chief guard 32 32 32 1.8 22 16 14 10

Health/Security Office personnel; Engineer (MCW, 

chemical)

15 15 15 7 14 15 11 8.1

Health Office: health surveyor/plant monitor 46 46 46 42 14 15 16 15

Instrument Shop machinist/technician 252 252 51 60 40 17 44 12

Laboratory Office personnel 100 100 10 5.6 5.6 5.6 2 42

Laboratory personnel - generic/MCW/Shotgun 245 245 24 25 23 30 10 42

Laundry operator/lead operator X? X? X? 4.5 19 11 19 6.2

Ledoux Lab technician/assisstant technician -

raffinate, MgF2

189 189 91 420 140 39 8.1 27

Ledoux Lab technician (K-65) 2,100 2,100 1,400 1,900 440 27 7.5 21

LF-9/brown/UO2 packager/unloader 38,990 38,990 364 350 350 X X X

Maintenance supervisor 50 50 50 38 42 10 13 140

Mechanic/Area mechanic: ore & furnace room, digest 

& feed, raffinate & C-3, Ether House, Nitric Acid 

House

189 189 2.7 28 28 28 29 24

Metal dissolver (#1, #2) X? X? X? 204 204 204 204 204

Metal room sampler X? X? X? 420 420 420 420 420

MgX operator 94 94 94 94 52 68 29 29

Miller (UO3/QM-2)*** 12,600 12,600 X X X X X X

Office: MCW - Clerk/Maintenance/Messenger/Porter/ 

Expeditor

50 50 50 48 48 20 17 15

Office: MCW - Other, AEC - all AEC except Engineer 50 50 50 0 6.7 2.9 2.2 2.2

Office: Production - Clerk/Secretary, Receiving - Clerk 161 161 52 99 17 18 19 9.1

Ore Room operator*** 13,720 350 392 370 170 140 140 X

Pilot Plant engineer 123 123 53 58 39 3.1 6.9 7.5

Pilot Plant lead operator/group leader 245 245 105 116 77 6.1 8.8 7.7

Pilot Plant technician 245 245 105 116 77 6 9.2 1,940

Pot Room operator 7770 770 336 100 190 45 113 234

Powder sample technician 3150 3150 217 217 57 57 57 57

Prod Research Lab personnel 84 84 12 30 13 5 2 3.7

Production superintendent/Asst production 

superintendent

25 25 25 25 26 56 21 18

QM-2 (orange) loader 5,320 5,320 1,400 420 420 420 X X

QM-2 (orange) packager 1,400 1,400 1,400 420 130 130 120 *****

Raffinate/Sump recovery operator 273 273 154 76 170 8.5 11 216

Sample Room supervisor 448 448 245 245 245 41 41 41

Stockroom foreman/clerk 21 21 21 33 15 34 14 3.7

Truck/forktruck operator/driver 75 75 75 75 63 20 19 20

Warehouse foreman/Assistant 

foreman/Warehouseman

196 196 84 70 38 20 10 5.8

Warehouse K-65 sampler/Weighmaster**** 196 196 84 230 270 350 350 X?

Notes

An “X” indicates that the job title did not exist during the indicated period. An “X?” indicates that it is uncertain if the job title existed, i.e., it is 

not certain when that job began.
Effective Date: 10/24/2003 Revision No. 00 Document No. ORAUT-TKBS-0005 Page 109 of 124

Table 22 (Continued)

* Outside sampling and U-con positions began in 1955.

** Pilot Plant work began in 1948.

*** Milling work ended by June 1949.

  These jobs appear to have ended in 1954, or to have been subsumed in other job titles.

** *** 1961 dpm/m3 in 1955, 268 in 1956-57.

Table 23. Uranium dust daily weighted average exposure levels, Plant 6E. Airborne dust exposures, alpha dpm/m3

Job Title

Oct 1950 -

Dec 1952 1953 1954 1955 1956 -1957

Area mechanic 23 23 23 36 73

Billet grinder 34 34 34 425 668

Blender; Lime/slag blender 8.9 14 13 11 5.7

Bottom/Lower "F" machine operator 23 15 24 52 29

Breakout operator/man 25 33 28 42 23

Brushing man X X X 47 47

Burnout man 160 26 23 31 39

Cage grinding/cage operator/man 55 111 20 35 349

Cage/4th saw man 17 80 21 38 38

Capping man/Crucible assembler 81 51 31 32 42

Chipper/Derby chipper 2,110 2110 2,110 2,110 19

Crafts: Maintenance/Electrician/Mechanic/Millwright; Graphite Shop 45 40 23 36 73

Crucible loader 28 87 49 19 60

Engineer/Chemical engineer/Technical engineer; Superintendent/Supervisor 27 14 15 24 5.7

F (machine) charger/Extra man/Utility operator 37 44 33 38 50

Foreman/Assistant foreman/General foreman 20 13 19 27 6.1

Furnace loader; Reduction furnace operator 15 20 23 23 23

Furnace operator/unloader; Bottom furnace operator; Generic lead operator 68 31 118 107 41

KB-2/Reduction lead operator; Shift foreman; Porter 45 40 26 36 27

Lift truck driver (operator) 30 15 17 22 8 

Office employees/Clerk; Production clerk 13 16 7.8 14 9.5

Production machinist 380 14 7.5 11 7

Recast furnaceYM-5 lead operator 49 20 47 30 71

Residue man 66 145 115 300 24

Saw operator/man 30 33 34 425 13

Slag building operator 110 110 110 224 18

Top/Upper/Generic "F" machine operator; Top(-ping) operator; Jolter 28 49 46 113 24

Top/upper furnace operator; Decontamination man 21 21 30 34 26

Notes

An “X” indicates that the job title did not exist during the indicated period. An “X?” indicates that it is uncertain if the job title existed, i.e., it is not 

certain when that job began
Effective Date: 10/24/2003 Revision No. 00 Document No. ORAUT-TKBS-0005 Page 110 of 124

Table 24. Uranium dust daily weighted average exposure levels, Plant 7 (including the Slag 

Separation Plant) and Plant 7E. Airborne dust exposures, alpha dpm/m3

Job title 1951 1952 1953 1954 1955

1956 -

1957

Plant 7

Area mechanic; Welder 37 14 37 7 13 14

Asst foreman/Plant superintendent/Engineer 22 21 10 7 70 32

Clerk/Record clerk; Porter 22 21 10 13 32 32

Decontamination man/Decontaminator 161 21 12 9 17 18

Filter/Tables operator* X X X X 9 9 

Foreman/Technical supervisor; Safety inspector/Fire marshal 28 28 22 12 16 18

Furnace operator/Utility operator 25 28 67 23 18 17

HF/Magnesium Room operator 8 14 9 5 5 5.3

Hoisting (slag) operator* x X X X 15 15

Lead (UO3-to-UF4, TA-7) Operator; 36' Level/Panel board operator** 38 21 17 9 30 30

Lift truck driver (operator) 14 14 10 14 26 41

QM-2 dumper/hoister 107 56 61 42 63 112

Safety clerk 17 17 17 9 14 14

Sampler and cleanup man X 28 286 8 8 9.1

TA-7 hoisting operator (hoister) 121 14 13 13 11 17

TA-7 (green/UF4) packager 242 28 103 68 49 24

Plant 7E (thorium/ionium) process)

Ionium plant operator/lead operator*** X X X X 0.1 0.3

Notes

An “X” indicates that the job title did not exist during the indicated period. An “X?” indicates that it is uncertain if the job title existed, i.e., it is not 

certain when that job began.

Plant 7E was in its startup phase when the only known measurements were taken in March 1955. Thus as AEC (1955c) noted, the full processing 

figures would likely be higher. Thus figures for 1956 and 1957 have been tripled to allow for full processing. Processing ceased at some point in late 

1956 or very early 1957, so a end date of March 1957 should be taken. 

* These positions began after July 1955.

**The 36' Level operator position began in 1953, the panel board operator position in 1952.

***The ionium plant work began after July 1955 and continued until March 1957.
Effective Date: 10/24/2003 Revision No. 00 Document No. ORAUT-TKBS-0005 Page 111 of 124

Table 25. Measured radon concentrations at various indoor and outdoor areas, in units of 1 x 10-10

Ci/L. 

 

1947

Med/

Mean

1948 1949 1951 1952

No. Min

Med/

Mean Max GSD No. Min

Med/

Mean Max No.

Med/

Mean GSD No.

Med/

Mean

INDOOR AREAS

Scalehouse/Ore 

Storage/Warehouse (1)

0.10 27 0.46 3.28 35 0.09 5.91 95 0.09

Scalehouse/Ore 

Storage/Warehouse (2)

193 0.03 2.02 32.8 158 0.03 1.03 13

Digest/Feed (1) 2 0.05 10.6

Digest/Feed (2) 57 0.03 0.39 7.82

Extraction Cells (1) 33 0.36 5.89 92 0.37

Extraction Cells (2) 14 0.03 1.08 4.14

Centrifuge Area (1) 35 0.06 4.77 95 0.12

Centrifuge Area (2) 141 0.03 1.13 11.7 48 0.03 1.55 12

Feinc/Filter/C-3/Raffinate/

Cloth Storage (1)

9 0.18 1.98 34 0.18 3.65 94 0.14

Feinc/Filter/C-3/Raffinate/

Cloth Storage (2)

186 0.03 12.3 894 172 0.03 0.75 468

Orange Packing

Pot Room

Ledoux Lab (!)

Ledoux Lab (2) 30 0.03 0.08 0.37

Shotgun Lab

Other Lab (Research/Control/ Xray/MY)

6E Breakout Area

6E Recast Area

Recovery area 0.21

Decontamination Room

Metal Dissolver Bldg

NA House

Ether House

Refrigeration Room

Receiving

Welding Shop

Millwright Shop

Electric shop

Maintenance Shop 0.09

Smoking Room

Production Office Area

Dispensary 0.11

YARDS AND OTHER 

OUTDOOR AREAS

General Plant 6 (1)

General Plant 6 (2)

35 0.03 0.19 2.06

Bldg 104/Scalehouse 0.07 9 0.07 1.47

Bldg 104 (Pilot Plant) (1)

Bldg 104 (Pilot Plant) (2)

7 0.03 0.15 0.6

Warehouse 17 0.10 2.05

Warehouse exhaust (1)

Warehouse exhaust (2)

33 0.06 0.91 3.3

Near warehouse during K-65 

drum welding

8 0.11 2.84

Labs, guard office 8 0.06 1.38

Ether House/ Bldg 109

Parking lot area 9 0.13 1.56

Scalehouse intake/ exhaust 3 0.12 2.06

Scalehouse exhaust (1) 18 0.13 3.08

Scalehouse exhaust (2) 24 0.04 2.2 48.7

Ore Storage intake/ exhaust 6 0.68 2.44

Ore Storage exhaust 2 0.31 1.77

Outdoor drum storage, Bldg 115/ 

Plant 6E
Effective Date: 10/24/2003 Revision No. 00 Document No. ORAUT-TKBS-0005 Page 112 of 124

Table 25. (Continued)

1952 1953 1954 1956 1957

GSD No.

Med/

Mean GSD No. Min

Med/

Mean Max GSD No.

Med/

Mean GSD No.

Med/

Mean GSD

INDOOR AREAS

Scalehouse/Ore 

Storage/Warehouse (1)

6.04 66 0.01 4.99 74 0.01 5.45 5 0.01 1.00

Scalehouse/Ore 

Storage/Warehouse (2)

Digest/Feed (1) 17 0.12 5.20 57 0.03 3.85 18 0.01 5.06 3 0.03 5.46

Digest/Feed (2)

Extraction Cells (1) 5.31 59 0.28 3.79 101 0.26 5.82 33 0.01 7.00 3 0.01 8.65

Extraction Cells (2)

Centrifuge Area (1) 4.96 55 0.05 5.59 51 0.07 6.46 7 0.01 5.18 3 0.01 1.00

Centrifuge Area (2)

Feinc/Filter/C-3/Raffinate/

Cloth Storage (1)

4.10 60 0.10 4.80 114 0.14 6.22 41 0.07 5.76 5 0.01 6.82

Feinc/Filter/C-3/Raffinate/

Cloth Storage (2)

Orange Packing 1 0.12 0.12 0.12

Pot Room 2 0.02 2.17

Ledoux Lab (!) 6 0.10 1.33 10 0.02 3.38

Ledoux Lab (2)

Shotgun Lab 2 0.04 5.80

Other Lab (Research/

Control/ X-ray/MY)

3 0.01 1.00 2 0.04 8.02

6E Breakout Area 3 0.01 1.00

6E Recast Area 4 0.01 1.00

Recovery area 1 0.11 0.11 0.11

Decontamination Room 1 0.01 0.01 0.01

Metal Dissolver Bldg 4 0.01 2.65

NA House 3 0.01 16.0

Ether House 2 0.02 2.67

Refrigeration Room 1 0.01 0.01 0.01

Receiving 1 0.01 0.01 0.01

Welding Shop 1 0.01 0.01 0.01

Millwright Shop 1 0.01 0.01 0.01

Electric shop 3 0.01 1.00

Maintenance Shop 3 0.04 6.95

Smoking Room 1 0.01 0.01 0.01

Production Office Area 1 0.01 0.01 0.01

Dispensary 2 0.04 6.13

YARDS AND OTHER 

OUTDOOR AREAS

General Plant 6 (1)

General Plant 6 (2)

Bldg 104/Scalehouse

Bldg 104 (Pilot Plant) (1)

Bldg 104 (Pilot Plant) (2)

Warehouse

Warehouse exhaust (1)

Warehouse exhaust (2)

Near warehouse during K-65 

drum welding

Labs, guard office

Ether House/ Bldg 109 1 0.01 0.01 0

Parking lot area

Scalehouse intake/ exhaust

Scalehouse exhaust (1)

Scalehouse exhaust (2)

Ore Storage intake/ exhaust

Ore Storage exhaust

Outdoor drum storage, Bldg 

115/ Plant 6E

3 0.30 8.62

Notes

Values are taken from (1) AEC 1948b, MCW various, and MCW 1951b and (2) AEC 1949.

Where there was only a single measurement for the time period, the value is given above in the Median/Mean column. Where there were only a minimum, 

average, and maximum reported, the three values are given above in the Min, Median/Mean, and Max columns. Where there were two or more measurements 

and the data was given in full, a lognormal distribution was used for the analysis and the geometric mean and geometric standard deviation (GSD) are given 
Effective Date: 10/24/2003 Revision No. 00 Document No. ORAUT-TKBS-0005 Page 113 of 124

above in the Median/Mean and GSD columns. In each case where the number of measurements was given in the reference, the number is given above in the 

No. column

Table 26. Middlesex ore storage worker 

radon exposures, January 1944–January 

1949 (AEC 1949, Table 1).

Level

Area Min Avg Max

Railcar unloading 200 250

Drying room 1 6.7 21.7

Storage area 0.1 0.8 2.9

Sampler rooms 0.1 0.2 0.5

Crushing area 0.1 0.2 0.9

Crusher pit 0.3 0.8 1.3

Sampling lab <0.1 <0.1 <0.1

Weigher's booth <0.1 <0.1 <0.1

Notes

Radon levels are given in multiples of the radon 

MPC/preferred level of 10-10 Ci/L. New blowerequipped railcars began to be used in about early 

1949, which reduced levels to below the MPC.

Table 27. Potential radon exposures from thorium processing. 

Isotope Maximum Ore Content, Ci Emanation per Working Hour, Ci Average Concentration, Ci/L

Rn-220 3.29 × 10-3 7.31 × 10-7 1.62 × 10-12

Rn-222 <1.30 × 10-12 2.89 × 10-16 6.42 × 10-22

Notes

The maximum ore content is from Table 6.

Emanation is based on an assumption of 10% outgassed to the room (this is conservative, since most would have been 

vented when the cake containers were opened in a well-ventilated area, as for ore, or when the cake was in the digestion 

vessels). The total number of working hours is 2000 per year for 2.25 years (1955-March 1957). Thus the emanation per 

hour is 10% of the maximum ore content, divided by 4500 hours. 

The average concentration is based on a conservative assumption of 2 air changes per hour (i.e., minimal ventilation) in a 

room measuring 3 m × 3 m × 2.5 m (about 10 ft × 10 ft × 7.5 ft), a conservatively small process area. This gives an air 

volume of 22,500 liters in the room and an air change volume of 45,000 liters. The emanation per working hour is divided by 

the air change volume to produce the average concentration.

Table 28. Ratio of alpha airborne activity to surface contamination (Becher 

1958, Table 2).

Source of Data Ratio*

Two operating areas having the highest contamination levels (1953)

 Shift-length air samples 0.64

 Spot air samples 1.9

Plant-wide operations (415 surveys over 9 months in 1958)

 Shift-length air samples 0.36

 Spot air samples 5.05

Special test conditions (1953)

 Simulated conditions 13

 Worst possible conditions (short periods) 20

* Units of the ratio are dpm/m3 of air per dpm/cm2 of surface
Effective Date: 10/24/2003 Revision No. 00 Document No. ORAUT-TKBS-0005 Page 114 of 124

Table 29. Contamination levels on various surfaces and 

pieces of equipment in the laundry (Utnage 1958b, Table 1).

Unit Beta + Gamma, cpm

Presser pad surface 2,000

Cloth hamper surface 5,000

Wooden hamper surface 1,000

Work tables and surfaces 200-1000

Lint ball/dryer lint trap/roof lint trap 4,000/2,000/3,000

Inside washer 300

Inside dryer 1,000-4,000

Under washer 20,000-60,000

Under dryer 100-40,000

Floors, average 300-500

Walls 100-300

Overhead pipes, etc. 300-1000

Table 30. Contamination levels and associated dose rates from work clothing (Utnage 1958b).

Spot Whole Garment or Group

Item Max Average Max Average Contamination Notes

Regulated 

coveralls

80% have ≥ 1-2 

in2 spot >3000 

cpm

Most coveralls 

have ≥1 spot 

>1 mr/hr

60% have ≥ 1 

mrep/hr over 

whole garment

1.5 mrep/hr; range, 0.2 to 12; 

100% have avg >100 cpm 

per 2 in2

Apparent U spots on 70%; 

area is from 10 cm2 to 

30% of total area 

Nonregulated 

coveralls

30% have ≥1 spot 

>1000 cpm

100% <3 mrep/hr 80% <1 mrep/hr; 0% with avg 

<100 cpm

5% visibly contaminated

Handkerchiefs >1,000 cpm <300 cpm 10% visibly contaminated

Socks 800 cpm 200 cpm

Underwear 800 cpm 200 cpm

Caps 300 cpm Low to moderate

Blue smocks >1,000 cpm 200 cpm Low to moderate

White smocks >1,000 cpm More than blue smocks 

(more spots)

Lab smocks >1,000 cpm <300 cpm Low (few spots)

Gloves ~5,000 cpm per 2 in2 All: heavy

Shoe covers >90% have >1,000 cpm "Destrehan": "high"

Note: Measurements were taken after wearing but before washing.

1 
Effective Date: 10/24/2003 Revision No. 00 Document No. ORAUT-TKBS-0005 Page 115 of 124

1 Table 31. Surrogate (comparable) worker inhalation intake values, in pCi/yr. 2 

Category Type Median or *Actual GSD

No. 

Cases Median or *Actual GSD

No. 

Cases Median or *Actual GSD

No. 

Cases

Plant 6 Period 123 Period 1 Period 23

Generic (mixed, miscellaneous, or unknown) 2.51 x 104 0.482 21 8.29 x 104 0.626 13 2.12 x 104 0.574 7

Mostly Cloth, Raffinate, Feinc -- -- -- 5.61 x 104 0.997 3 2.87 x 104 0.146 3

 Mostly Digest 2.10 x 104 0.529 3 4.45 x 104 0.933 3 *1.98 x 104 -- 1 

Mostly Ether House 2.19 x 104 0.493 5 3.08 x 104 0.200 4 2.32 x 104 0.0016 3

Mostly Packaging -- -- -- 1.28 x 105 1.636 3 -- -- --

Mostly Pot Room, Orange Oxide -- -- -- 1.80 x 105 0.697 5 2.26 x 104 0.843 5

 Period 12 Period 1 Period 2

 Mostly Ore Room

*2.45 x 105-4.96 x

104 -- 2 

*4.83 x 104 - 5.96 x

104 -- 2 *3.84 x 104 -- 1 

Plant 6E Period 123 Period 1 Period 23

 Generic (mixed, miscellaneous, or unknown) 2.98 x 104 0.450 12 4.79 x 104 0.831 4 3.12 x 104 0.439 13

Bomb Step (UF4→Derby) 3.19 x 104 0.335 3 2.30 x 104 0.410 4 2.59 x 104 0.665 6

 Recast (Derby→Ingot)

*2.49 x 104 - 2.02 x 

104 -- 2 4.71 x 104 0.575 3 3.44 x 104 0.507 4 

Period 12

Graphite Shop 1.56 x 104 0.450 3

Plant 4 Period 123 Period 1 Period 23

4 Metal (Derby, Ingot) -- -- -- 1.24 x 105 0.931 11 -- -- --

4 Pilot Plant (Dingot, Metallurgical, etc.)

*1.65 x 104 - 1.04 x 

104 -- 2 2.29 x 104 0.874 3 2.92 x 104 0.484 8

Plants 4 & 7 Period 123 Period 1 

(Green Salt) UF4 Production Work, First at 4, Then at 7 1.81 x 104 0.449 8 1.18 x 105 1.499 7

Plant 7 Period 23 Period 2 

Generic (mixed, miscellaneous, or unknown) 1.67 x 104 0.408 16 2.18 x 104 0.624 7

Multi-Plant Period 123 Period 1 Period 23

 AEC

*7.20 x 103 - 6.94 x

103 -- 2 1.96 x 104 0.465 5 

*6.01 x 103 - 1.60 x

104 -- 2 

 Decon/Cleanup 1.44 x 104 0.209 3 -- -- -- -- --

Boiler House/Power House 1.11 x 104 0.203 6 1.73 x 104 0.439 6 *1.23 x 104 -- 1

Engineering 1.08 x 104 0.336 7 3.38 x 104 0.436 10 9.71 x 103 0.222 10

 Instrument Shop 1.26 x 104 0.263 6 2.13 x 104 0.647 7 -- -- --

Laboratories (Anal., Research, Ledoux, etc.) 1.11 x 104 0.406 24 2.30 x 104 0.742 22 1.17 x 104 0.352 23

 Laundry 1.10 x 104 0.360 6 1.73 x 104 1.036 8 

*9.06 x 103 - I72.20 x 

104 -- 2 

Maintenance
Effective Date: 10/24/2003 Revision No. 00 Document No. ORAUT-TKBS-0005 Page 116 of 124

Carpenter *9.17 x 103 -- 1 -- -- -- -- -- --

 Electrical 1.66 x 104 0.438 7 5.19 x 104 0.487 7 -- -- --

Insulator/Pipe Coverer *3.25 x 104 -- 1 -- -- -- -- -- --

Machinist/Machine Shop *4.17 x 104 -- 1 -- -- -- -- -- --

Mechanic/Millwright 1.75 x 104 0.544 10 7.05 x 104 0.580 9 2.40 x 104 0.507 4

Miscellaneous/Mixed 1.46 x 104 0.372 12 3.57 x 104 0.799 12 -- -- --

Multi-Plant, Oiler *1.76 x 104 -- 1 -- -- -- -- -- --

 continued Painter 1.49 x 104 0.349 5 1.70 x 104 0.446 6 -- -- --

 Pipefitter 2.42 x 104 0.693 7 2.65 x 104 0.845 9 -- -- --

 Rigger

*3.24 x 104 - 1.48 x 

104 -- 2 -- -- -- -- -- --

 Tinner *2.49 x 104 -- 1 *3.23 x 104 -- 1 -- -- --

 Welder 2.62 x 104 0.433 7 3.30 x 104 0.298 6 

*6.63 x 103 - 9.43 x 

103 -- 2 

Office Workers 9.74 x 103 0.310 23 1.98 x 104 0.507 15 1.10 x 104 0.592 6

Porter/Custodian 1.12 x 104 0.260 7 2.94 x 104 0.895 9 8.55 x 103 0.213 3

Safety & Health, Fire 8.23 x 103 0.314 3 2.06 x 104 0.669 6 1.35 x 104 0.625 5

Shipping & Receiving, Warehouse, 

Storeroom,

 Fork Truck Operator 1.11 x 104 0.380

22

1.52 x 104 0.843

20

9.63 x 103 0.180

3

 

Supervision & Management (where not

 specified by plant) 1.95 x 104 0.394

7

3.46 x 104 1.059

7 -- -- --

Notes

A lognormal distribution was used. The geometric median and the geometric standard deviation (GSD) are shown. However, where only one or two suitable cases were found for a category 

and time period, no distribution could be formulated. For these cells, only the one or two actual results from IMBA are given; this is indicated by an asterisk (*). The non-distribution case 

results thus indicate what may be known regarding the exposure of a comparable worker.

The sample size was assumed to be 1.4 liters; the assumed activity equivalence is .676 pCi per μg nat U (or 676 pCi/mg); the sample data was in mg U/l; and IMBA uses 365 days of 

exposure per year. Hence the conversion factor for input data in mg U/l was 676 x 1.4 = 946, to give pCi; the conversion factor for IMBA output data in pCi/day was 365, to give pCi/year; and 

the overall conversion factor was thus 517.90.

Period 1 is from 1948-1951; Period 2 from 1952-1955; and Period 3 from 1956-1958. Period 123 includes all three periods, Period 23 includes only the last two periods, and so forth. Major 

improvements were made in 1949-1951. Thus after this period, exposures went down in many areas and for some occupations. Increases in exposures in other areas and for other 

occupations are likely due to increases in production.

In IMBA, the assumption was made that all of the uranium measured was U-234. The daughters assumed to be in equilibrium in the U-238 and U-235 chains must be added by the dose 

reconstructor when appropriate (see Section 6.1 of the text).
Effective Date: 10/24/2003 Revision No. 00 Document No. ORAUT-TKBS-0005 Page 117 of 124

Table 32. Estimate of annual intake of thorium dust during Plant 7E 

operation. 

Jul-Dec 1955 1956 Jan-Mar 1957

Alpha dpm/m3 0.1 0.3 0.3

pCi/m3 0.0450 0.135 0.135

pCi inhaled in the period 63.1 378 94.6

Notes

All of the source is assumed to be Th-230 (with negligible build-in of Ra-226, etc.)

The breathing rate is taken to be 1.4 m3/hr and the working time to be 2000 hours 

per year, ratioed as necessary to correspond to the fraction of a year.

Table 33. Exposure rates (mR/hr) from drums of K-65 and Q-11 residue in a railcar (AEC 1949, 

Figure 21).

Source Position Distance, feet

Dose 

rate

Dose 

rate A

Dose 

rate B

Dose 

rate C

96 drums of K-65, along 

the full width of a railcar 

and ~46 ft along its length 

(57 ft)

Drum group centerline 

perpendicular to long side of 

railcar, 4 ft from ground

1 85

3 72.5

5 50

8 37.5

12 27

18 19

24 12.5

30 9.4

Top of railcar Contact 104

87 drums of Q-11, 

distributed in two groups 

at ends of boxcar

One drum group centerline 

perpendicular to long side of 

boxcar (A); at center door (B); at 

end (C)

2 37 7.75 24

4 24 8.75

6 18 9 9.5

8 16 9.5

10 14 9.7 6.6

12 12 9.7

14 10.5 9.5 5.5

16 9 8.5

Top of car, over one group (A); 

over empty center (B)

Contact 21 11

5 boxcars in a row 

containing unspecified 

amount of K-65

Along a line perpendicular to the 

axis of the line of cars, even with 

the center of the middle car

10 50

25 22

43 13

50 10.2

Note

The total length of a car is given as 57 feet.
Effective Date: 10/24/2003 Revision No. 00 Document No. ORAUT-TKBS-0005 Page 118 of 124

Table 34. Middlesex ore storage worker gamma-beta doses, 

1944–1949 (AEC 1949, Figure 3)

Worker type

Number of 

workers

Weekly gamma-beta 

dose, mrep

Guard 11 150

Laborer 20 500-600

Labor Foreman 1 250

Laboratory Technician 2 300

Maintenance 10 150-500, avg 300

Office --- 100

Timekeeper 1 250

Note: Doses were 65% gamma. K-65 residue was still coming here at this time.

Table 35. Measured dose rates at various positions in Plant 6 (MCW 1946).

Location Exposure, % tolerance

Center of 4 barrels of GLC (3 days old) 150

Between 2 barrels of GLC (3 days old) 120

Under M-3 containing GLC 120

Operator position while filling drums of GLC 110 (2)

Operator position while milling GLC - Sample Prep Room 130

Near drying oven -- Sample Prep Room 20

Skids of acid press cake (2-3 days old) 200 (5)

Floor surface under Feinc platform 30 (8)

In front of Feinc filter while filtering reslurry batch 210 (2)

Fork truck driver position while loading GLC(rail?)car 140

By M-1: on platform, about 250 gal GLC (3 days old) 90

By M-2 on platform, 5 hrs after addition of 6000 lbs ore 100

By M-3 on platform while adding ore (2000 lbs in tank) 70

On platform at M-19 while reslurrying skid cake batch 120

At operator position between skid and M-14 on platform 120

Above skid of GLC on M-14 280

By M-72, containing wash water 50

By M-83, 3/4 full of GLC (2 days old) 220

Floor around Feinc filters 110 (7)

Sump recovery skids, all full 80 (5)

Dempster body of Chemical 6BC 30

Dempster body of Chemical 6BP 40

Laboratory sample room, center 0

Laboratory sample room, in vault 80

Sample Prep Room, 1 foot from pile of K-65 in hood 60

Sample Prep Room, 1 foot from a 3-gal bottle of K-65 at equilibrium 160

Sample Prep Room, 1 foot from 3 2-qt bottles of K-65 at equilibrium 100

Sample Prep Room, operator position at Ro-Tap 50

Notes

Data is based on measurements taken in July and August 1946 at one foot from the source.

Figures are given in units of per cent of tolerance, where tolerance is defined only as the "amount of exposure that a 

person can receive 8 hrs/day for an indefinite period of time", presumably corresponding to 500 mrep/ wk at this time.

Parentheses indicate the number of measurements used to form the average.
Effective Date: 10/24/2003 Revision No. 00 Document No. ORAUT-TKBS-0005 Page 119 of 124

Table 36. Weekly external dose values, April 1942–December 1945. 

Lognormally-distributed doses (rem)

Median weekly photon dose GSD Median weekly electron dose GSD

0.1 1.85 0.124 1.79

Table 37. Organ doses (rem)a for a beam quality (HVL) of 2.5 mm Al and 80 kVp (1942-1958)

PA Chest Lat Chest Total (b)

Entrance Air Kerma (cGy) 0.20 (c) 0.50 (c)

Thyroid 7.5 x 10-2 (f) 3.8 x 10-2 1.1 x 10-1

Ovaries 3.4 x 10-2 2.8 x 10-2 6.2 x 10-2

Testes 1.8 x 10-3 1.7 x 10-3 3.5 x 10-3

Lungs (male) 8.4 x 10-2 9.7 x 10-2 1.8 x 10-1

Lungs (female) 9.0 x 10-2 1.1 x 10-1 2.0 x 10-1

Breast 9.8 x 10-3 1.3 x 10-1 1.4 x 10-1

Uterus 3.0 x 10-2 2.2 x 10-2 5.2 x 10-2

Bone Marrow (male) 1.8 x 10-2 1.9 x 10-2 3.7 x 10-2

Bone Marrow (female) 1.7 x 10-2 1.5 x 10-2 3.2 x 10-2

Total Body (male) 2.6 x 10-2 3.2 x 10-2 ( e) 5.8 x 10-2

Total Body (female) 2.4 x 10-2 3.0 x 10-2 (e) 5.4 x 10-2

Thymus 9.0 x 10-2 1.1 x 10-1 2.0 x 10-1

Liver/Gall Bladder 3.4 x 10-2 2.8 x 10-2 6.2 x 10-2

Eye/Brain 4.6 x 10-3 6.9 x 10-2 7.4 x 10-2

Esophagus 9.0 x 10-2 1.1 x 10-1 2.0 x 10-1

Stomach 9.0 x 10-2 1.1 x 10-1 2.0 x 10-1

Urinary Bladder 3.4 x 10-2 2.8 x 10-2 6.2 x 10-2

Colon/Rectum 3.4 x 10-2 2.8 x 10-2 6.2 x 10-2

Bone Surface 9.0 x 10-2 1.1 x 10-1 2.0 x 10-1

Skin (d) 2.7 x 10-1 6.9 x 10-1 9.6 x 10-1

Remainder 9.0 x 10-2 1.1 x 10-1 2.0 x 10-1

(a) For organs listed in ICRP 34 (1982) and proximal organs for input to IREP

(b) Sum of dose from the PA and lat views

(c) Kathren et al, to be published

(d) Skin dose is entrance skin exposure calculated from air kerma, multiplied by a backscatter 

factor of 1.35 from NCRP 102, Table B-8 

(e) The ICRP 34 DCFs appear to have been switched male/female

(f) The DCF for the AP cervical spine was used, corrected by a 43% depth dose factor (NCRP 

102, Table B-8)
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Table 38. Decontamination and decommissioning years, 1959-1961: maximum annual doses

(rem).

Plant

Exposure Potential Exposure Mode 4 6 6E 7

Inhalation 4.73 x 10-1 1.61 x 100 9.39 x 10-1 3.08 x 100

High Radon 2.13 x 10-2 7.77 x 10-2 4.52 x 10-2 1.48 x 10-1

Direct gamma 1.60 x 10-3 1.18 x 10-2 1.12 x 10-3 2.00 x 10-4

Total 4.96 x 10-1 1.70 x 100 9.86 x 10-1 3.23 x 100

Inhalation 2.39 x 10-1 6.73 x 10-2 7.25 x 10-2 7.52 x 10-2

Moderate Radon 1.07 x 10-2 3.23 x 10-3 3.49 x 10-3 3.62 x 10-3

Direct gamma 2.40 x 10-4 2.00 x 10-4 8.00 x 10-5 8.00 x 10-5

Total 2.50 x 10-1 7.07 x 10-2 7.61 x 10-2 7.89 x 10-2

Inhalation 1.22 x 10-1 3.09 x 10-3 1.07 x 10-2 2.28 x 10-3

Low Radon 5.71 x 10-3 1.48 x 10-4 5.17 x 10-4 1.10 x 10-4

Direct gamma 6.40 x 10-5 8.00 x 10-5 0.00 x 100 4.00 x 10-5

Total 1.28 x 10-1 3.31 x 10-3 1.12 x 10-2 2.43 x 10-3

Notes

Data in this table was calculated using the RESRAD-BUILD computer code (ANL 2003 ) and measured data 

from Refs. MCW 1958 and MCW 1959.

Both surface contamination and bulk (floor and wall contamination were taken into account in calculating the 

inhalation and radon contributions.

"High" exposure potential represents those working in the most contaminated areas, i.e., the former process 

areas; "moderate" represents those accessing the less contaminated areas or infrequently accessing the former 

process areas; and "low" represents those accessing the uncontaminated areas.
Effective Date: 10/24/2003 Revision No. 00 Document No. ORAUT-TKBS-0005 Page 121 of 124

Table 39. Postoperations years, 1962-1995: maximum annual doses (rem). 

Former Plant 6E Buildings Former Plant 7 Buildings

Exposure Potential Exposure Mode

Annual 

Dose (rem) Exposure Potential Exposure Mode

Annual 

Dose (rem)

Inhalation 3.74 x 10-3 Inhalation 6.08 x 10-3

High Radon 6.09 x 10-3 High Radon 1.01 x 10-2

Direct gamma 1.29 x 10-4 Direct gamma 8.00 x 10-4

Total 9.96 x 10-3 Total 1.70 x 10-2

Inhalation 2.40 x 10-3 Inhalation 2.31 x 10-3

Moderate Radon 4.06 x 10-3 Moderate Radon 3.79 x 10-3

Direct gamma 4.29 x 10-5 Direct gamma 4.00 x 10-4

Total 6.50 x 10-3 Total 6.50 x 10-3

Inhalation 9.56 x 10-4 Inhalation 8.00 x 10-4

Low Radon 1.62 x 10-3 Low Radon 1.36 x 10-3

Direct gamma 0.00 x 100 Direct gamma 2.00 x 10-4

Total 2.58 x 10-3 Total 2.36 x 10-3

Notes

The data in this table was calculated using the RESRAD-BUILD computer code (ANL 2003) and measured data from 

MCW (1961). Both surface contamination and bulk (floor and wall contamination were taken into account in calculating 

the inhalation and radon contributions.

"High" exposure potential represents those working in the most contaminated areas, i.e., the former process areas; 

"moderate" represents those accessing the less contaminated areas or infrequently accessing the former process 

areas; and "low" represents those who accessing the uncontaminated areas.
Effective Date: 10/24/2003 Revision No. 00 Document No. ORAUT-TKBS-0005 Page 122 of 124

Table 40. Summary of postoperational years exposure data as measured by two survey groups. 

1981 1986 1990 Overall Averages, 1981

Plant

Bldg 

Remaining

Radon 

daughters 

(WL)

Radon 

Yearly Avg, 

WLM/yr

Radon 

Yearly 

Avg, 

WLM/yr Source BG

Average Measured Levels

at St. Louis Site

1 K1E 0.009-0.02 0.05 0.12 Radon in air <1 pCi/L

0.4 – 37 pCi/L, avg daytime;

69 pCi/L max

1 25-1 0.001 0.01

Radon daughters 

in air <0.01 WL

0.0009-0.07 WL, avg daytime 

conc

1 25-2 0.0009 0.01

Gamma from Ra, 

U daughters 8 uR/hr 8-290 uR/yr @ 1 m above floor

1 38

1 40

2 50 0.0003

2 51 0.0005 0.02 0.01

2 51A 0.02 0.01

2 52A 0.07

2 52

0.0007-

0.001 0.00

6 100 0.02

6 101 0.10 0.01

6E 116-1 0.00 0.04

6E 116-2 0.00

6E 116B 0.01

6E 117-1 0.00

6E 117-2 0.00

7 700 0.00

7 704 0.03 0.00

7 705 0.00

7 706 0.00

7 708 0.01

Measurements are from ORNL (1981) and Applied Nuclear Safety (1986; 1990)
Effective Date: 10/24/2003 Revision No. 00 Document No. ORAUT-TKBS-0005 Page 123 of 124

APPENDIX A

Notes on How the "Tolerance" or "Preferred" Level for 

Insoluble Uranium Compounds in Air Was Calculated in 1948

(From AEC 1949)

1. The "tolerance" alpha radiation level to the lung is 30 mrep/week or 4.3 mrep/day.

Assumptions

2. The fraction of inhaled material retained in the lungs and pulmonary lymphatic tissue is 0.25.

3. The biological half-life of insoluble uranium compounds in the lung is 90 days.

4. The weight of a pair of lungs is 1000 grams.

5. An individual inhales 10 cubic meters per 8-hour working day.

1. μCi in lung required to deliver 4.3 mrep/day: = 

Calculations

2.2 × 106 × 60 × 24 × 8.86

 (5.2 × 107) (0.0043) 

where

5.2 × 107

 = number of MeV/g in one rep

0.0043 rep/day = daily acceptable dose rate

2.2 × 106

 = number of dis/min per μCi

60 × 24 = number of minutes per day

8.86 = sum of energies of alpha radiation from U-238 and U-234 in equilibrium, in 

MeV

= 8 × 10-6 μCi/g

2. Total μCi in lungs for 4.3 mrep/day:

= 1000 × 8 × 10-6

= 8 × 10-3 μCi

3. μCi per 10 m3 (inhaled in 8 hours) which will give 8 × 10-3 μCi to the lung at equilibrium (assuming 

exposure every day)

where

8 × 10-3 = μCi in lungs at equilibrium

0.25 = fraction of inhaled material deposited in the lung

90 = assumed biological half-life in the lungs, in days

= 2.54 × 10-4 μCi per 10 m3

1.4 = factor to convert half-life to mean life

4. μCi per m3

: 

= 2.54 × 10-5 

= 56 dpm per m
Effective Date: 10/24/2003 Revision No. 00 Document No. ORAUT-TKBS-0005 Page 124 of 124

5. Adjustment for actual exposure occurring up to 6 days a week, when 7 days was assumed:

56 dpm * (7 days/6 days) = 65 dpm)

(i.e., 5-6 days per week actually, since 56 dpm * (7 days/5 days) = 78 dpm

and

= 70 dpm per m3

= 50 μg per m3

Note: As AEC 1949 states, these calculations use no factor to account for nonuniform distribution in 

the lungs. It is also stated that the acceptable weekly dose rate for alpha dust exposures was going to 

be changed (presumably by AEC) to either 30 mrep/week or 15 mrep/we

1