ORAU TEAM

Dose Reconstruction

Project for NIOSH

Oak Ridge Associated Universities I Dade Moeller & Associates I MJW Corporation

Page 1 of 225

DOE Review Release 06/08/2009

Document Title:

Basis for Development of an Exposure Matrix for 

the Mallinckrodt Chemical Company St. Louis 

Downtown Site and the St. Louis Airport Site, St. 

Louis, Missouri

Document Number: ORAUT-TKBS-0005

Revision: 02 PC-1 

Effective Date: 05/25/2009

Type of Document: TBD

Supersedes: Revision 02 

Subject Expert(s): Janet L. Westbrook and Cindy W. Bloom

Site Expert(s): N/A

Approval: Richard Merrill Signature on File for Approval Date: 05/21/2009

Joseph S. Guido, Document Owner

Concurrence: Signature on File Concurrence Date: 05/21/2009

John M. Byrne, Objective 1 Manager

Concurrence: Signature on File Concurrence Date: 05/21/2009

Edward F. Maher, Objective 3 Manager

Concurrence: Signature on File Concurrence Date: 05/21/2009

Kate Kimpan, Project Director

Approval: Signature on File Approval Date: 05/25/2009

James W. Neton, Associate Director for Science

 New Total Rewrite Revision Page Change

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

REVISION AND DISCARD / DESTROY ALL COPIES OF THE PRIOR REVISION.
Document No. ORAUT-TKBS-0005 Revision No. 02 PC-1 Effective Date: 05/25/2009 Page 2 of 225

PUBLICATION RECORD

EFFECTIVE

DATE

REVISION 

NUMBER DESCRIPTION

10/24/2003 00 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.

03/10/2005 01 Approved Revision 01 initiated to incorporate comments on Sections 

7.0 and 8.0; to incorporate additional lately captured information 

regarding site work and external and radon exposure rates; to 

incorporate neutron exposure information; to incorporate information 

about the St. Louis Airport Site (SLAPS); and to provide additional 

guidance to dose reconstructors. Incorporates formal NIOSH review 

comments. Section 7.0, regarding external dose, remains on hold. 

Initiated by Janet L. Westbrook.

06/14/2007 02 Approved Revision 02 initiated to reflect comments of the ABRWH.

Constitutes a total rewrite of the document. Modified Purpose and 

Scope, and Appendix A to incorporate SEC information as instructed 

by NIOSH. Made minor editorial changes to eliminate readily 

discernable errors and eliminated extraneous intake tables. Added 

information to reconstruct doses from available monitoring and 

coworker data for the period 1949-1962, and to clarify some 

discrepancies between the early-in-the program draft and current 

guidance. Deleted Appendix D (which was formerly C). This revision 

changes the method for assigning uranium, radium and thorium 

coworker intakes and radon exposures and eliminates the application 

of equilibrium amounts of uranium progeny intakes to reconstruct 

dose. Incorporates formal internal and NIOSH review comments. 

Revised dates for operational period in sections 6 and 7 to end in 

1958. Removed references to unmonitored thorium and radium 

exposures (since reconstruction of such exposures has been 

determined to be not feasible by the SEC action). Added a section 

titled “unmonitored raffinate exposures” to cover the period 1948 – 

1958 and inserted the text “reserved” to indicate that reconstruction of 

internal exposure from such pathways is not covered in this TBD for 

the time period 1948 – 1958. This revision results in an increase in 

assigned dose and a PER is required. Training required: as 

determined by the Task Manager. Initiated by Cindy W. Bloom and 

Joseph Guido.
Document No. ORAUT-TKBS-0005 Revision No. 02 PC-1 Effective Date: 05/25/2009 Page 3 of 225

05/25/2009 02 PC-1 Page change initiated to incorporate SEC information on pages 10-13

in Section 1.0 (extension of class to 12/31/1958) as instructed by 

NIOSH. Updated NIOSH required language on pages 10 and 11 in 

Section 1.0. Added information on pages 106-108 in Section 6.5. 

Updated NIOSH reference on pages 106-108, 124, 128 and 216 in 

Sections 7.1, 7.3, 8.5, 8.6, and Attachment A, respectively. Added 

reference on page 141 in Reference Section. On pages 158-160, 

updated NIOSH and ORAUT references in Reference Section. On 

page 210, updated NIOSH reference in Table A-34. Table A-40 on 

page 215 was revised to clarify that tabulated intakes represent gross 

alpha activity and to provide guidance on isotopic mixture to be 

applied. Incorporates formal internal and NIOSH review comments. 

Training required: as determined by the Objective Manager. 

Initiated by Joseph S. Guido.
Document No. ORAUT-TKBS-0005 Revision No. 02 Effective Date: 06/14/2007 Page 4 of 224

TABLE OF CONTENTS

SECTION TITLE

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

PAGE

9

1.0 Purpose and Scope.................................................................................................................10

2.0 Introduction .............................................................................................................................13

3.0 History of Site Use ..................................................................................................................14

3.1 History of Use of the St. Louis Downtown Site (SLDS) ................................................14

3.2 History of Use of the St. Louis Airport Storage Site (SLAPS or 

SLAPSS) .....................................................................................................................21

4.0 Description of the Mallinckrodt Uranium Refining Process, Other Processes, 

and SLAPS Storage and Disposal Activities............................................................................24

4.1 The Basic Process.......................................................................................................24

4.2 The Wartime Period (April 1942 - April 1945) ..............................................................26

4.3 The Early Postwar Period (May 1945 - December 1949).............................................27

4.4 The Later Postwar Period (1950 - 1958)......................................................................28

4.5 Other Processes..........................................................................................................31

4.6 Ores and Other Feed Forms........................................................................................32

4.7 Residues and Effluents................................................................................................34

5.0 Radiological Characteristics, Conditions, Considerations, and Available Data ........................36

5.1 Units, Limits, and Recommendations...........................................................................38

5.2 Radioactivity Content and Handling of the Ore, Uranium Products, 

and Residues ..............................................................................................................41

5.2.1 Ores and Other Feeds and the Initial Process Steps........................................41

5.2.2 Uranium Products ............................................................................................45

5.2.3 Residues and Other Wastes and Reprocessing...............................................47

5.2.4 Sampling and Laboratory Activities for All Forms .............................................53

5.2.5 Safety, Health, Decontamination, and Laundry Facilities..................................56

5.3 Internal Dose Considerations.......................................................................................56

5.3.1 Particle Size, Solubility, and Composition Considerations; Air 

Sampling Methods ...........................................................................................57

5.3.2 Airborne Dust Levels........................................................................................60

5.3.3 Respirator Use................................................................................................. 68

5.3.4 Radon ..............................................................................................................69

5.3.5 Surface Contamination.....................................................................................73

5.3.6 Information and Available Data Regarding Urinalyses......................................77

5.3.7 Information and Available Data Regarding Other Types of 

Bioassay ..........................................................................................................79

5.4 External Dose Considerations .....................................................................................81

5.4.1 Gamma, Beta (Electron), and Nonspecific Beta-Gamma 

Exposures........................................................................................................82

5.4.2 Neutrons ..........................................................................................................88

5.4.3 Information and Available Data Regarding Film Badges and 

Extremity Dosimeters.......................................................................................90
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5.4.3.1 Information Regarding the Type and Composition of 

Film Badges Used at Mallinckrodt .................................................... 90

5.4.3.2 Film Badge Technical and Processing Information ........................... 91

5.4.3.3 Film Badge Monitoring Periods and Wearing 

Practices........................................................................................... 93

5.4.3.4 Film Badge Record Types, Arrangement, and 

Availability ........................................................................................ 94

5.4.3.5 Extremity Dosimeters........................................................................ 95

5.4.3.6 Film Badges Used as Area Dosimeters ............................................ 96

5.4.4 Information and Available Data Regarding Occupational X-Ray 

Examinations ...................................................................................................96

5.4.5 Skin Contaminations and Other Radiological Incidents ....................................96

5.5 Other Data of Dosimetric Interest ................................................................................97

5.5.1 Number of Workers..........................................................................................97

5.5.2 Number of Hours Worked per Week ................................................................98

5.5.3 Job Types, Job Histories, Work Areas, and Work, Access, and 

Clothing Practices ............................................................................................98

5.5.4 Miscellaneous Product Information ................................................................101

5.5.5 Missouri and St. Louis Area Background Levels ............................................102

6.0 Reconstructing Internal Doses During the Operating Years ..................................................102

6.1 Uranium Intakes ........................................................................................................103

6.1.1 Assignment of Uranium Intakes for Unmonitored Workers or 

Unmonitored Periods .....................................................................................103

6.2 Radium Intakes .........................................................................................................104

6.3 Thorium Intakes.........................................................................................................105

6.4 Radon Exposure........................................................................................................105

6.5 Unmonitored raffinate exposures (1949 - 1958).........................................................106

7.0 Reconstructing External Doses During the Operating Years .................................................107

7.1 Assumptions and General Instructions ......................................................................107

7.2 Information Regarding the 1950 AEC Cumulative Dose 

Reconstruction Study ................................................................................................108

7.3 Gamma and Beta Dose .............................................................................................108

7.3.1 Selecting a Surrogate Cohort and Formulating a Surrogate 

External Data Set...........................................................................................109

7.3.1.1 Notes Regarding Dose Summaries................................................. 109

7.3.2 Determining Dose - Monitored Period 1949 - 1958 ........................................110

7.3.2.1 Monitored Workers ......................................................................... 110

7.3.2.2 Unmonitored Workers – Plants Still Operating under 

MED/AEC Contracts....................................................................... 110

7.3.3 Unmonitored Workers – Plants No Longer Operating under 

MED/AEC Contracts ......................................................................................110

7.4 X-Ray Dose...............................................................................................................111

7.4.1 Mallinckrodt-Specific Information ...................................................................111

7.4.2 Chest X-Rays................................................................................................. 111

7.5 Neutron Dose ............................................................................................................111

7.6 Other Dose Contributions ..........................................................................................112

7.6.1 Extremity Dose...............................................................................................112

7.6.2 Submersion Dose ..........................................................................................112

7.6.3 Shallow Dose................................................................................................. 112
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8.0 Determination of Exposures Due to Residual Contamination Remaining From 

MED/AEC Operations ...........................................................................................................113

8.1 Description of the Decontamination and Demolition Work at the St. 

Louis Main Site, Including Health and Safety Precautions .........................................113

8.2 Post-Decontamination Conditions at the St. Louis Main Site .....................................117

8.3 General Considerations and Approach for St. Louis Main Site 

Decontamination and Post-Decontamination Dose Reconstruction ...........................118

8.4 General Considerations and Approach for SLAPS Post-Operations 

Dose Reconstruction ................................................................................................. 118

8.5 Calculation of Decontamination and Post-Decontamination Internal 

and External Exposures for the St. Louis Main Site When Individual 

and Surrogate Data Are Lacking................................................................................120

8.5.1 Assumptions Made in RESRAD-BUILD and Related Internal 

Exposure Calculations ...................................................................................120

8.5.2 Results of RESRAD-BUILD Calculations and Conversion to 

Dosimetrically Useful Quantities.....................................................................122

8.5.3 Assumptions for and Results of Ingestion Calculations ..................................123

8.5.4 Assumptions for and Results of External Exposure 

Calculations ...................................................................................................123

8.6 Calculation of Decontamination and Post-Decontamination Internal 

and External Exposures for SLAPS When Individual and Surrogate 

Data are Lacking .......................................................................................................124

8.6.1 Assumptions for Work Time and Worker Types .............................................125

8.6.2 Assumptions for Radioactive Source Terms...................................................125

8.6.3 Assumptions for Inhalation Exposure Calculations.........................................126

8.6.4 Assumptions for Ingestion Exposure Calculations..........................................127

8.6.5 Assumptions for and Results of Radon Exposure Calculations ......................127

8.6.6 Assumptions for and Results of External Exposure 

Calculations ...................................................................................................127

8.7 Use of Decontamination and Post-Decontamination Exposure Data for 

St. Louis Main Site Dose Reconstruction...................................................................128

8.8 Use of Decontamination and Post-Decontamination Exposure Data for 

SLAPS Dose Reconstruction.....................................................................................129

9.0 Attributions and Annotations ................................................................................................. 129

References .......................................................................................................................................130

ATTACHMENT A, TABLES ..............................................................................................................162

ATTACHMENT B, U. S. DEPARTMENT OF LABOR CORRESPONDENCE TO 

NIOSH CONCERNING DOSE RECONSTRUCTION FOR MEMBERS OF 

THE MALLINCKRODT DESIGNATED SPECIAL EXPOSURE COHORT ............................220

ATTACHMENT C, NOTES ON HOW THE “TOLERANCE” OR “PREFERRED” 

LEVEL FOR INSOLUBLE URANIUM COMPOUNDS IN AIR WAS 

CALCULATED IN 1948 (FROM AEC 1949B) .......................................................................223
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ATTACHMENT D, EXAMPLE CALCULATION OF DAILY WEIGHTED AVERAGE 

EXPOSURE (DWE) .............................................................................................................225

LIST OF TABLES

TABLE TITLE

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

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

PAGE

165

A-2 Summary chronology of operations at the Mallinckrodt St. Louis Downtown 

Site and at the St. Louis Airport Storage Site...................................................................... 168

A-3 Principal changes made in sites, processes, and equipment .............................................. 169

A-4 Types and quantities of material produced in association with Mallinckrodt 

uranium refining and related operations.............................................................................. 170

A-5 Functional and process keywords and codes ..................................................................... 175

A-6 Thorium and daughter content of the AM-7 residue ............................................................ 180

A-7 Airborne uranium particle size in process areas.................................................................. 180

A-8 Uranium dust concentrations, Plants 4 and 6, in alpha dpm/m3 .......................................... 181

A-9 Uranium dust concentrations, Plants 6E and 7, in alpha dpm/m3 ........................................ 182

A-10 Early air sample data, short-term sample readings in alpha dpm/m3................................... 182

A-11 Airborne uranium dust concentrations in Plant 4 Areas, 1948: AEC versus 

Mallinckrodt measured data................................................................................................ 183

A-12 Plant 4 measured daily weighted average exposure concentrations ................................... 184

A-13 Plant 6 measured daily weighted average exposure concentrations ................................... 185

A-14 Plant 6E measured daily weighted average exposure concentrations................................. 187

A-15 Plant 7 measured daily weighted average exposure concentrations ................................... 188

A-16 Plant 7E measured daily weighted average exposure concentrations................................. 188

A-17 Average and highest airborne dust concentrations in the laundry ....................................... 189

A-18 Job titles and classifications, with geometry factors ............................................................ 190

A-19 Uranium dust daily weighted average exposure levels, Plant 4........................................... 194

A-20 Uranium dust daily weighted average exposure levels, Plant 6........................................... 195

A-21 Uranium dust daily weighted average exposure levels, Plant 6E ........................................ 196

A-22 Radioactive dust daily weighted average exposure levels, Plant 7 (including 

the Slag Separation Plant) and Plant 7E............................................................................. 197

A-23 Early radon data, 1945 ....................................................................................................... 198

A-24 Measured radon concentrations, 1947-1949, in units of 1 × 10-10 Ci/L ................................ 199

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

of 1 × 10-10 Ci/L................................................................................................................... 201

A-26 RESERVED........................................................................................................................ 202

A-27 Contamination levels and associated dose rates from work clothing................................... 202

A-28 RESERVED........................................................................................................................ 202

A-29A RESERVED........................................................................................................................ 202

A-29B RESERVED........................................................................................................................ 202

A-29C RESERVED........................................................................................................................ 202

A-29D RESERVED........................................................................................................................ 202

A-29E RESERVED........................................................................................................................ 202

A-29F RESERVED........................................................................................................................ 202

A-30 RESERVED........................................................................................................................ 202

A-31A RESERVED........................................................................................................................ 202

A-31B RESERVED........................................................................................................................ 202
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A-31C RESERVED........................................................................................................................ 202

A-31D RESERVED........................................................................................................................ 202

A-31E RESERVED........................................................................................................................ 202

A-31F RESERVED........................................................................................................................ 202

A-32 Exposure rates from K-65 residue and Q-11 ore, in mR/hr ................................................. 203

A-33 Measured dose rates in various work areas........................................................................ 204

A-34 Neutron dose rates and doses from the alpha-neutron reaction sources and 

from a RaBe source ............................................................................................................ 210

A-35 External gamma dose rates from processing AM-7 residue to concentrate 

thorium ............................................................................................................................... 211

A-36 Reserved............................................................................................................................ 211

A-37 Monitoring experience from the decontamination/demolition period, 1960-

1961 ................................................................................................................................... 212

A-38 Summary of post-decontamination exposure data as measured by two 

survey groups..................................................................................................................... 213

A-39 Measured data used to produce source terms for decontamination and postdecontamination exposure calculations .............................................................................. 214

A-40 Annual internal exposures during the decontamination and postdecontamination periods..................................................................................................... 215

A-41 Annual external photon organ doses, decontamination and postdecontamination periods..................................................................................................... 216

A-42 Annual external electron organ doses during the decontamination and postdecontamination periods..................................................................................................... 218

A-43 Annual dust inhalation, ingestion, and radon daughter intakes during the 

SLAPS postoperations period............................................................................................. 218

A-44 Annual external beta doses, gamma doses and gamma organ doses during 

the SLAPS postoperations period....................................................................................... 219
Document No. ORAUT-TKBS-0005 Revision No. 02 Effective Date: 06/14/2007 Page 9 of 224

ACRONYMS AND ABBREVIATIONS

AEC Atomic Energy Commission

AP anterior-to-posterior

CER Center for Epidemiological Research (Oak Ridge Associated Universities)

Ci/L curies per liter

D&D decontamination and decommissioning

DCF dose conversion factor

DWE levels time-weighted daily average exposure levels

EEOICPA Energy Employees Occupational Illness Compensation Program Act

FMFL fluorinated MFL

GSD geometric standard deviation

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

HVL Half-value layer

LOD limit of detection

lfm linear feet per minute

MAC Maximum Allowable Concentration

ME or MEP Minor Elements Production

MED Manhattan Engineer District

MFL MgF2 liner

MgX magnesia concentrate feed

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
Document No. ORAUT-TKBS-0005 Revision No. 02 PC-1 Effective Date: 05/25/2009 Page 10 of 225

Code terms for the various uranium forms (e.g., TA-7) are given in Table A-5 in Attachment A.

1.0

Technical basis documents and site profile documents are not official determinations made by the 

National Institute for Occupational Safety and Health (NIOSH) but are rather general working 

documents that provide historic background information and guidance to assist in the preparation of 

dose reconstructions at particular sites or categories of sites. They will be revised in the event 

additional relevant information is obtained about the affected site(s). These documents may be used 

to assist NIOSH staff in the completion of the individual work required for each dose reconstruction.

PURPOSE AND SCOPE

In this document the word “facility” is used as a general term for an area, building, or group of 

buildings that served a specific purpose at a site. It does not necessarily connote an “atomic weapons 

employer facility” or a “Department of Energy [DOE] facility” as defined in the Energy Employees 

Occupational Illness Compensation Program Act [EEOICPA; 42 U.S.C. § 7384l(5) and (12)]. 

EEOICPA defines a DOE facility as “any building, structure, or premise, including the grounds upon 

which such building, structure, or premise is located ... in which operations are, or have been, 

conducted by, or on behalf of, the Department of Energy (except for buildings, structures, premises, 

grounds, or operations ... pertaining to the Naval Nuclear Propulsion Program)” [42 U.S.C. § 

7384l(12)]. Accordingly, except for the exclusion for the Naval Nuclear Propulsion Program noted 

above, any facility that performs or performed DOE operations of any nature whatsoever is a DOE 

facility encompassed by EEOICPA.

For employees of DOE or its contractors with cancer, the DOE facility definition only determines 

eligibility for a dose reconstruction, which is a prerequisite to a compensation decision (except for 

members of the Special Exposure Cohort). The compensation decision for cancer claimants is based 

on a section of the statute entitled “Exposure in the Performance of Duty.” That provision [42 U.S.C. § 

7384n(b)] says that an individual with cancer “shall be determined to have sustained that cancer in the 

performance of duty for purposes of the compensation program if, and only if, the cancer ... was at 

least as likely as not related to employment at the facility [where the employee worked], as 

determined in accordance with the POC [probability of causation1

The statute also includes a definition of a DOE facility that excludes “buildings, structures, premises, 

grounds, or operations covered by Executive Order No. 12344, dated February 1, 1982 (42 U.S.C. 

7158 note), pertaining to the Naval Nuclear Propulsion Program” [42 U.S.C. § 7384l(12)]. While this 

definition excludes Naval Nuclear Propulsion Facilities from being covered under the Act, the section 

of EEOICPA that deals with the compensation decision for covered employees with cancer [i.e., 42 

U.S.C. § 7384n(b), entitled “Exposure in the Performance of Duty”] does not contain such an 

exclusion. Therefore, the statute requires NIOSH to include all occupationally-derived radiation 

exposures at covered facilities in its dose reconstructions for employees at DOE facilities, including 

radiation exposures related to the Naval Nuclear Propulsion Program. As a result, all internal and 

external occupational radiation exposures are considered valid for inclusion in a dose reconstruction. 

No efforts are made to determine the eligibility of any fraction of total measured exposure for inclusion 

in dose reconstruction. NIOSH, however, does not consider the following exposures to be 

occupationally derived (NIOSH 2007): 

] guidelines established under 

subsection (c) ...” [42 U.S.C. § 7384n(b)]. Neither the statute nor the probability of causation 

guidelines (nor the dose reconstruction regulation, 42 C.F.R. Pt. 82) define “performance of duty” for 

DOE employees with a covered cancer or restrict the “duty” to nuclear weapons work (NIOSH 2007). 

 

1 The U.S. Department of Labor (DOL) is ultimately responsible under the EEOICPA for determining the POC. 
Document No. ORAUT-TKBS-0005 Revision No. 02 PC-1 Effective Date: 05/25/2009 Page 11 of 225

• Background radiation, including radiation from naturally occurring radon present in 

conventional structures

• Radiation from X-rays received in the diagnosis of injuries or illnesses or for therapeutic 

reasons

Under EEOICPA, employment at an AWE facility is categorized as either (1) during the DOE contract 

period (i.e., when the AWE was processing or producing material that emitted radiation and was used 

in the production of an atomic weapon), or (2) during the residual contamination period (i.e., periods 

that NIOSH has determined there is the potential for significant residual contamination after the period 

in which weapons-related production occurred). For contract period employment, all occupationally 

derived radiation exposures at covered facilities must be included in dose reconstructions. This 

includes radiation exposure related to the Naval Nuclear Propulsion Program and any radiation 

exposure received from the production of commercial radioactive products that were concurrently 

manufactured by the AWE facility during the covered period. NIOSH does not consider the following 

exposures to be occupationally derived (NIOSH 2007):

• Background radiation, including radiation from naturally occurring radon present in 

conventional structures

• Radiation from X-rays received in the diagnosis of injuries or illnesses or for therapeutic 

reasons

For employment during the residual contamination period, only the radiation exposures defined in 42 

U.S.C. § 7384n(c)(4) [i.e., radiation doses received from DOE-related work] must be included in dose 

reconstructions. Doses from medical X-rays are not reconstructed during the residual contamination 

period (NIOSH 2007). It should be noted that under subparagraph A of 42 U.S.C. § 7384n(c)(4), 

radiation associated with the Naval Nuclear Propulsion Program is specifically excluded from the 

employee’s radiation dose. This exclusion only applies to those AWE employees who worked during 

the residual contamination period. Also, under subparagraph B of 42 U.S.C. § 7384n(c)(4), radiation 

from a source not covered by subparagraph A that is not distinguishable through reliable 

documentation from radiation that is covered by subparagraph A is considered part of the employee’s 

radiation dose. This site profile covers only exposures resulting from nuclear weapons-related work. 

Exposures resulting from non-weapons-related work, if applicable, will be covered elsewhere.

The following information from the Department of Energy’s Office of Worker Advocacy (DOE 2009) 

defines the EEIOCPA dose reconstruction periods.
Document No. ORAUT-TKBS-0005 Revision No. 02 PC-1 Effective Date: 05/25/2009 Page 12 of 225

Site: Mallinckrodt Chemical Co., Destrehan St. Plant

Alternate names: St. Louis Downtown Site, Mallinckrodt Chemical Works, MCW

Location: St. Louis, Missouri

Covered Period: 1942-1962 

Remediation Period: 1995

Site: St. Louis Airport Storage Site (SLAPS)

Alternate names: Robertson Airport, Robertson Storage Area

Location: St. Louis, Missouri

Covered Period: 1946-1966 

Residual Period: 1967-1998

This document provides information for the reconstruction of radiation doses to workers at the 

Mallinckrodt Chemical Company St. Louis Downtown Site, which refined and processed uranium 

under contract to the United States government and for the St. Louis Airport Storage Site, which 

received residues from Mallinckrodt operations from 1946 until 1958, after which it was used until 

about 1967 for residue storage and disposal. 

This document describes exposure conditions 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 

period of such operations began in April 1942. Additionally, exposure due to residual contamination is 

considered for the period 1959-1962 and exposure due to remediation activities is considered for

1995.

This document also describes exposure conditions for Mallinckrodt workers who performed operations 

at the St. Louis Airport (Storage) Site (SLAPS or SLAPSS), the waste repository site north of the 

former St. Louis Municipal Airport in Robertson, Missouri, from 1946 to 1958. Additionally, exposure 

due to residual contamination left at the site and during further work at the site is specifically 

considered for the period 1959-1962, but would likely be an overestimate for the operational period 

through 1966, and the residual contamination period from 1967 through 1998. 

NIOSH has determined, and the Secretary, Health and Human Services has concurred, that it is not 

feasible to reconstruct radiation doses (both internal and external) for individuals who worked at 

Mallinckrodt prior to 1949 (HHS 2005a). For the period 1949-1957 NIOSH has determined, and the 

Secretary, Health and Human Services has concurred, that it is not feasible to reconstruct internal 

radiation doses from internal exposures to non-uranium radionuclides for Mallinckrodt raffinate 

workers, nor is it possible to identify which workers were raffinate workers (HHS 2005b). The period 

of applicability for this class was extended through December 31, 1958 (HHS 2009).

For any claim referred to NIOSH regarding an employee, 

• who was employed during the Cohort period but because of limited employment during this 

period is not a member of the Cohort, or 

• who is a member of the Cohort and whose cancer is not defined as a specified cancer under 

EEOICPA 
Document No. ORAUT-TKBS-0005 Revision No. 02 PC-1 Effective Date: 05/25/2009 Page 13 of 225

(and so is not eligible for compensation under EEOICPA without a dose reconstruction), NIOSH will 

continue to attempt to complete a Mallinckrodt dose reconstruction for the exposure period beginning 

in 1949, using whatever information is available about that member's entire work history. 

Department of Labor’s letter to NIOSH dated December 7, 2005, included in Appendix A of this 

document, notes that the Secretary of Health and Human Service’s report on the Special Exposure 

Cohort designation for Mallinckrodt Destrehan St. Plant for the years 1949-1957, discusses that that 

there is insufficient information to establish certain individual internal doses for Mallinckrodt workers 

but that it is feasible to reconstruct individual external exposures. This time period was subsequently 

extended to include all of 1958 (HHS 2009). Since NIOSH has determined that it is possible to 

estimate external exposure, it will continue to perform dose reconstructions for cases with nonspecified cancers for the years 1949-1958. While Health and Human Service’s report on the Special 

Exposure Cohort designation does not reference NIOSH's ability to reconstruct internal radiation 

doses in any manner, the Mallinckrodt site profile is written to allow for the reconstruction of certain 

internal exposures during the years 1949-1958. 

Consequently, only internal radiation doses associated with specific bioassay types (uranium, radon in 

breath, and thorium) and radon area measurements will be estimated for individuals who worked for 

Mallinckrodt during the period 1949 through 1958. 

2.0

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-Hilliard

1967; FUSRAP undated b). The work began immediately and by July 1942 Mallinckrodt was 

producing almost a ton of UO2 per day (Fleishman-Hilliard 1967; Mallinckrodt 1994; FUSRAP undated 

b). Later in 1942, Mallinckrodt started production of UF4 and in July 1943, 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 2002).

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. Like most of the early MED plants, the early Mallinckrodt facilities were large-scale 

expansions of laboratory processes and techniques and were designed without pilot plant studies or 

any special provision for health and safety (AEC 1951b). Also, temporary types of construction 

methods were used; even Plant 6, which was completed in mid-1946 and did have the benefit of an 

engineering pilot plant study, was of temporary-type construction because it was to process only highgrade pitchblende and the available supply was thought to be limited (AEC 1951b). The permanent 

status of the Mallinckrodt site was not established until 1946 (AEC 1951b; Mason 1958a). 

Engineered and administrative provisions for health and safety were instituted as high exposures 

were found in the course of monitoring (Hickey and Dupree 1984). 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 (except possibly for a pilot project starting in 

mid-1945) and urinalysis in the summer of 1948 (MCW 1950c) (some urinalysis was done earlier, but 

there was some question about its validity – see Section 5.3.6). 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 1955d; AEC 1950a), AEC’s Health and Safety Laboratory (HASL) also did 
Document No. ORAUT-TKBS-0005 Revision No. 02 Effective Date: 06/14/2007 Page 14 of 225

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, except for some special cases. 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 for a time (MCW 1955d), but 

eventually Mallinckrodt did its own urinalyses again (ORAU 1983b). It appears that most of the 

records of this monitoring are available. The external dose records are mostly available from 1946 

on, but are missing for the period 1942-1944 and for most of 1945. Some external dose records from 

1954 to the end of operations are available. The internal dose records (as bioassays) are mostly 

available from 1948 to the end of operations, but are missing for the period 1942-1947. Records for 

the post-operations decontamination and decommissioning should be found with the Weldon Spring 

records since film badges were issued from Weldon Spring (MCW 1961b), but to date the Weldon 

Spring records are only available for individual claims.

The St. Louis Airport Storage Site (SLAPS or SLAPSS) was acquired by AEC in 1946 for use in 

storing residues arising from pitchblende ore processing (AEC 1959; ORNL 1979). Later, it was also 

used to store other types of residues and some ore (AEC 1956d; Mallinckrodt 1994). Although 

MED/AEC operated the site itself from 1946 until July 1953, it appears that Mallinckrodt and perhaps 

some Ledoux Company workers actually did the work (MCW 1949d; MCW 1949p; ORAU 1989); 

Mallinckrodt ran the site under contract to AEC from July 1953 to 1958 and possibly until 1962, when 

control of the site passed to a private concern that was given a license to take over the stored 

residues (AEC 1959; ORAU 1989; AEC 1972). Thus the film badge records for airport workers from

1946 until at least 1958 should be found among the Mallinckrodt film badge records; film badge 

records after that may possibly be among the Weldon Spring records.

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.

3.0

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 

A-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 A-2.

HISTORY OF SITE USE

MED was required to store some of the residues produced from the pitchblende ore (AEC 1959) 

because the ore supplier, the African Metals Company, retained ownership of the radium and valuable 

metals remaining in the residues. Because of the volume of the residues and the high gamma and 

radon exposures associated with storage at the Mallinckrodt site, MED acquired a tract of land located 

in Robertson, Missouri near the St. Louis Municipal Airport (Labert Field) (AEC 1959) to use to store 

the residues and some of the pitchblende ore (AEC 1959). This storage area was formally called the 

St. Louis Airport Storage Site (SLAPS or SLAPSS), but in Mallinckrodt and MED/AEC documents it 

was often referred to simply as “the airport.”

3.1 HISTORY OF USE OF THE ST. LOUIS DOWNTOWN SITE (SLDS)

Plants 1 and 2 were already in existence and used for other operations by Mallinckrodt at the time 

uranium-refining operations started in 1942 (Mason 1977). MED (1944p) states that as of November 

1944, MED operations occupied a small part of Plant 1, part of Plant 2, and all of Plant 4.
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Laboratory development work started in April 1942 in Plant 1’s Building 25 (second floor) and in the 

alley on the southwest side of the building (Mason 1977). (Mallinckrodt 1994 states that work began 

in Plant 2 in uranium-refining operations in April 1942, but the Mason 1977 chronology appears to be 

more plausible.) Additional laboratory and 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).

In Plant 1, Building 25 contained the project offices, while Building P contained the engineering office. 

MED (1944p) states that Building Z contained other offices; Mason (1977) specified that the early 

uranium project was a semi-works operation for which administrative services were furnished from 

Mallinckrodt headquarters in Building Z. Building A in Plant 1 was the base for the plant mechanical 

department, which provided general maintenance services (Mason 1977; MED 1944p). Mason (1977) 

states that the laboratory in Building 25-2W provided quality control services.

Facilities for batch production were installed in Plant 2’s Buildings 50, 51, 51A, and 52 to produce 

uranium trioxide (UO3) from ore concentrates (Mason 1977; Hickey and Dupree 1984; Mallinckrodt 

1994). The concentrates were “digested” (dissolved) in nitric acid in Building 51 to produce uranyl 

nitrate (Mason 1977; ORNL 1981; Mallinckrodt 1994; Fleishman-Hilliard 1967). The digested liquid 

was then transferred to Building 52 to be purified by ether extraction to uranyl nitrate hexahydrate 

(UNH) (MED 1944p; Mason 1977; ORNL 1981; Mallinckrodt 1994); however, sources such as ORNL 

(1981) suggest that an area outside of Buildings 51 and 52, referred to as 51X and 52X, was covered 

with a canopy and used temporarily for extraction. The UNH was converted in Building 51A first to 

UO3 and then to UO2 (Mason 1977; ORNL 1981; Mallinckrodt 1994). The UO2 was apparently 

packaged in Building 51A (Mason 1977). By July 1942 Mallinckrodt was producing approximately a 

ton of uranium oxide (UO2) per day (Mason 1977; Mallinckrodt 1994).

Part of Building 38 was used as the change house for the uranium project workers (Mason 1977). 

Buildings 40, 45, 45A, and 47 were used as warehouse buildings for raw, in-process, and finished 

materials (MED 1944p). Building 50 was also used as a utility area, as a tank storage area for liquids 

to be used in Building 51, as a storage area for incoming feed materials and packaged products, and 

as a mechanical repair area (Mason 1977; Mallinckrodt 1994).

Building 55 contained the “shotgun” laboratory that tested UO3 samples (MED 1944l; Mason 1977; 

ORNL 1981) and also operated as a pilot plant from 1943 until some time in 1947. The shotgun 

laboratory was designated as a special restricted area because of the radium-beryllium (RaBe) source 

used for the sample testing (Mason 1977). 

Work began in 1942 on production of UF4 by the high-temperature gas-solid reaction of UO2 with HF; 

for a short time, one corner of Building 50 (the “sulfur burner room”) was used for development work 

on this process (Mason 1977; ORNL 1981).

Experimental processing of high-grade pitchblende ores began in Plant 1 in about July 1944 (Mason 

1977; Mallinckrodt 1994). Because of the high radium content of these ores and the residues 

resulting from the processing, Mallinckrodt wanted to confine the materials to a limited area and to 

prevent their spread into other plant areas (Mason 1977). Mallinckrodt therefore instituted strict 

controls of the developmental work and also decided to do pilot-scale work to obtain design data for 

eventual construction of a production-level refinery (Mason 1977). The laboratory for the testing of 

radium extraction methods was set up in Building 25-2 and the pilot plant in Building K-1E 

(Mallinckrodt 1994; ORNL 1981; Hickey and Dupree 1984 (which states that the pilot plant was in 

Building 25-2)) and in the alley (Fleishman-Hilliard 1967). The radium-free feed liquor that resulted 

from the Building K-1E digestion and precipitation processes was conveyed in containers by hand cart 
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to Building 52, where it was first treated in vessels outside the building (presumably in the alley) 

(Mason 1977). The liquid was then taken into Building 52 for conversion as usual to UO3 and then to 

UO2 (Mason 1977). By the end of 1944, Mallinckrodt had obtained sufficient data to begin the 

engineering design of a pitchblende ore refinery (Mason 1977). Mason (1977) states that “radiumbearing materials were not processed or stored at any time in the Plant 2 buildings,” but consistent 

with the usage of the term “radium-bearing materials” in various documents from the 1940’s and 

1950’s, it may be the high-radium-level residues that are being referred to. Mason (1977) also states 

that solid residues from the extraction were drummed, stored temporarily, and transferred from Plant 2 

to other government sites for scrap recovery.

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). It is not clear whether it is this plant or the Building K-1E plant that is the 

one referred to as the “Pilot Plant” in early film badge records; perhaps the term included both. Work 

at Plant 2 continued until 1946, when the plant was closed in preparation for the startup of the newly 

built Plant 6 (Mason 1958a; Mason 1977; Eisenbud 1975; ORNL 1981; Mallinckrodt 1994). Mason 

(1977) also states that when Plant 6 started up, all uranium work at Plants 1 and 2 stopped and the 

buildings were sealed to await further cleanup (but see below).

Plant 1 was not used after about 1945 (ORNL 1981) or early 1946 (Mason 1977); the offices and 

laboratories apparently moved to Plant 6. In particular, administrative and related services moved 

from Building Z to Plant 6 at this time; however, the quality control services moved to a laboratory in 

Building 400 of Plant 4 (Mason 1977). Although as noted above the Plant 2 operating area was said 

to have been closed in 1946 when production moved to Plant 6, Mason (1958a) states that milling of 

UO3 and unloading of (some) pots was not discontinued at Building 51 (Plant 2) until about 1947 and 

ORAU (1977) states that this building was closed about January 1947. Thus it can conservatively be 

assumed that Plant 2 was in limited operation through 1946.

From late November 1942 through April 1943, Plant 4, a lumber sash and door works, was hastily 

converted for uranium processing (AEC 1951b; Mason 1958a) and dubbed “the metal plant” (AEC 

1949b; AEC 1951b). It replaced the uranium metal production work done in Building 25-2 (Hickey and 

Dupree 1984). Production of uranium tetrafluoride (UF4) began in Building 400 in April 1943 after the 

process had been scaled up from the early work in Building 50 of Plant 2 (AEC 1951b; Mason 1977; 

Fleishman-Hilliard 1967). Production of uranium metal began in Buildings 400 and 401B in July 1943 

(AEC 1951b; Mason 1977; Mallinckrodt 1994). Magnesium metal for the process was stored in 

Building 406, while the residues from the process (e.g., slag) were stored in Building 408 (Mason 

1977). The principal production of metal moved from Plant 4 to Plant 6E in 1950 (Mallinckrodt 1994) 

although some metal was still produced in Plant 4 (AEC 1951b; Hickey and Dupree 1984) and some 

UF4 was still produced there until perhaps early 1953 (Hickey and Dupree 1984). After the 

development of a continuous furnace process, UO2-to-UF4 production moved to Building 705 of Plant 

7 (7N) in late 1951 (overlapping Plant 4’s production) (AEC 1951b; Fleishman-Hilliard 1967; Mason 

1977 (which says that Plant 4 UF4 production was completely shut down in 1951); Hickey and Dupree 

1984; Mallinckrodt 1994). In about 1950 or 1951 (ORNL 1981 says 1956, but this is evidently 

incorrect), 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” (Mason 1977; ORNL 1981). 

The “dingot” metal production process was developed and conducted at Plant 4 in the mid-1950’s 

(AEC 1956a; Fleishman-Hilliard 1967), along with sporadic ordinary metal “derby” production on a 

developmental basis. Plant 4 was used until 1956 (Mason 1977; ORNL 1981; Mallinckrodt 1994; 

FUSRAP 2003c). 
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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; ORAU 1983a; Mallinckrodt 1994)), Plant 6 

was built in 1945-46 on a large site fronting on Destrehan Street (Mason 1958a; Fleishman-Hilliard 

1967; Mallinckrodt 1994); it began operation in July 1946 and was then referred to as “the refinery” 

(AEC 1949b; AEC 1951b). Mason (1977) states that the choice of this site was based on proximity, 

security considerations, and space availability for constructing administrative and other support 

buildings that would enable the uranium operations to function as an independent unit. Most of the 

administrative offices, laboratories, and support facilities for the uranium refining operations were in 

fact located there. In early 1946, the ore-to-UO2 part of the refining process was moved to Building 

104 of Plant 6 from Plant 2 and apparently the laboratory work moved to Building 102 from Plant 1, 

while the UO2-to-metal production remained at Plant 4 (FUSRAP 2003c; Hickey and Dupree 1984; 

AEC 1949b; 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 

(1950c). Mason (1977) states that expansion of Plant 6 (new construction and production increases) 

continued through 1954.

In November 1947, Mallinckrodt cast 100 special billets for use by Hanford in an alpha extrusion 

experiment (AEC 1949d). These special billets contained higher nickel levels than usual and were 

prepared as part of the research into green salt production, in the belief that the new continuous green 

salt production reactor might give a product with higher levels of nickel than usual (AEC 1949f). From

late 1947 to early 1948, Mallinckrodt was engaged in producing natural uranium castings for the 

Schenectady zero power pile (AEC 1949d). Also from late 1947 to early 1948, several tanks and 

centrifugal filters were added to the digest process line to allow the reprocessing of the radium (K-65) 

and barium sulfate residues and recovery of 95% of the 2% of the original uranium from the ore that 

was in these two residues (AEC 1949d; AEC 1948g). The K-65 and barium sulfate residues already 

stored in the St. Louis area were then returned from storage and processed. Subsequently this socalled C-3 process was used on all the K-65 residues produced and the companion process on all the 

barium sulfate residues produced (AEC 1949d). 

The incoming ore arrived by rail and was stored in Building 110; however, in late 1950 or early 1951 

an outdoor onsite ore storage area was added for pitchblende ore (Q-11) (AEC 1950i). 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). The new refinery 

process was designed to include the new operations developed in the laboratory and pilot plant, i.e., 

to recover radium and raffinates as separate residues and to remove sulfate as a barium sulfate cake 

(Mason 1977). In 1949, following laboratory and pilot-scale studies, a second digest line was added

to Building 104 to process uranium ore (mill) concentrates (Mason 1977; Mallinckrodt 1994). This 

appears to have included a magnesia concentrate designated MgX or MGX and a calcium uranate 

concentrate designated CX, both of which were prepared in the Congo from low-grade tailings (AEC 

1951b; AEC 1949f). Building 114 was used for temporary storage of residues until they could be 

transferred elsewhere (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 1949b). Pitchblende 

ore use was phased out over the period 1952-1954, after which the pitchblende digest line was placed 

on standby (Mason 1977). Feed concentrates were used exclusively from 1955 through the end of 

production in 1958 (Hickey and Dupree 1984).

When equipment was added to Plant 7 to allow continuous UO3-to-UF4 conversion, Plant 6 continued 

to produce UO2 until early 1953 and then began to produce only UO3 (Hickey and Dupree 1984; 

Mallinckrodt 1994). Milling of UO3 at Plant 6, milling of ore at Plant 6, and pre-digestion ore grinding, 

conducted initially at Plant 4 and later at Plant 6, were discontinued in 1950, mid-1949, and 1955 
Document No. ORAUT-TKBS-0005 Revision No. 02 Effective Date: 06/14/2007 Page 18 of 225

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 had become clear to AEC that process improvements to Plants 4 and 6 were not enough to 

bring about satisfactory control of dust and other hazards (AEC 1949b; AEC 1951b). As Mallinckrodt 

reminded AEC, Plant 6 had been built as an ordinary processing plant with no special provisions for 

health hazards because it was not expected to be used more than eight months (AEC 1951b; MCW 

1950t). AEC authorized funding for dust control and mechanization improvements, which were 

installed in 1949-1950 (see Table A-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 (MCW 1950e; AEC 1951b). 

Mallinckrodt and AEC agreed to build a new pair of plants that were located at the Destrehan Street 

site (AEC 1949b; Mason 1958a). The first was Plant 6E, the new metal plant, which was constructed 

from early 1948 to June 1950, began startup operations in June 1950, and went into operation in July

1950 (AEC 1949e; AEC 1951b; Mason 1977). Note that there was a laborers’ and hod (a tray, trough 

or shallow box on a stick used to carry a load of bricks or mortar) carriers’ strike from 28 March into 

early April 1949 that slowed construction of Plant 6E (AEC 1949e; AEC 1949d). 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 (Mason 1977; Mallinckrodt 

1994). An expansion project was started in March 1951 and finished in about September 1951 (AEC 

1951b). 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 was constructed from November 1950 

to March 1951 and was turned over to Mallinckrodt for operation in March 1951 (AEC 1951b). Note 

that AEC (1949b) appears to indicate that Plant 7 operations began in late 1952, while Mason (1958a) 

indicates a 1948 start date and Mason (1977) indicates a 1951 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. The buildings involved at first were Buildings 703 through 708 (Mason 

1977). At Plant 7, a continuous process replaced the batch-type process used at Plant 4 and the 

UO3-to-UO2 production process at Plant 6, although UO2 could still be produced as needed (AEC 

1951b; Mason 1977); however, Mallinckrodt (1994) states that it was only later that equipment was 

added to allow the continuous production of UF4 from UO3 directly.

Uranium metal recovery and some storage operations were moved to Plant 7 in late 1951 (AEC 

1951a; AEC 1951b; Mallinckrodt 1994) or early 1952 (Mason 1958a). Some reversion of UF4 to UO2

or UO3 was done in 1953 and perhaps into 1954 (MCW 1953c). Building 700 was placed in service in 

1954 to serve as a warehouse (Mason 1977). A new wet slag (interim residue) recovery operation 

was added in late 1955 in Building 701 (Mason 1977; Mallinckrodt 1994); this was apparently what 

was referred to as Plant 7W and may have been a pilot plant (Hickey and Dupree 1984). Some time 

in 1954 (Mason 1977) or in the late 1950’s (Mallinckrodt 1994) machining of (nuclear) reactor core 

(elements) was done on a temporary basis in a fabrication facility in Building 700. In 1955, very low 

enrichment uranium (probably only a small amount) as UF4 was processed at Plant 7 (AEC 1955d) 

and in August 1956, about 5.5 kg of 20%-enriched uranium was processed, presumably in Plant 7 

(MCW 1956e).

Plant 7E, whose sole building (712) was constructed as a temporary facility in 1954-1955 and which 

was regarded administratively as part of Plant 7, was used from 1955-1957 to process pitchblende 
Document No. ORAUT-TKBS-0005 Revision No. 02 Effective Date: 06/14/2007 Page 19 of 225

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

solution (FUSRAP undated b, FUSRAP 2003a; Mason 1977 says from 1955-1956) by an acid 

digestion process similar to the uranium ore digestion (AEC 1955e). The concentrate was sent to the 

Mound site for further processing.

A new facility to replace the so-called “Ledoux Laboratory” was completed in November 1952 (MCW 

1953a). The purpose of this was to provide improved dust control for the sampling and processing of 

K-65, the radium-containing solid waste. Further improvements were made in 1953 (MCW 1953a). 

Other sampling and assay work was done, such as the evaluation of slugs produced by National Lead 

Company of Ohio from extruded rods that had been manufactured by the Bridgeport Brass Company 

from dingot billets apparently produced by Mallinckrodt (MCW 1956d). 

In 1957 or 1958, all uranium refining operations ceased at all the plants (FUSRAP 2003c; 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 (processing 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 already 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; Mason 1977, which says the work began in 1949); 

contaminated materials and some contaminated equipment were collected and delivered to AEC for 

disposal, while the rest of the contaminated equipment was transferred to Plant 6 (Mason 1977). 

After AEC was satisfied as to the completion status of the decontamination, AEC performed a final 

survey (Mason 1977), with presumably some touchups by Mallinckrodt and resurveys by AEC. In 

1951, AEC released these plants to Mallinckrodt for unrestricted use and no further AEC work was 

performed there (Mason 1977; DOE 1981). Mallinckrodt (1994) states that further decontamination 

took place in 1954 and 1970. Some of these buildings were still extant as late as 1990 (Applied 

Nuclear Safety 1991). Plants 4, 6, 6E, and 7 were surveyed by AEC and Mallinckrodt in 1958-1959 

and were decontaminated or taken down in 1959-1961 by Arch Wrecking Company (under contract to 

AEC), with safety and health coverage by Mallinckrodt (Mason 1977). Mallinckrodt performed a final 

decontamination survey for AEC during 1960-1961 (Mason 1977). The buildings removed during the 

takedown included all of the Plant 4 buildings and all wet-process buildings in Plant 6 (ORNL 1981; 

Mallinckrodt 1994) (see Table A-1 for details). The site was released to Mallinckrodt without 

restriction in 1962 (ORNL 1981). These decontaminations were done to the AEC standards then in 

force, i.e., not to background levels (ORNL 1981; DOE undated). The remaining Plants 6, 6E, and 7 

buildings were decontaminated to modern standards in the 1990’s and demolished in 1997 (FUSRAP 

2002).

Note that the term “Destrehan Street site” seems to have been used in documents 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. In distinguishing between these plants and the SLAPS 

airport site (see Section 3.2), these plants as a group will be referred to in this technical basis 

document as the Mallinckrodt St. Louis main site.
Document No. ORAUT-TKBS-0005 Revision No. 02 Effective Date: 06/14/2007 Page 20 of 225

To summarize, work performed at the Mallinckrodt St. Louis main site 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 (DOE 1997; Mallinckrodt 1994; MED 1946a)

2. Production of UF4, with some being shipped to Harshaw and K-25 (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 (DOE 1997; Mallinckrodt 1994; MED 

1946a; FUSRAP undated b)

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 or small-batch basis included the 

following.

1. Production of dingots using a thermite reduction process (Mallinckrodt 1994; FleishmanHilliard 1967; AEC 1956a) 

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

3. Casting of special billets (ingots) and other special uranium castings (AEC 1949d) 

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

Mallinckrodt 1994; AEC 1954g) 

5. Production of UO2F2 (FUSRAP undated b)

6. Reprocessing of pitchblende residues to recover uranium (AEC 1949d; AEC 1951b) 

7. Recycling of slag for use as liner material in the processing of uranium metal (AEC 

1956a; Fleishman-Hilliard 1967) 

8. Extraction and concentration of Th-230 from pitchblende raffinate (FUSRAP undated a; 

FUSRAP undated b; Mallinckrodt 1994; AEC 1955e) 

9. Experimental processing of very low enrichment UF4 (FUSRAP undated b; Mallinckrodt 

1994; ORNL 1981) and some 20% uranium metal (MCW 1956e) 

10. Production of small research quantities of uranyl nitrate hexahydrate (UNH) [AEC 

1951b]; conversion of small research quantities of aqueous uranyl nitrate hexahydrate to 

UO3 (DOE 2000)

11. Production of U3O8 and UO2 using an experimental continuous denitration furnace

(Mallinckrodt 1994)

12. Experimental extraction of uranium using tributyl phosphate (TBP) to replace ether 

(Mallinckrodt 1994), apparently in a pilot plant circa 1950-1951
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13. Miscellaneous analysis and assay work (e.g., MCW 1956d) 

FUSRAP (2003b) states that uranium ore assay was done in the “Ledoux Laboratory” located in 

Buildings 110A and 111 of Plant 6. However, from MCW (1954a) and other contemporary references, 

it appears that only the K-65 residue assays were done in the Ledoux Lab, which work had been 

contracted to Ledoux by African Metals (ORAU 1989a), which retained the rights to the radium in the 

residue. 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 (1944p) 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 

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.

Mallinckrodt processing of black oxide (pre-milled ore), soda salt (sodium diuranate, or Na2U2O7), 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 (uranium derbies), 

slag processing, ingots (billets), croppings, sawdust processing, and other metal production under 

Contract W7405-Eng 29, 2nd Phase (MED 1945a; MED 1944p).

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 

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 1949h); 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. 

3.2 HISTORY OF USE OF THE ST. LOUIS AIRPORT STORAGE SITE (SLAPS OR 

SLAPSS) 

The main supplier of the pitchblende ore to MED, the African Metals Company, retained ownership of 

the radium and valuable metals remaining in the residues after the ore was processed (AEC 1959). 

Thus by contract, MED was required to store the radium-bearing residue (K-65 or gangue lead cake) 

and the metal-rich barium sulfate cake residue until they could be transferred to African Metals (AEC 

1959). From 1942 to 1945, the residues and other solid wastes were stored temporarily at the 

Mallinckrodt site and then shipped to the Lake Ontario Ordnance Works. Because of the volume of 

the residues and the high gamma dose rate and radon emanation rate of the K-65 residue, MED 

obtained consent in March 1946 to use for residue storage purposes a 21.7-acre tract of land located 

in Robertson, Missouri north of the former St. Louis Municipal Airport (also known as Lambert Field) 

(AEC 1959). In January 1947, MED obtained title to the property by condemnation (AEC 1959). This 

storage area was formally called the St. Louis Airport Storage Site (SLAPS or SLAPSS), but in 
Document No. ORAUT-TKBS-0005 Revision No. 02 Effective Date: 06/14/2007 Page 22 of 225

Mallinckrodt and MED/AEC documents it was often referred to simply as “the airport.” The SLAPS 

site was operated by MED/AEC from 1946 until July 1953 (AEC 1959; ORNL 1979), when the 

operation was turned over to Mallinckrodt (AEC 1959); Mallinckrodt then ran the site under contract to 

AEC from July 1953 to 1958 and possibly until it was turned over to a private concern that assumed 

control of it under an AEC license (AEC 1959; ORAU 1989; AEC 1972). Although AEC owned 

SLAPS, Mallinckrodt workers appeared to have been doing the residue delivery and placement work 

even in 1946-1953 (MCW 1949d; MCW 1949p). 

Initially, it was only the radium-bearing waste residues (K-65), the barium sulfate residues (AJ-4), and 

the pitchblende raffinates (AM-7 or “airport cake”) produced at Mallinckrodt that were stored at SLAPS 

(Mallinckrodt 1994). Later, SLAPS was also used to store other types of residues and wastes and 

some ore and to serve as a storage and burial site for contaminated objects and debris (AEC 1959; 

AEC 1956d; Mallinckrodt 1994; AEC 1948e; AEC 1949m). The ore was stored there as part of AEC’s 

strategic stockpiling program (AEC 1949d). The other types of residues included the residue left after 

reworking of the AM-7, called AM-9; Colorado (domestic non-pitchblende) ore raffinate; and a residue 

formed from the precipitate of the first-phase ether extraction columns (“Sperry cake,” apparently a 

synonym for or subset of the AM-7) (AEC 1956e; AEC 1964; Mallinckrodt 1994). Metal scrap, 

dolomite slag (“C-liner” or “C-oxides”), and tailings of uranium scalping operations from magnesium 

fluoride slag (the latter two from the derby metal production) were also sent to SLAPS (Mallinckrodt 

1994). Some of the debris buried at the site appears to have been from the decontamination and 

decommissioning of the Plants 1 and 2 area in 1948-1950. From perhaps mid-1948 on, pitchblende 

ore (Q-11) was stored at SLAPS (MCW 1949p).

In 1948 and 1949, the K-65 residue was brought back from SLAPS in lots to be “reworked” 

(reprocessed to recover more uranium) (AEC 1959), after which it was sent to the Lake Ontario 

Ordnance Works (AEC 1959). The drums in which it had been stored were returned to SLAPS for 

storage. From 1955-1957, most of the AM-7 was brought back from SLAPS to be processed to 

recover Th-230 as a solution that was sent to Mound; the resulting residue, AM-9, was sent back to

SLAPS (AEC 1959). Also in 1955, the tailings (D-701) from the processing of the magnesium fluoride 

slag produced in the metal plant were stored at the site (AEC 1959) or at SLAPS (AEC 1956d; 

Mallinckrodt). The barium sulfate cake remained at SLAPS because eventually African Metals 

relinquished ownership of it (AEC 1959). 

In 1952, some of the western part of the SLAPS plot, where contaminated metal and debris were

buried (presumably from the decontamination of Plants 1 and 2), was filled in deeply with earth (AEC 

1964). In April 1959, a railroad siding and loading facilities were installed (AEC 1959), apparently in 

preparation for moving out the residues and other material (AEC 1959); the tailings from the 

magnesium fluoride slag processing were then sent to the Fernald site (AEC 1959). In about 1959-

1960, some of the scrap and rubble from the demolition and decontamination of the Mallinckrodt St. 

Louis site facilities was buried at the west end of the site (AEC 1972; ORNL 1979). At some point, a 

building consisting of a changeroom, a shower, and office space was moved to the site (AEC 1959; 

AEC 1964); it is not clear when this occurred, but since earlier references mention only a truck wash 

pad as a cleanup area, the building was likely not installed until the burial activities of 1959-1960 took 

place.

In March 1962, AEC issued an invitation to bid on the residues and on September 1962, AEC issued 

a license to the successful bidder, a commercial entity, to possess and remove 125,000 tons of 

uranium and thorium residues stored at SLAPS and at the licensee’s processing facility (NRC 2004). 

Although there was little work by Mallinckrodt or SLAPS employees at the site thereafter, some further 

history of the site is given below.
Document No. ORAUT-TKBS-0005 Revision No. 02 Effective Date: 06/14/2007 Page 23 of 225

The 1962 license was terminated in February 1966. This was possibly due to the licensee’s going out 

of business (AEC 1967). A new license was issued to a new purchaser in 1966. In 1966-1967, this 

purchaser moved the residues from SLAPS (AEC 1967, FUSRAP 2004a) to a site at Latty Avenue in 

Hazelwood, Missouri (AEC 1967); this was said to have required ten dump trucks operating for five 

months to move a total of 100,000 tons of residue (AEC 1967). 

AEC apparently proposed to the City of St. Louis that the city take over the site, specifically the cityowned airport management organization. Thus in 1969-1970, under a November 1969 acquisition 

permit from AEC and with some radiological support from the City of St. Louis Health Department, the 

St. Louis (Lambert) Airport Authority demolished the structures on the site and excavated some of the 

contaminated earth (AEC 1969; AEC 1972; ORNL 1979; FUSRAP 2004a). Excavated material, 

including some residual barium sulfate residues, was trucked to the AEC quarry at Weldon Spring

(AEC 1969; AEC 1972), while the demolished structures, including the concrete floors and pads, were 

buried at the site and clean fill was placed over the entire site (AEC 1972). More fill was placed over 

several areas following a radiation survey performed in December 1969 to meet the requirements of 

the permit (ORNL 1979). In November 1971, AEC representatives performed a satisfactory final 

radiological survey of the site (AEC 1972; ORNL 1979). 

In 1973, AEC relinquished ownership of the site to the St. Louis Airport Authority by quit claim deed 

(FUSRAP 2004a). After that, access was controlled by the St. Louis Airport Authority and casual 

entry was precluded (ORNL 1979). Over the period from about 1972 through 1978, the site served as 

an isolation or buffer space on the side of a runway approach at the adjacent airport and it was also 

used as a dump site for nonradioactive rubble from the airport; the fence remained in place, but the 

area was not patrolled (AEC 1979). Without maintenance, the site deteriorated, in that erosion 

occurred and minor contamination was transported to the drainage ditch that surrounded the site 

outside the fence and to the creek that traversed the lower end of the site (AEC 1979). This was 

shown in the radiological characterization surveys performed by Oak Ridge National Laboratory 

(ORNL) for AEC in November 1976 and August 1978 (ORNL 1979; AEC 1979). On the basis of these 

surveys, it was decided that remedial action was required (AEC 1979; FUSRAP 2004a) and in 1981 

the drainage ditches along the north side of the site were designated for remedial action under 

FUSRAP (2004a). 

The residues that had been sold and taken out of the site went to the Latty Avenue (St. Louis) site to 

await processing by the purchaser; however, the purchaser went bankrupt before processing actually 

began. Under a Nuclear Regulatory Commission (NRC) license, the Cotter Corporation then dried the 

residues in a kiln at the Latty Avenue site and sent them off to their Colorado facility. However, the 

Latty Avenue site was not properly decontaminated before the NRC license was terminated and the 

site was sold to a private purchaser. After an ORNL survey at the Latty Avenue site, the owner was

advised by NRC to stop property development and the NRC asked the Department of Energy (DOE) 

to allow materials from the future decontamination of the site to be taken to the Weldon Spring Plant

(AEC 1979). However, DOE considered the Weldon Spring raffinate pit to be unsuitable for 

permanent disposal of these materials. NRC then suggested that the material be returned to SLAPS

and DOE agreed, providing that this could be done in an environmentally acceptable manner (AEC 

1979). 

Eventually, the US Congress, via the Energy and Water Development Appropriate Act of 1984 (Public 

Law 98-360), directed DOE to reacquire and stabilize the SLAPS site and to use it as a storage and 

disposal site for the Latty Avenue materials (Bechtel 1987; DOE 1986). Although it postponed 

acquisition of the site even after being directed to do so by Congress, DOE had begun to have routine 

monitoring of SLAPS done by Bechtel National Incorporated in 1983 and as of 1987 was planning to 

dispose of the waste materials from the Hazelwood Interim Storage Site (including the Latty Avenue 
Document No. ORAUT-TKBS-0005 Revision No. 02 Effective Date: 06/14/2007 Page 24 of 225

materials) at SLAPS eventually (Bechtel 1987). In October 1989, the Environmental Protection 

Agency (EPA) placed SLAPS on the National Priorities List, which meant that the cleanup would 

proceed according to the guidelines of the Comprehensive Environmental Response, Compensation, 

and Liability Act (CERCLA) (FUSRAP 2004a). DOE then began to work with EPA to determine how 

to clean up the site and where to put the contaminated soil (FUSRAP 2004a). The two agencies 

signed an agreement establishing an environmental review process and remediation schedule, with 

one of the options to be considered being storage of the SLAPS and HISS materials at SLAPS; DOE 

declined to accept SLAPS back from the City of St. Louis until the CERCLA review process was 

completed (FUSRAP 2004a).

In 1990, the St. Louis Board of Aldermen approved a plan to transfer SLAPS to DOE on condition that 

a permanent disposal cell for radioactive wastes would not be constructed on the site (FUSRAP 

2004a). In 1994, DOE formulated a remediation plan and presented it to the St. Louis community. 

The community rejected it because it included a disposal cell at SLAPS, contrary to the terms of 

transfer enacted by the board of aldermen; the site was therefore never transferred to DOE and it 

appears unlikely that it will ever be (FUSRAP 2004b). However, as of July 2004, FUSRAP was 

working to develop a Record of Decision to document the final cleanup remediation plan for the North 

County sites, including SLAPS – a plan that did not include a disposal cell at SLAPS (FUSRAP 

2004b).

4.0

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 A-4; Table A-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, OTHER 

PROCESSES, AND SLAPS STORAGE AND DISPOSAL ACTIVITIES

It is also important to understand the nature of the work performed at SLAPS in order, again, 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 discussion of SLAPS work is included in 

the discussion of Mallinckrodt-produced residues (Section 4.7 below). 

4.1 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 in nitric acid in a digestion vessel or tank (MED 1946a; 

AEC 1949b; AEC 1967; Fleishman-Hilliard 1967). Sulfuric acid was added to the solution in the 

vessel to precipitate the radium and lead as sulfates (MED 1946a; AEC 1949b; AEC 1951b; AEC 

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

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

vacuum filter (AEC 1967) referred to as a Feinc because of the name of its manufacturer (Federal 

Engineers Incorporated or FE Inc). The filtered solids formed a radium-bearing residue referred to as 

the K-65 residue (AEC 1949b; AEC 1967) 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 1949b; AEC 1951b; AEC 1967; Fleishman-Hilliard 1967). To remove the solids, 
Document No. ORAUT-TKBS-0005 Revision No. 02 Effective Date: 06/14/2007 Page 25 of 225

the mixture was run in a continuous solid-bowl centrifuge (also referred to by the name of its 

manufacturer, Bird) (MED 1946a; AEC 1949b; AEC 1967). 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 to concentrate it (MED 1946a; AEC 1951b). 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 1967; 

Fleishman-Hilliard 1967).

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

extraction process, diethyl ether was added to the solution in extraction columns; the double 

extraction was to remove the acid-insoluble molybdenum (AEC 1951b). 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 re-extraction into 

water (MED 1946a; AEC 1949b; AEC 1951b; AEC 1967; Fleishman-Hilliard 1967). Eisenbud (1975) 

stated 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 1949b; 

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 1949b; 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 1944o).

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 1949b; Fleishman-Hilliard 

1967). Usually the UO2 was further processed at Mallinckrodt, but some was shipped to nonMallinckrodt 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 1949b). As needed, it was 

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

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

reactor (MED 1946a; AEC 1949b; 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 (HF) 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 1949b). Most of the UF4 was further processed at 

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

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

there. Excess HF was neutralized with lime or potassium hydroxide solution to recover uranium 

(Mason 1977).

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

powder and mixed in a tumbler (AEC 1949b; Fleishman-Hilliard 1967; Mason 1977). A special firing 

container (the “bomb”) was lined with a refractory material, variously reported to be lime (1945, per 
Document No. ORAUT-TKBS-0005 Revision No. 02 Effective Date: 06/14/2007 Page 26 of 225

MED 1946a); dolomite (from 1946 on, per FUSRAP undated a, or 1942 on, per Fleishman-Hilliard 

1967); or recyclable magnesium fluoride from about 1954 on (FUSRAP undated a; Fleishman-Hilliard 

1967). The bomb was “jolted” (shaken) on a mechanical jolter until the liner was packed sufficiently 

hard. 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 1949b; Fleishman-Hilliard 1967). The molten mass formed a 

“derby” or “biscuit” of solidified metal as it cooled (MED 1946a; AEC 1949b; Fleishman-Hilliard 1967). 

(The derby form was also called KB-2 and there were subtypes of derbies called Number 1, Number

2, etc.) After the bomb cooled, the derby was taken out by “breaking out” the bomb shell (Mason 

1977) and the slag on the derby was chipped off pneumatically, crushed, and sent to uranium salvage 

(recovery), first to Vitro and eventually to onsite recovery (MED 1946a; AEC 1949b; Fleishman-Hilliard 

1967; Mason 1977).

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

process 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 

1949b; 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 1949b; Fleishman-Hilliard 1967). A sample was taken from the billet by power hacksaw 

(AEC 1949b) and the billets were sawed to suitable lengths (presumably only when necessary) 

(Mason 1977). The billets were then packed, stored temporarily in Building 400 (Mason 1977), and 

shipped to the appropriate non-Mallinckrodt site for further processing (MED 1946a; AEC 1949b; 

Mason 1977). 

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

1949b; 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 1967).

The precipitate from the first extraction column was on occasion dewatered using a Sperry filter press, 

producing a supernate and a batch of cake called Sperry cake (AEC 1967). The supernate from the 

press and the aqueous uranium tails from the wash were de-etherized and treated with a hydrated 

lime slurry (AEC 1967). 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 AM-7, called “airport cake” (AEC 1967), from its being stored for several years at the 

AEC’s SLAPS site that had been an airfield. Note that the Sperry cake appears to be a subset of the 

AM-7 or perhaps synonymous with it. Also, several other types of cake (e.g., the barium sulfate cake) 

were also referred to loosely as “airport cake” due to their being sent there for storage.

4.2 THE WARTIME PERIOD (APRIL 1942 - APRIL 1945)

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 (DOE 1997; AEC 1967; 

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

Plant 4, during the same month (Fleishman-Hilliard 1967). Mason (1977) stated that there were no 

wet processes in Plant 4 (ever).

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). The ore had to be ground to “four mesh” 

size for optimum digestion as well, if the ore particle size was larger (MCW 1947). Little information is 

available as to how or where the early milling and grinding operations were performed, but it was 

apparently in Plant 2 (see Table A-1), later in Plant 4 (AEC 1945f; AEC 1945h; AEC 1945i), and still 

later in Plant 6 (e.g., MCW 1950e). Some soda salt was also used as feed material (DOE 1997). 

Note that the radium-lead precipitation step was necessary for the high-radium pitchblende ores, but 

not necessarily for other types of ore.

Plant 4 began production-level UO2-to-UF4 conversion in April 1943 and production-level UF4-to-metal 

production in July 1943. The ore→UO3→UO2 conversion continued to be done in Plant 2. MED 

(1942) stated that the ether extraction was done by a completely enclosed process of “recirculating” 

and that a solution of purified uranyl nitrate was boiled in large open receptacles to decompose the 

nitrate and produce the UO3. The UO3 was then ground and sifted to yield the appropriate size for 

“roasting” in a furnace to produce the UO2. 

AEC (AEC 1951b) stated the following. The process of refining pitchblende ore by using nitric acid, 

including the radium removal technique and acid extraction were developed by Yale University but 

expanded to production scale by Mallinckrodt during this period; also, double extraction to remove 

molybdenum when pitchblende ores were processed and the balance of the processing to UO3 were 

developed by Mallinckrodt. The UO2 hydrofluorination process was developed by DuPont and 

Harshaw, while the process for reduction of UF4 to uranium metal was developed by Iowa State 

College at Ames. The research for the continuous UO3-to-UF4 process was done by Mallinckrodt.

4.3 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 

4 (AEC 1945f; AEC 1945h; AEC 1945i) and then to Plant 6 after the latter began operation in 1946. 

Because of the dust hazards of the pitchblende ore, AEC had Mallinckrodt develop a method to permit 

automatic unloading and loading of the pitchblende ore in the grinding operation (AEC 1948f). 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 after that apparently still had to be ground to some 

extent (e.g., AEC 1947a). 

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). The sulfur removal (barium 

sulfate precipitation) step was added when ores containing high levels of sulfur were used. The 

continuous countercurrent process eliminated the necessity of moving the intermediate materials 

batchwise from one process location to another as had been necessary at Plants 1 and 2; rather, the 

materials passed from vessel to vessel or filter via piping. There was apparently still some possibility 

of “splatter or slopping over” at some point if the chemical reaction was too violent, but this would not 
Document No. ORAUT-TKBS-0005 Revision No. 02 Effective Date: 06/14/2007 Page 28 of 225

be true of the column extraction process because it was completely enclosed (AEC 1949e). 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 

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. Also, during the first year of operation (1946-

1947), it had become apparent that not all of the uranium in the pitchblende ore was being dissolved 

in the nitric acid step, but was remaining in the “K-65” or gangue sludge residue that was to be stored 

and returned to African Metals Company; a process was developed and installed in 1948-1949 to 

allow greater capture of the uranium from the ore (see Section 4.7). 

It was realized that the UO2-to-UF4 production in Plant 4 was particularly problematic from a dust point 

of view and AEC therefore had Mallinckrodt embark on an experimental program to produce a 

continuous-process reactor for the production of UF4 (AEC 1948f). 

4.4 THE LATER POSTWAR PERIOD (1950 - 1958)

Milled ore was sent from Middlesex to Mallinckrodt up to 1955, when the Middlesex work was 

transferred to the Fernald plant, which thereafter was the sender of milled ore to Mallinckrodt (DOE 

1997). An Ore Room and addition were added by 1950 to Plant 6. Early on, there was grinding of ore 

in a rod mill and opening, dumping, cleaning, and closing of drums (MCW 1950u); later, the duties of 

the operator included opening (deheading) and cleaning ore drums (AEC 1953). Hence this was not 

simply a storage area. During 1950, the average concentration of raw materials arriving at 

Mallinckrodt dropped to less than 50% (as U3O8), compared to ~65% during 1947-1949 (AEC 1951b). 

This was due to a gradual decrease in the content of the pitchblende and to the introduction of the 

magnesia concentrate MgX, which was produced from low-grade tailings. The new feeds had higher 

impurity content and the MgX contained 25-30% moisture compared to the normal pitchblende 

content of less than 3%. In order to maintain the production level, a new extraction column was 

designed and installed by Mallinckrodt in December 1950 to permit direct solvent extraction of the 

MgX slurry without prior filtration.

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 A-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. In late 1950 or 1951, in order to increase production, Plant 6E was placed in six-day, three-shift 

operation as quickly as new crews could be trained (AEC 1951b), while Plant 7 was designed to 

operate on a six-day, three-shift schedule.

AEC (AEC 1951b) stated the following. In 1951, the Plant 4 UO2-to-UF4 conversion equipment 

consisted of six large gas-fired furnaces containing five reactor banks each. The individual reactors 

were composed of a graphite tray assembly inside of a steel shell. The Plant 4 UF4-to-biscuit 

equipment included six individual gas-fired furnaces, and the biscuit-to-ingot conversion equipment 

included four high-vacuum induction furnaces. The Plant 6E UF4-to-biscuit equipment included 16 
Document No. ORAUT-TKBS-0005 Revision No. 02 Effective Date: 06/14/2007 Page 29 of 225

electrically heated induction furnaces and the biscuit-to-ingot equipment included six ultrahighvacuum furnaces. In Plant 6E, operations were mostly remotely controlled and/or totally enclosed, for 

health and safety reasons. In Plant 7, there were three parallel reactor banks, each made up of one 

UO3-UO2 reactor tube and three UO2-UF4 reactor tubes, with the four reactors making up the bank 

operated in series. Each reactor tube was equipped with a ribbon screw that moved the material 

continuously through the reactor into the feed area for the next tube. A single UO3-UO2 tube was 

installed to provide for the production of UO2 alone. The later expansion called for additional UO2-UF4

tubes to be added.

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 1956b). 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).

The Plant 6E operations became more enclosed and more mechanized over the years. In 1955, the 

billet (ingot) postprocessing grinding and inspection method in Plant 6E was described as follows 

(MCW 1955q); it is likely that this was the method used from the start but with the addition, over time, 

of the grinding enclosure for operation. The operator moved the billet from the water cooling tank into 

grinding position by means of a remote-controlled hoist. Through canvas gloves installed in the 

enclosure wall, he ground off the fin (raised line) from the mold joint. Using a remote-controlled 

piston, he then ejected the billet from the grinding rollers on the scale through a small opening in the 

enclosure. Using a rag, he wiped off the excess black oxide as he rotated the billet slowly, checking it 

for defects. He then moved the billet by hoist onto the billet cart to move it to the saw for a sample to 

be taken.

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

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

step instead of two, eliminating the need for recasting and the associated impurity contamination. 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 MgF2 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 applied. The bomb was removed by 

hoist to a “Hevi-duty” furnace, where the thermite (metal-metal reduction reaction) took place.

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, and 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 
Document No. ORAUT-TKBS-0005 Revision No. 02 Effective Date: 06/14/2007 Page 30 of 225

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, and 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 to have 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.

Feed forms and quantities (in tons) used for a period in1955.

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

Miscellaneous --- --- --- --- ---

Dissolver 24 77 20 20 20

Sawdust 5 6 5 5 5
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A list of process cells, dust collectors, and tanks appears in reference MCW 1955g. 

4.5 OTHER PROCESSES

Some uranium recovery operations consisted 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 (from the sawing of uranium metal billets) was 

stored under oil until it could be processed in this way and converted to ingots (Fleishman-Hilliard 

1967; MED 1944o; MED 1944p); the sawdust was also recovered by oxidizing it at high temperature 

to U3O8, and then sending it back to Plant 6 as a feed material (AEC 1949g). As mentioned earlier, 

some uranium was recovered from neutralizing excess HF in the UF4 production process (Mason 

1977).

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 on 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 (1956d), 

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” or “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.

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). See Sections 5.2.1-5.2.3 for more details. The ore areas and residue storage areas 

appeared to be separate from the product warehouse areas, however.

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

a short reference in an AEC Plant 7 air dust study report (AEC 1954g) to the work of a panel board 

operator’s work: 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 often the “C-” designated scrap material or waste 

material captured in a dust collector. Later such AEC reports also mention that the reverter was being 

used by this operator (AEC 1955d, AEC 1956b) but not mentioned in an earlier report (AEC 1952a), 

so presumably this work continued from at least 1953 (the date of the AEC 1954g survey) to the end 

of Plant 7 operation.
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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 1955e). 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 have been of a routine process nature. 

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

nitrate hexahydrate, both 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 1955d) 

regarding an 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, Fleishman-Hilliard (1967) 

commented that the reason for adding a little UO2F2 to a bomb was to slow down the process 

thermally, resulting in a better separation of slag and metal.

An experimental permit was issued in July 1956 that allowed the pickling of 14 tons of high-carbon 

scrap metal in Plant 6, which was done on 29-30 July 1956 without the health and safety department’s 

knowledge (MCW 1956f). It is not known if any more of this work was done.

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

constituted less than 0.3% 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.

Regarding specific laboratory sampling and assay processes, there is limited information available. 

This is given in Section 5.2.4.

4.6 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 (Na2U2O7), also called soda 

salt, some of which was sent to Mallinckrodt for refining to UO2 and UO3 (DOE 1997, MED 1945a; 

MED 1943e); other soda salt may have come from Belgium and Port Hope, Ontario (MED 1943e). 
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This appears to have been digested in the same fashion as the black oxide. Soda salt and other nonore 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 1956e; AEC 1959), 

when the plant switched to processing various types of concentrates. However, MED (1943e) 

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 1967), the so-called African ore. There is some 

disagreement as to how it came to St. Louis. DOE (1997) states that from 1943 on, the receiving and 

storage facility operated at Middlesex, New Jersey (DOE 1997) assayed, crushed, riffled, and 

redrummed the ore as it came into the United States, then sent it to the various refineries, including 

Mallinckrodt (AEC 1949b, Eisenbud 1975). MED (1945n) states that the African ore came through the 

Eldorado (Port Huron, Canada) area for processing before being sent on to the US refineries. AEC 

(1967) states that the concentrate was shipped from the Belgian Congo to St. Louis in 55-gallon 

drums (but this could mean that it came by way of Middlesex). Whatever the case, it was not likely 

that significant milling of this ore (as opposed to simply grinding to somewhat smaller size) was done 

at Mallinckrodt. In any case, it arrived by rail in carload lots and was stored in Building 110, the 

warehouse (Mason 1977).

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. The ore was usually stored in the 

warehouse or in another dedicated area on site, but presumably because of the volume of the 

pitchblende ore (Q-11) and its hazards, from about mid-1948 on at least some of it was stored at 

SLAPS (AEC 1949d) in a dedicated shed or shack (MCW 1949g; AEC 1949d, which states that ten 

25-ton lots were shipped to Mallinckrodt in May 1949 and an additional twelve lots were sent to 

SLAPS).

From 1955 on, ore concentrates were mostly used. This included some Belgian Congo ore tailings

concentrates (resulting from the leaching of raw ore during pre-processing in the Belgian Congo in 

previous years before the ore was shipped as a concentrate) (AEC 1951b). Soda salt appears to 
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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 1956b) and as noted Vitro 

also supplied soda salt until it was shut down.

4.7 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 1967); thus the 

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

From 1942-1945, MED solid wastes were stored temporarily at the Mallinckrodt site and then shipped 

to the Lake Ontario Ordnance Works in Lewiston, New York (Mallinckrodt 1994). The radium-bearing 

pitchblende residues (K-65 or gangue lead cake) were shipped by rail to the Middlesex, New Jersey 

repository for storage for a time, up to 1946 (AEC 1949b). But from the time AEC acquired the 

SLAPS site in 1946 and up through 1948, all residues produced at Mallinckrodt were stored 

temporarily in Building 114 of Plant 6 until they were shipped by truck to SLAPS (MCW 1949p; AEC 

1967; Mason 1977; Mallinckrodt 1994). From 1948 to at least 1949, the K-65 residues were shipped 

to Middlesex (AEC 1949b) or to the Lake Ontario Ordnance Works (AEC 1959) after being drawn 

back from SLAPS and reprocessed. After about 1949, K-65 was shipped to the Lake Ontario 

Ordnance Works (FUSRAP undated a; AEC 1959; AEC 1949b); although Mallinckrodt (1994) says 

that these were still sent to SLAPS up to 1956, it was clearly only the other types of residues and 

wastes that were sent after 1949, not the K-65. Some uranium scrap, including the mainly 

magnesium fluoride slag immediately above the derby (in the metal reduction step) was sent to the 

Vitro Manufacturing Company for uranium recovery (AEC 1949b).

Initially, it was only the K-65, the barium sulfate residues (AJ-4 or “airport cake”), and the pitchblende 

raffinates (AM-7) produced at Mallinckrodt that were stored at SLAPS (Mallinckrodt 1994). The K-65 

was trucked in drums (AEC 1947a), while the barium sulfate (AJ-4) and pitchblende raffinate (AM-7) 

cakes were collected in dumpster-type containers, loaded into AEC dump trucks, and transported on 

a daily basis to SLAPS (Mallinckrodt 1994). The K-65 drums were stored in a dedicated shed 

structure (AEC 1947a), but the other residues were simply dumped on the ground in piles (AEC 

1947f; AEC 1947a; AEC 1947e; AEC 1949b) because they were insoluble in water and were thus 

regarded as fairly immobile (AEC 1947a). 

Later, SLAPS was also used to store other types of residues and wastes and some ore and to serve 

as a storage and burial site for contaminated objects and debris, such as empty drums, Rauschig 

rings from the ether extraction columns, bituminous floor materials, and some materials from the 

decontamination and decommissioning of Plants 1 and 2 (AEC 1959; AEC 1956d; Mallinckrodt 1994; 

AEC 1949m). As noted earlier, the other types of residues included the AM-9 residue left after 

reworking of the AM-7 to recover thorium; Colorado (domestic carnotite and other low-grade nonpitchblende) ore raffinate (AM-10); and the precipitate residue of the ether extraction columns (“Sperry 

cake,” apparently a subset of the AM-7) (AEC 1956e; AEC 1967; AEC 1964; Mallinckrodt 1994). 

There were also small amounts of other raffinates present at the SLAPS waste storage site from the 

processing of lower-grade uranium ores (AEC 1967). The material obtained from processing carnotite 

and other low-grade domestic ores was kept separate from the K-65 (AEC 1967).

Metal plant waste, such as metal scrap, dolomite slag (“C-liner” or “C-oxides”), and tailings of uranium 

scalping operations from magnesium fluoride slag were also sent to SLAPS (Mallinckrodt 1994). The 

C-liner was mainly used and nonrecyclable dolomite liner with less than 2% uranium content 

(FUSRAP undated b; AEC 1949b) and 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 
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the dolomite liner was replaced by a recycled magnesium fluoride liner (FUSRAP undated b). There 

was another bomb waste called “C-special” (AEC 1949b), 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).

Within the Mallinckrodt plants, 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 1944o). Floor sweepings 

and spillages, swept into sumps; scrapings off the walls and equipment; and dust collected in 

collectors, usually by vacuum systems (MED 1944o) were also sent to a recovery system. 

Some reprocessing of residues was also done to recover uranium. AEC (1951b) states that following 

the development and installation of a sodium carbonate/sodium bicarbonate leach process in 1948-

1949 to recover more of the uranium, the K-65 produced up to that point was brought out of storage at 

SLAPS in lots and reprocessed in about 1949 (MCW 1949o and AEC (1959) imply 1948-1949), 

resulting in a final uranium content in the residue of 0.05%. This was apparently done in the same 

vessels and in the same general manner as ore would be processed, including heating the K-65 

drums in the thawing oven used to thaw ore drums (MCW 1949o). The new residue was sent to the 

Lake Ontario Ordnance Works (AEC 1959) to await return to African Metals and the corroded drums 

in which the K-65 had been stored were returned to SLAPS for storage (AEC 1949m). At least some 

of the barium sulfate cake was reprocessed similarly (AEC 1951b; AEC 1948g); however, the 

resulting barium sulfate residue was taken back to SLAPS because (eventually) African Metals 

relinquished ownership of it (AEC 1959). Some of the Sperry Cake, a good source of protactinium231, was brought back from SLAPS and shipped to Mound, which processed approximately 20 tons 

(about eighty 55-gallon drums) of it and obtained approximately two grams of protactinium-231 (AEC 

1967).

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 1967; 

FUSRAP undated b). Starting in 1955, the thorium-bearing AM-7 raffinate residue was brought back 

from storage at SLAPS, stored at Plant 6, and conveyed by dumpster from Plant 6 to Plant 7E as 

needed (AEC 1955b). The processing was first done on a crash basis in early 1955, apparently on a 

laboratory level (AEC 1955c; ORAU 1991), to produce a solution containing several hundred grams of 

thorium. Mallinckrodt later undertook the process on a pilot scale, but insisted on only a limited 

number of workers and strict safety precautions (ORAU 1991). The resulting residue, AM-9, was sent 

back to SLAPS (AEC 1959).

AEC (1955b) described the thorium solution extraction process as follows. The AM-7 was first 

digested in nitric acid in a tank and filtered. The resulting “liquor” was then processed in a TBP 

(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 (apparently in solution). The thorium fluoride was sent to the Hot Lab in Plant 6, 

where it 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-9, 

was sent back to storage at SLAPS. From the beginning of 1955 into 1957, a total of 350 tons of the 
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AM-7 was processed (AEC 1959; AEC 1967). ORAU (1991) says the processing was performed in 

1958). Note that the Th-230 was also called “ionium” and was referred to that way in Mallinckrodt 

records. 

As noted earlier, significant earthmoving operations to cover buried contaminated items were done in 

1952 (AEC 1964); the tailings from the magnesium fluoride slag processing were sent to the Fernald 

site in 1959 (AEC 1959); and in 1959-1960, some of the scrap and rubble from the demolition and 

decontamination of the Mallinckrodt St. Louis site facilities was buried at the site (AEC 1972; ORNL 

1979). 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 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 1949b). A small additional 

amount of acid waste flowed to the Mississippi via an acid sewer (AEC 1949b); Mason (1977) states 

that excess HF (from UF4 production) was neutralized with lime or potassium hydroxide and sent to 

process sewers, so it is unclear what the acid referred to by AEC (1949b) might have been. 

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 A-1 Notes column).

5.0

As AEC (1950a) observed, radiation measurements or evaluations of dust exposure for the purpose 

of dose determination were not 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. Thus only a few instantaneous 

measurements were made and time-weighting was generally not considered. 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 radium-bearing pitchblende ores), they determined them 

to be “considerable” with respect to the long term (AEC 1950a). NYOO and Mallinckrodt began a 

program of workplace and personnel monitoring.

RADIOLOGICAL CHARACTERISTICS, CONDITIONS, CONSIDERATIONS, AND 

AVAILABLE DATA

MED 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 1945 and a

formal dust measurement program in 1948 (AEC 1950a). Although urinalysis for uranium was done 

as a “screening experiment” in acute and chronic exposures as early as 1944 (MED 1944j), a routine 

urinalysis program appears to have begun in 1947 or possibly 1948 (MCW 1950c) (see Section 

5.3.6).

Some extrapolation of existing data to cover unmonitored periods is necessary, as AEC itself tried to 

do (AEC 1950a). The sections below provide information as to the available data and other 

information to address dose reconstruction. 

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

reduce exposures (AEC 1949b). However, with the increase in production, these goals were not met 

in all cases, although there were some successes and 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.

An AEC effort in evaluating dose to workers is documented in AEC 1950a, which is a report of an 

attempt by AEC in late 1949 to estimate the cumulative exposures of Mallinckrodt workers at Plants 4 

and 6 who were working at any time between July 1942 and 1 October 1949 and who had more than 

six months of exposure to radioactive materials. The base data came mostly from Mallinckrodt 

through its health physics staff, who did some calculations and sent them to AEC (e.g., MCW 1950c). 

This study is important because it was apparently AEC’s first large-scale study of chronic exposure to 

all types of workers; ORAU (1977) commented that this was the best information available about 

individual employment dates, jobs, and dust concentration during the period since more than 90% of 

all full-time uranium workers were included; also, the exposures were chronic and potentially sizeable 

-- production was in thousands of tons per year and the Mallinckrodt main facilities were “factories, not 

laboratories” (ORAU 1977). Of the 650 workers included in the study, 99 were working in Plant 4, 294 

were working in Plant 6, and about 250 had previously terminated (ORAU 1977). 

However, the potential usefulness of this data is lessened by the fact that the Plant 4 work records 

were somewhat deficient and Mallinckrodt was thus not able to evaluate past dust exposures for 

workers who had terminated employment or who had transferred out of the AEC project work prior to 

October 1949 (MCW 1950c). Also, some of the estimates were difficult for Mallinckrodt to do because 

the particular type of work was variable (e.g., maintenance work) (MCW 1950c). Mallinckrodt used 

the data of 1948 to calculate exposures prior to 1948 although they thought that many of the 

exposures were likely somewhat higher (MCW 1950c). Finally, since reported dust concentrations 

were based on a gross alpha measurement rather than on a chemical (element-specific) basis, the 

reported concentrations and the doses calculated from them might not always indicate a strictly 

uranium exposure; it was known that some workers, such as those working with K-65 residue, were 

exposed to radium dust (beyond that normally present in uranium dust as per uranium-radium parentdaughter ratio) (MCW 1950c). Still, correlation of the urinary uranium concentrations with calculated 

dust exposures was considered to be good for Plant 6, although poor for Plant 4 (MCW 1950c).

AEC’s estimates for the dose to the lung were based on air samples of alpha-emitting dusts 

(translated to a daily weighted average exposure level); to the bone, on breath radon analysis (to 

determine the fixed radium burden); and to various organs, on film badge data (AEC 1950a). ORAU 

(1977) stated that there was a Mallinckrodt lifetime tolerance dose exposure level, but did not state 

what the level was; however, it did report that by October 1949, 52 employees had accumulated 

100% of this level, with 45 exceeding 100% of it and 10 exceeding 200% of it.

Plant 6 was the most hazardous area in terms of dust levels, radon levels, and external gamma dose. 

Both Mallinckrodt and AEC made periodic studies of the various areas and continuous incremental 

improvements were made over the years of operation. An AEC consultant noted in 1949 that film 

badge records were well kept and the external exposures could be seen to correlate with the amount 

of radium in the plant (in ores and residues); despite problems with airborne radioactivity, progress 

had been made in reducing external dose and airborne dust levels (AEC 1949k). However, in 

October 1955, a Mallinckrodt health and safety official was asking an operations manager what had 

happened to Plant 6, listing in detail the many areas that were dirty and had leaking or nonfunctional 

equipment and the many questionable practices and violations of requirements (e.g., doors left open 

and respirators not being worn) that were resulting in unnecessary exposures to workers, particularly 

dust exposures (MCW 1955d). This official stated that many of these items had been pointed out 
Document No. ORAUT-TKBS-0005 Revision No. 02 Effective Date: 06/14/2007 Page 38 of 225

repeatedly, to no avail. Thus although the various hazards were progressively identified and 

addressed over the course of the years of operation and exposure levels did go down significantly, in 

the last few years some hazards were allowed to continue for extended periods before being 

addressed and some exposures were seen to go up. Even so, it appears that some studies and 

improvements were made very late in the life of the plant (e.g., the cleanup following the January 

1957 study of radon and gamma levels in the Cloth Storage Room (MCW 1957)). Because of this 

somewhat up and down safety history, it is important to consider what was going on at the plants in 

any given year in order to assess the dose to an individual.

5.1 UNITS, LIMITS, AND RECOMMENDATIONS

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

milliroentgen (mR) for gamma doses and millirep for beta doses, with the rep being equal to 0.93 rad; 

the abbreviations in the film badge and other records were mr and mrep respectively. Air exposures 

were expressed as disintegrations per minute per cubic meter sampled or inhaled (dpm/m3

), while 

radon exposures were usually given in units of 10 x 10-10 Ci/L. AEC also expressed exposures in 

terms of “tolerance” levels set on the basis of assumed safe levels; a tolerance level was equal to a 

given external dose level in mR per week or mrep per week, or radioactive air concentration in

dpm/m3

, etc., but the given values changed with time as MED/AEC set the safe levels lower and 

lower. The tolerance was regarded as a recommendation rather than a rigid limit, but AEC pressed 

the contractors to get and keep exposures to below the tolerance level. The units and the tolerances 

are further explained below.

In the early days of the war, MED decided that its tolerance levels of external exposure for uranium 

processing would be 0.1 R per eight-hour day or 0.6 R per week for gammas and 0.5 R per eight-hour 

day or 3.5 R per week for betas (MED 1943c, MED 1945b). This was based on the findings of the 

University of Chicago that the gamma dose rate from an infinite plane source (a 2π geometry) of 

uranium metal was 0.04 R per eight-hour day and for a room of uranium metal (a 4π geometry) it was 

0.09 R per eight-hour day (MED 1943c). Later in the war, the limits were set at 700 mR per week to 

the whole body and 3500 mrep per week to the hands (MED 1945b, AEC 1949b; AEC 1950b). MED 

(1945b) stated that the entire dose for one week of either type of radiation could be received in one 

day or less; AEC (1950b) stated that the limit 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). 

MCW (MCW 1950f) states that the tolerance level as 12.5 mR/hr, i.e., the hourly average for the 0.1R 

per eight-hour day; presumably this was according to AEC usage as well.

At some point, when NYOO had assumed the job of reading the film badges, the tolerance level 

appeared to have been lowered again 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 factor of 0.93. In 1947, the basic dose limit was said to be 0.1 rep/day and the 

relative biological effectiveness (RBE) for alphas to be 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 
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1949b). An AEC internal memo dated January 1949 (AEC 1949c) states that the Plant 6E design by 

Singmaster & Breyer was intended to meet the following AEC-specified criteria: whole-body beta dose 

less than 250 mrep per week (based on an maximum permissible (tolerance) level of 500 mrep per 

week) and beta dose to the hand less than 500 mrep/week (based on a maximum permissible level of 

1500 mrep per week). It was noted that AEC considered these criteria to be the minimum 

requirements and did not include any factors of safety beyond the “acceptable” (i.e., tolerance levels). 

New shielding added in 1948-1949 was designed on the basis of 50 mR per day of gamma radiation 

(MCW 1950e; AEC 1947b).

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 a 1950 letter to AEC, Mallinckrodt summarized AEC’s tolerance levels in the postwar era, as given 

in the table below (MCW 1950t). Mallinckrodt (MCW 1950I) suggested that AEC use absolute units 

instead of tolerance units to report exposures because the tolerance values were apt to change with 

time and this could lead to confusion in interpreting older records.

Summary of AEC tolerance levels in the postwar years.

Date

Beta tolerance 

level, mrep/week

Gamma tolerance 

level, mR/week

Total beta + gamma 

tolerance level, mrep/week

Before June 1948 700 700 1,400

June 1948 - August 1949 500 500 1,000

August 1949 - January 1950 300 300 600

January 1950 on --- --- 300

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 1950e); this was apparently based on the fact that 

Mallinckrodt was already using a system of personnel rotation to reduce doses (MCW 1949p; 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). Subsequently, 

Mallinckrodt 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).

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 1949b). 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 
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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” (also referred to as a tolerance level) of 50 μg/m3 for 

uranium dust, assuming a routine exposure of 8 hours a day, 6 days a week (AEC 1949b). This was 

taken to be equivalent to 70 dpm/m3 for alpha and was based on animal studies (Hursh 1975). 

Appendix B 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 (1955e) 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 1949 the AEC also recommended against respirator use except in emergency 

situations (AEC 1949b), suggesting that before the relevant period of Mallinckrodt work, extensive use 

of respirators was still tolerated as a means of minimizing exposure.

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). 

The “generally accepted level” for radium fixed in the body (as a body burden) was taken to be 0.1 μCi 

(MED 1946c). In the highest concentrations handled by Mallinckrodt (i.e., of the pitchblende ores), 

0.1 μCi corresponded to about 0.5 g of ore (MED 1946c). Since the radium was present in insoluble 

or only slightly soluble form, the original MED value for allowable dust concentrations of ore was 150 

μg/m3 (MED 1946c) (as mentioned above). When the tolerance for other kinds of dust was raised to 

500 μg/m3

, MED apparently did not mean for the ore and radium-bearing waste dust tolerance level to 

rise to 500 μg/m3 as well, but Mallinckrodt apparently assumed so, based on the statement in MED 

(1946b) that the maximum allowable concentration for chronic exposure for U3O8, UO2, and UF4 had 

been raised from 150 μg/m3 to 500 μg/m3

, effective June 1945. Thus in late 1946, a Mallinckrodt 

manager told MED that they were using 150 μg/m3 as the tolerance level for (apparently unground) 

ore dust, but 500 μg/m3 for U3O8, UO3, UO2, and UF4; MED (1946c) soon told them to return to using 

150 μg/m3 for uranium compounds and “5 x 10-11 grams radium per cubic meter [units approximately 

equivalent to μCi per cubic centimeter]” for ore residues with the uranium removed (for Ra-226 50

picogram/m3 is equal to about 110 dpm/m3

). As noted, above the uranium tolerance level was 

reduced to 50 μg/m3

, equivalent to 70 dpm/m3, by 1949; 70 dpm/m3 is also the figure given as the 

tolerance level for the radium-bearing residue in 1953 (MCW 1953b).

There was a separate tolerance level of 40 μg/m3 set for the “shotgun” residue dust (i.e., for sampling 

and work done in the shotgun laboratory), according to MED (1945d); this applied only if the

concentration of the Th-234 and Pa-234 decay products (called TI1 and TI2 in the reference) had not 

been increased by concentration to more than 400 times the equilibrium value (presumably secular 

equilibrium).

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 1949b) stated explicitly that the 

acceptable radon level for “environmental” air (room or outside air breathed by occupational workers) 
Document No. ORAUT-TKBS-0005 Revision No. 02 Effective Date: 06/14/2007 Page 41 of 225

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 1950k). 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 favorable to claimants 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 1949b). 

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 1949b). No information 

is available as to Mallinckrodt’s approach to effluent control, e.g., whether it followed the AEC 

recommendation.

For workers employed in years during which no (radiological) urinalysis was done, AEC (1950a)

prepared a cumulative exposure estimate study of Mallinckrodt workers who had been employed 

between July 1942 and October 1949. AEC’s estimates for the dose to the lung were based on air 

samples of alpha-emitting dusts (translated to a daily weighted average exposure level); dose to the 

bone was based on breath radon analysis (to determine the fixed radium burden); and dose to various 

organs was base on 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. It should be noted that workers were not identified by name or work category in the AEC report 

but the tabulations that produced the bottom-line figures for the report may be available by worker 

name in dose reconstruction project files.

Also with regard to the AEC 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 AEC did not include radon dose because they could not estimate an average 

concentration and they assumed that dose form airborne particle inhalation would dominate; that AEC 

assumed all the uranium to behave like UO2 in the lung; and that AEC 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). They stated that conditions probably were not more favorable and may have been 

“moderately” more severe. Thus they thought that their extrapolations could possibly be somewhat 

nonconservative.

5.2 RADIOACTIVITY CONTENT AND HANDLING OF THE ORE, URANIUM PRODUCTS, 

AND RESIDUES

5.2.1

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 

Ores and Other Feeds and the Initial Process Steps
Document No. ORAUT-TKBS-0005 Revision No. 02 Effective Date: 06/14/2007 Page 42 of 225

dose due to inhalation, with the radon releases again being highly correlated to the radium content. 

Since the concentration of radium and other daughters present in the ore, processed uranium, and 

processing residue at any given time depended most strongly on the concentration of uranium in the 

ore, on a per-ton-of-ore-processed basis (ignoring process differences), the various 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% (Eisenbud 1975) or 

30% (Mason 1977) uranium by weight, up to a maximum of 65-70% (DOE 1997; Dupree-Ellis et al. 

2000). 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). (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.)

Ores from the Belgian Congo had average concentrations of 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 1949b). 

(Mason (1977) gives 300 milligrams per ton, but this appears to be in error.) 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. A 1947 study done for AEC by the University of Rochester evaluated the relative amounts of 

uranium and radium in ore (see Section 5.3.1). The radon in the Belgian Congo ore 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 A-4 gives more information about the quantities and radiological characteristics of the 

ore.

As stated earlier, most of the pitchblende processed by Mallinckrodt was obtained as a concentrate 

from the Belgian Congo by AEC in 1944 and was shipped to St. Louis in 55-gallon drums from 

whatever North American receiving point was distributing it at a given time. 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.

In 1942, there was some contact (i.e., it was open to the air and potentially could be touched) with 

non-pitchblende uranium in the form of U3O8 when it was mixed with nitric acid for digestion (MED 

1942). As noted earlier in Section 3.1, the pilot plant operations in Building K-1E yielded a liquor that 

was conveyed by hand cart to vessels outside Building 52 and as suggested by Mason (1977), the 

further transfers also appear to have involved some closeup, manual handling. But when the 

pitchblende ore began to be used, contact was reduced as much as possible. To minimize personnel 

exposure, the ore (storage) room (addition) was divided by brick piers into corridors, with each 

corridor being wide enough to hold four drums of ore side by side, with a layer of four more drums 

stacked on top of them (MED 1946c). It was thought that this design would eliminate the need for

workers to pass between or close to stacks of ore during the ore storage and transfer operations. 

However, due to the blocking of some corridors and the filling of others because of the volume of ore 
Document No. ORAUT-TKBS-0005 Revision No. 02 Effective Date: 06/14/2007 Page 43 of 225

drums, it was after all necessary for workers to pass (close) by the ore drums stored in the partly filled 

corridors (MED 1946c). As of at least 1947, barrel handling of ore barrels was done with a forktruck, 

thus eliminating nearly all of the necessity of manual handling of the barrels (AEC 1947a). 

In 1948-1949, ventilation and remote control systems were installed in the Ore Room and its addition. 

This helped to reduce the radon concentrations in the operating areas, except for a point in the Ore 

Room at the junction of the two conveyor systems: radon levels were found to be high and to remain 

high for nearly half an hour after pitchblende drums were opened (MCW 1949m). Operators were not 

supposed to be near the drums after they were opened, but operators sometimes still had to enter the 

Ore Room or the conveyor alleys while drums were standing open on the conveyor because of 

operating difficulties. These difficulties included the need to make frequent visual inspections of the 

flow through the Syntron feeder mechanism into the rod mill, incorrect placement of a drum on the 

conveyor, failure of conveyor stops, and the jamming of some fiber (cardboard) drums on the 

conveyor. In the case of the second and third items, the operator had to crawl into the conveyor alley 

to remedy the problem and in the case of the fourth item, had to go into the Ore Room to remove the 

open drum manually.

In February 1950, a study was done of gamma and radon exposure to three workers conducting a 

special pitchblende ore inventory (MCW 1950g). This included moving 1.5 lots (probably a total of 9 

drums) by forktruck, weighing the ore drums, checking the numbers, and counting all the ore drums 

present (not just the checked drums). They received gamma doses of 45-75 mR, compared to the 

then-tolerance dose of 60 mR per day. Radon exposures averaged 1.02 x 10-10 Ci/L, which was 

higher than normal because the ventilation system was operated in a special mode to facilitate radon 

sampling for the study. It was concluded that these exposures were acceptable if the special 

inventory was done monthly, but if more extensive inventories were done, rotation of personnel and 

extra planning would be necessary.

In 1950, it was observed during a study that it was very dusty in the Ore Room proper and that while 

the rod mill (grinder) was running it was necessary for the operator to enter once for 1.5 minutes to go 

to the top of the skip hoist and dislodge the ore stuck in the supply drum; this was done using a 

hammer (AEC 1949e). Other than that, the operator’s time in the Ore Room was spent opening, 

dumping, cleaning, and closing the drums, with the average number handled being 22 per shift over 

about four days (MCW 1950u). In 1954, it was noted that operators were still being forced to enter 

the Ore Room frequently, due to the ineffective drum-handling system: drums would hang up on the 

conveyor and have to be pushed free manually (MCW 1954c). Also, large drums had to be removed 

from the skip hoist by hand because the system could not handle large drums (MCW 1954c).

Contamination control instructions were not always enforced in these known dusty operations. For 

example, in 1954 operators in the Ore Room were observed to be loosening lids on the ore drums on 

the conveyor before moving the drums to the skip hoist, which allowed dust to come out under the lids 

as the drums were moved; similarly, these operators took drums from the skip hoist and placed them 

on pallets with the lids positioned only loosely and without the clamps being fastened on (MCW 

1954c). Drums of ore were also observed to be sitting in the Scale House without lids on, with the lids 

and clamps being put on in the open area of the Scale House or even outside and not on the 

conveyor as procedure required (MCW 1954c). The reason for the irregularities with the drum lids 

and clamps was that by this time many of the drums, having been used and re-used, were old and 

slightly out of true, so that the lids had to be beaten on and the clamps were fastened on only with 

great difficulty (MCW 1954c). When the feeder (conveying) screws that removed the ore from the 

hopper became blocked, ore had to be scooped from the hopper by hand; it was noted at least once 

in 1954 that this was done without respirators (MCW 1954c). The dust released was not confined to 
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the ore area, since the doors to the Ore Room and Ore Room addition were often left open through 

negligence or for convenience (MCW 1954c).

Similar issues arose in the MgX area, used from 1949 on. Various memos and an August 1953 report 

(MCW 1953d) expressed concern about the MgX area, in particular about the inadequate design and 

maintenance of the conveying facilities that created dust levels in the breathing zone and general area 

up to 300-350 dpm per m3

. This report noted that the drum washer was not working properly, so that 

there were high levels of dust when the workers hammered the ring clamps onto the drum; the airflow 

into the exhaust on the operator’s platform was so inadequate that dust often came out of the opening 

when the operator released the mechanical “grab”; the operator had to go to the top frequently to 

push the feed material down the hopper into the screw feeder with a paddle, with dust coming out of 

the hopper opening as he did so. Also, the housing was warped, the gasket was loose, and there 

were broken windows in the drum washer housing (MCW 1953d). The problems had still not been

solved in 1954, when leakage was observed coming from the feed hopper to the tanks at the feeder 

screw housing because of flanges that were not securely fastened and loose packing at the screw 

shaft; the tops of the digestion tanks were thus covered with this dust (MCW 1954c).

In 1947, apparently after some months of experience in working with the high-level pitchblende ores, 

Mallinckrodt wanted to install steel shielding around several of the digestion tanks along the catwalks 

(platforms) and around the Feinc press filters that produced the K-65 waste (AEC 1949f). However, 

AEC rejected shielding around Tank M-14 because sludge was shoveled into the tank from the 

platform and the shielding would impede the shoveling (AEC 1949f). AEC apparently rejected all of 

the extra shielding eventually (MCW 1947). Mallinckrodt protested, stating that they had agreed to 

fewer provisions for health protection than they thought advisable in the design of Plant 6, on 

condition that the provisions would be added “if trouble later developed”; the difficulty was that after 

the plant was built and high-level pitchblende ores were being handled, AEC and Mallinckrodt 

disagreed as to what constituted “trouble developing” (MCW 1947). Thus although some shielding 

appears to have been added in 1948-1949 (AEC 1949b, MCW 1950e), notably for the ore grinding 

and C-3 filtration operations, the relatively high dose rates from the digestion tanks and the Feinc 

“bowls” were a continuing personnel protection problem at Mallinckrodt. This was noted on an 

inspection of the Mallinckrodt facilities by an AEC consultant, who stated that shielding was reaching 

a practical limit because with the tanks and vessels being shielded, the existing piping had become an 

important source of dose and the piping was not easily shielded (AEC 1949k). Further, rearranging 

the piping was not an option because the pitch and length of the sections was important to the 

process (AEC 1949k).

Dust and radon could be a continuing problem even in some of the “wet” areas. It is known that one 

reason for the spread of dust and radon was that doors to the cells containing the process equipment 

were often left open through negligence or for convenience; as late as 1954, a Mallinckrodt internal 

safety report gave as examples the leach and wash (Oliver) cells, the Feinc cells, and the M-14 cells 

(MCW 1954c). Tank lids were also left open unnecessarily, most notably the centrifuges, which were 

sometimes opened on one shift for cloth changing and not closed until the next shift (MCW 1954c); 

this allowed K-65 residue to dry out and drift out of the centrifuge and onto its outer surfaces and the 

floor. Also, as equipment became worn, there could be substantial leakage out various openings. In 

the case of the centrifuges, leakage was observed around the shaft, plow, and scoop openings at the 

top; this leakage was often allowed to collect and dry out (MCW 1954c).

From about mid-1948 on, at least some pitchblende ore (Q-11) was stored at SLAPS (AEC 1948e, 

MCW 1949p). Apparently at least one forktruck driver (to move the drums) and one truck driver were 

required; possibly a sampling worker was needed at times. Guards were said to have been 

“maintained” at the site from 1946 to 1951 (AEC 1959), apparently constantly; after that, it appears 
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that guards were present only for routine patrols and actual worker entry into the site for deliveries or 

removals of ore. It is not clear whether the storage workers and guards were always Mallinckrodt 

employees or not; e.g., Mason (1977) states the raffinate cake and barium sulfate cake collected in 

dumpsters in Plant 6 were taken away in dump trucks “which AEC operated.” However, Mallinckrodt 

warehouse workers and guards appeared to have been doing the residue delivery and placement 

work even in 1946-1953 (MCW 1949d; MCW 1949p), when AEC was running the site. There are 

notations of “airport” in urinalysis and film badge records, so it may be possible to determine from 

such records whether and when a particular individual did work at the site. 

Material was taken between the main Mallinckrodt site and SLAPS by truck (Mallinckrodt 1994; AEC 

1947a; MCW 1949p), so it can be inferred that ore drums were received by rail at Mallinckrodt and 

then trucked to SLAPS. The ore drums were stored on a concrete pad next to the K-65 shed (the pad 

also formed the floor of the shed) (AEC 1948j). Mallinckrodt (MCW 1949p) restricted the time that 

warehouse workers spent at SLAPS because of the hazards from the K-65 drums and because of 

other ore work at the main Mallinckrodt site, such as unloading ore drums from railcars. The task of 

obtaining ore drums from the airport pad involved three trips per week (MCW 1949p) from Plant 6 

(where the warehouse and yard workers were based) to SLAPS; apparently at least one forktruck 

driver (to move the drums) and one truck driver were required. Because of the proximity to the K-65 

storage in the enclosed shed, it appears that the K-65, with its much higher content of Ra-226 than 

the ore, may have contributed non-negligibly to the exposure of the ore workers.

It should be noted that in the later years, many booths or walk-in areas were called “enclosures.” 

These were generally not entered bodily by the operator, unless there was a malfunction of the 

equipment; rather, he inserted his arms into such enclosures through openings or used remotely 

operated devices to perform work within the enclosure. Some “stations” were actually enclosures 

such as this, some were hoods, and some were semi-enclosed booths.

5.2.2

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 in processing (except for the radium-bearing residue). Radium and radon would again build up 

to significant levels from the uranium parent, but this took more time than the apparent typical 

digestion-to-shipout time at Mallinckrodt. The principal hazard was thus the uranium-bearing dust.

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); in 1942, it was noted that contact could be 

made with UO3 when removing it from the open receptacles (boilers) to transfer it to the furnaces, and 

with UO2 when packing it. For example, in Plant 6 (and presumably in its predecessor Plants 1 and 

2), the UO3 dry powder was unloaded from the reaction pots by hand-scooping (Mason 1958a), i.e., 

manually using handheld scoops. The UO3 was scooped into a drum through a grated funnel that had 

a vacuum connection on it to draw the UO3 dust out of the drum (MCW 1956c) (the vacuum 

connection was probably in use only from the late 1940’s on). The large pieces caught in the grated 

funnel were broken up by hand, which was always recognized as being a particularly dusty step and a 

principal reason why respirators were supposed to be worn during the operation. In 1954, there was 

a temporary top that was supposed to be used during the filling of the UO3 drums to keep the dust 

down, but since it did not allow the operator to see when the drum was full, it was usually left off 

(MCW 1954c). The dust on the outside of the drums was supposed to be vacuumed off, but the 

operators often used a rag to do it (MCW 1954c), which meant that much of the dust went into the air 
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and onto the floor or the conveyor. When the dust collection bag for the UO3 area was changed, a 

spool piece was supposed to be installed in the ductwork, presumably to allow for an alternate suction 

source to keep the dust from coming out, but this was not done even through many bag changes 

(MCW 1954c).

The drums of UO3 were taken to the loading area and trays were filled with the material. After being 

weighed, the trays of UO3 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. This was 

described in Rochester 1948b as follows. In the tumbling area, the newly formed and cooled UO2 was 

placed in tumbling drums and rotated on a tumbler to break up clumps; the drums were taken to the 

skip hoist, where the tumbling drums were opened and transferred to the hopper of the Fiberpack 

loader; at the Fiberpack loader, a vibrating chute fed the material into 75-lb containers and the 

containers were sealed on a roller conveyor; and from the UO2 production area, the UO2 containers 

were trucked to the warehouse for storage.

The major handling improvement of 1949, the 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 1956b). A 1956 

Mallinckrodt report (MCW 1956c) explained why: when the pneumatic system was not operational and

the UO3 was ready to be unloaded from the pots, the unloading could not be postponed because acid 

would condense in the ducts and run into the pot, ruining the batch of UO3. So it was recommended 

that while the pneumatic system was out of order, the exposure be reduced somewhat by collecting 

the large pieces caught in the funnel grate and waiting until the pneumatic system was back in order 

before breaking up the large pieces (MCW 1956c). 

UO2 produced at Plant 6 was trucked over to Plant 4 in small fiber containers (AEC 1949b); 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). AEC (1947a) states that the UO2 was hand-scooped into trays and leveled off 

with a stick; the gloved hand thus came into direct contact with the material. Manual handling 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 Plant 4 and 

part of Plant 6. 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 1955d), this was not always so. AEC (1955e) 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 weight (in 

loading a container), with the operator’s (presumably gloved) fingers dipping into the material 

frequently.

In 1955, a new derby “pickling” facility went into operation in Plant 6. The derbies made in Plant 6E 

were transported to this facility to be pickled – i.e., cleaned in a dilute acid bath – and then returned to 

Plant 6E in baskets of eight or metal boxes of twelve (AEC 1955e, MCW 1955p). Prior to being 

processed in the recast step, the derbies were dried with a flame in a hood, if necessary (AEC 

1955e). Some chipping in the chipping enclosure was still necessary (AEC 1955e) because during 

the drying (whether by air or flame), the surface of the derby would oxidize to black oxide (U3O8). 

Thus the dust in the chipping operation was at least partly U3O8. Although the use of the flame for 

drying was supposed to be discontinued (MCW 1955o), it appears that it did continue (MCW 1955p). 

The billet (ingot) was initially low in beta activity because during the melting in the vacuum furnace 

(i.e., in vacuo), the dross and slag floated to the top of the crucible and the beta-emitting UX1 and 
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UX2 (Th-234 and Pa-234) sublimed and cooled on the underside of the furnace lid (Mason 1977). 

The dross from the crucibles and the other furnace residues was packaged separately from the slag 

and the billet (Mason 1977). However, as the billet aged, the UX1 and UX2 started to build up again.

In November 1953, a study was done of Plant 4 recast operation to determine the cause of high film 

badge readings (MCW 1953e); although Plant 6E made most of the metal, some metal was still 

produced on a pilot or experimental basis in Plant 4. The following dose-producing practices were 

found. The furnace jack was warped, so the operator had to enter the furnace enclosure to guide the 

jack with his foot each time the furnace was loaded. After unloading the furnace, the operator had to 

weigh the crucible and put it into a drum by hand. The crucible dose rate varied with the number of 

times used and the type of metal; the crucible in the study read almost 400 mrep/hr (beta plus 

gamma) before it was loaded because it had been used before (and would read more after the recast 

was done). Aged metal billets were stored on the floor beside the operator’s desk and were left there 

until the lead operator or the engineer examined and approved them, a period of one to three days. 

The operator spent four to six hours sawing “betatron slices” (samples to be sent to Granite City Steel 

(aka General Steel Castings) in Granite City, Illinois for testing) because the shape of the slices 

required the operator to spend several minutes setting the metal on the saw. The slices were stored 

on a table in a storeroom until enough had accumulated for shipment, a period of three to four weeks. 

When enough slices had accumulated, the operator would examine each closely to determine the 

weight and lot number stamped on it, which took a total of 30 to 45 minutes.

Besides these practices, MCW (1953e) noted that in their one run per shift, the same two Plant 4 

recast operators had to load the crucibles, clean the furnace, saw the metal, degrease the metal (by 

hand until some time in late 1953), and load the betatron slices. There was no mechanical method to 

move the crucible from the furnace to the scale, so one of the operators had to climb physically into 

the furnace bottom and attach a chain to the billet, remove the fire brick from the billet, remove the 

empty crucible, and weigh the crucible. Meanwhile, at Plant 6E there were separate groups for each 

step, i.e., saw operators, furnace operators, crucible loaders, and cage men (who ran the degreasing 

machine). At Plant 6E, after the metal was sawed, it was kept covered and was stored away from 

contact with workers. Also, the metal at Plant 6E read fairly low compared to the aged Plant 4 metal. 

Thus as MCW (1953e) pointed out, the hazards faced by the Plant 4 recast workers were not being 

considered to the same degree as the Plant 6E recast workers. It is not clear whether the subsequent 

film badge readings went down because changes were made or because the type of operation was 

reduced or discontinued, but it appears that the handling of the crucible was minimized thereafter.

5.2.3

Because of the chemical processing, various residues and wastes had content that was concentrated 

or depleted in particular radionuclides, depending on where in the process the wastes were formed. 

Details are given in Table A-4 and in the text for specific radionuclides. In general, it can be stated

that the waste was disposed of promptly because otherwise storage space would have been 

exhausted. Thus once the waste was taken out of the process equipment or area, exposure to the 

waste was limited to the radioactivity existing in it at the time it was created or collected, with the 

exceptions of the K-65 drums that might be stored until they were transported to SLAPS, the slag that 

was recycled in the later years, the low-level solid wastes (such as rags) that might be accumulated 

prior to incineration; and the residues brought back from SLAPS for reprocessing. In the case of the 

K-65 drums, the exposure would be to radon emanating from the drums and to the gamma radiation 

from the radium; in the case of the slag, the exposure would be to dust and mostly beta radiation; in 

the case of the low-level solid wastes, which were probably kept in dumpsters or similar large bin-type 

containers, the exposure would be dust and external radiation; and in the case of the reprocessable 

residues (e.g., the AM-7), the exposure would be to dust and external radiation. It does not appear 

Residues and Other Wastes and Reprocessing
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that any of the wastes had sufficient time before disposal to build up daughters much beyond what 

was already in the waste, except to some extent the residues brought back from SLAPS.

The residues dumped on the ground or stored in drums at SLAPS are a special case, since they sat 

there for a number of years, but only a few workers entered the storage area intermittently. In the 

case of the residues dumped on the ground, the exposure would be to dust and to some degree from 

external radiation and in the case of the drummed K-65 residue, it would be to radon and gamma 

radiation.

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 1967; AEC 1949b). 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, also called gangue lead cake or GLC. K-65 and other wastes were trucked to 

SLAPS but sent by train to sites outside Mallinckrodt. Radiation levels from the K-65-containing 

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 1949b).

A 1947 dust study done for AEC by the University of Rochester evaluated the relative amounts of 

uranium and radium in the K-65 residue (see Section 5.3.1). It was reported in February 1948 that the 

K-65 contained about 600 mg of radium per ton and about 0.2% uranium; at that time it was being 

sent to the Lake Ontario Ordnance Works at an estimated rate of 6000-12000 pounds per day (AEC 

1949m). AEC (1949b) stated that as much as 100 grams of Ra-226 contained in the K-65 residues 

was produced per month at the Mallinckrodt facilities. This was in fact the quantity produced in 

December 1948 (AEC 1949b). It was reported that 200 grams had been transported at one time to 

Middlesex (AEC 1949b), meaning that up to this quantity had been in storage at the Mallinckrodt site 

or at SLAPS and had had to be loaded for transport at one time.

In addition to the K-65 residues, there were other solid and liquid residues and wastes, as reported in 

February 1949 (AEC 1949m). The other two major solid refinery residues were the barium sulfate or 

AJ-4 residues and the pitchblende raffinate or AM-7 residues. The barium sulfate contained about 4 x 

10-9 grams of radium per gram of residue and about 0.1% uranium; because of the low radioactivity 

content, it was being dumped on the ground at SLAPS at the rate of about 6800 pounds per day. The 

AM-7 consisted of various metal hydroxides with about 0.1% uranium; it too was dumped on the 

ground at SLAPS at an estimated rate of 23,000 pounds per day. The liquid wastes (from the refinery 

processes) contained about 0.002% uranium; they were discharged through a sewer to the 

Mississippi River at an estimated rate of 3000 gallons per day, with about 12 pounds per day of it 

being uranium. Miscellaneous waste byproducts of the UF4 production process were sent to Vitro for 

recovery of uranium, but in relatively insignificant quantities, while the slag from the uranium metal 

production process contained about 0.3% uranium and was stored at the airport. All of the various 

processes generated contaminated scrap metal, with the most problematic, in terms of quantity and 

degree of contamination, being Raschig rings from extraction columns, corroded drums in which K-65 

had been stored before reprocessing, and floor materials (such as bituminous floor coverings being 

replaced).

The K-65 residues were stored in drums. Some of the waste sent to SLAPS was said to have been 

hand-packed by Mallinckrodt workers into barrels (drums) (Eisenbud 1975). A concrete-lined pit (“the 

swimming pool”) was constructed to store the drums in, but was never used “due to health reasons” 

(i.e., the dose rates and perhaps the radon concentrating in the pit would be at excessively high 

levels) (AEC 1959) and because the corrosion of the K-65 drums would have made retrieval from a pit 
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difficult (MCW 1949g). However, AEC (1959) stated that the pit was used to store tailings from the 

slag reprocessing operation from 1955 on. The same storage design and precautions as for the ore 

were followed for the radium-containing waste (MED 1946c), except that the drums may have been 

kept in the open until the storage shed was built in about August 1947 (AEC 1947a). (See also 

Section 5.2.1. above.) Shield walls were included in this to divide it into smaller areas (AEC 1947a). 

Workers who performed operations where the K-65 and ore were stored at the Mallinckrodt main St. 

Louis site and at SLAPS (i.e., the warehouse/yard workers) were placed on a rigid time schedule to 

keep them from exceeding the tolerance dose (MCW 1949g; MCW 1949s; MCW 1949p; MCW 

1950v). For example, MCW (1949s) states that a warehouse worker (including forktruck drivers) 

could, in one week, unload one ore car and make one trip to SLAPS involving K-65 handling and 

other duties within 50 feet of the K-65/ore pad or

Concern was also expressed in 1949 about the bulldozer and crane operators working at the fresh 

and aged raffinate (AM-7) heaps at SLAPS (MCW 1949g). This was studied and it was found that the 

beta dose rate at the cab of a bulldozer pushed all the way into a heap of raffinate was considerably 

below tolerance. Since the bulldozer operator averaged less than 8 hours per week on the raffinate 

heap work, this operator’s exposure was a small fraction of tolerance. These workers were thus not 

issued film badges.

he could work loading one railcar with K-65 drums 

and make one trip to SLAPS ditto; an additional restriction was that no more than 2 hours per week 

could be spent at SLAPS, with only one hour per day actually working at the pad. As Mallinckrodt and 

AEC recognized (MCW 1949g), prior to the reprocessing of the K-65, corrosion of the drums was a 

severe problem: although the K-65 was insoluble, stream pollution at SLAPS was a possibility, but 

repackaging the K-65 would involve significant worker dose. MCW (1949g) suggested that some of 

the “sand” (inferentially, the AM-7 residue, as furnace sand with the radioactivity deposited on it) was 

being put into the K-65 drums, presumably to lower the dose rate from individual drums.

As noted in Section 4.7, most or all of the K-65 was brought back in drums from SLAPS and 

reprocessed starting in early 1948. In January 1949, the gamma dose to the hands of operators 

opening drums of return K-65 was studied by placing film badges on their wrists (MCW 1949e). See 

Section 5.4.1 for results. No correction of the high hand dose situation was made, however, because 

the K-65 drum opening was done over a limited period of time (i.e., once all the K-65 stored at SLAPS 

had been brought back and reprocessed, the operation was over) and it was not realized how high the 

doses were until the study was done late in the reprocessing campaign (MCW 1949e).

The K-65 waste residues included not only the residues themselves, but also the used cloths from the 

Feinc press filters; drum disposal was the only solution for disposal of the cloths, since (1) they were 

impregnated with the radium-lead salt sludge and thus would not burn and (2) they were made of 

plastic fibers and thus would not dissolve in acid (AEC 1949f). It was reported that three used filter 

cloths, stored in two 55-gallon drums, produced 30 times the tolerance level dose rate at contact with 

the drums (AEC 1949f). Thus the cloths and the operations associated with them were of major 

concern with respect to external exposures in Plant 6.

In 1949, a study was done of the beta and gamma exposures associated with preparing, repairing, 

cleaning, and changing Feinc filter cloths by the cloth operators (MCW 1951g). Due to a reduction in 

the external dose tolerance levels by AEC, it had been recommended that an additional (third) worker 

be added to the rotation schedule for this job. Major revisions as a result of the 1949 study were that 

a basket was provided for carrying the used cloths, the filters were partly cleaned by the area operator 

before the cloth operator began work, and a fourth worker was to be added to the rotation schedule in 

May 1950. However, the cloth operators did not like to use the basket because it was hard to haul it 

up the stairways of the filter platforms and they still had to spend some time cleaning the filter. Also, 

during the summer of 1950 several dozen workers were transferred from Plant 6 to Plant 1 (see 
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Section 5.0) and this disruption of operations and the need to retrain new workers delayed the 

assignment of new people to the rotation, which was not done until November 1950. This also 

increased the two principal cloth operators’ doses significantly over the other cloth operators’ doses. 

As a result, it was agreed in 1950 that the area operators would better clean the filters and blankets in 

preparation for the cloth operators’ work, the cloth change interval would be increased from two days 

to three, and a fifth worker would be added to the rotation (MCW 1951g; MCW 1951a). In January 

1951, a new study was done of beta and gamma exposures associated with preparing, repairing, 

cleaning, and changing filter cloths (MCW 1951a); this included cloths for the Feinc, C-3, and 

Recovery filters. Dose rates measured with instruments were combined with time measurements to 

estimate doses; the gamma doses were said to agree well with film badge readings but not the beta 

doses, due to the technical difficulties of field measurements of beta radiation. These results were 

said to be comparable to those found in the 1949 study, although some of the tasks the cloth 

operators were doing in 1949 had been given to the area operators to do (MCW 1951a), but clearly 

the weekly doses had increased at the time of the 1951 study. Also, the operators were still not using 

the basket provided because of its bulkiness. But waiting in the cell for work to start and other work 

practices that increased the “close to Feinc” (distance of less than three feet) time had been changed 

and the cloth change interval was increased from two days to three (MCW 1951a), so the doses were 

expected to decrease. 

There were varying quantities of miscellaneous waste arising from various sources. For example, 

rags and paper with small amounts of uranium forms and residues were created from cleaning and 

wiping procedures, such as the black oxide from wiping derbies, billets, and mold strips and the 

residue from wiping the outsides of drums before transport. Other types of contaminated materials 

included packaging, sweeping compound, old uniforms and gloves, and like items that were used in 

the process areas or were in contact with radioactive matter. All of these types of materials could be 

burned in an incinerator outside Building 115 at Plant 6 or outside Building 403 in Plant 4 (Mallinckrodt 

1994) and the ash could be processed or sent elsewhere for recovery. The filter cloths were also 

incinerated (MCW 1951a), at least during some periods of operation.

Dust from the various processes was collected mostly in bag-type filters, although some was collected

with a rotoclone (a water spray type of collector) and run into a recovery tank (MED 1944o). Dumping 

these filters, i.e., dumping the bags of dust into drums, was recognized as a high-dust operation (e.g., 

AEC 1954a, AEC 1954d, AEC 1954c). Mallinckrodt estimated that bags were changed about every 

six weeks for the Plant 6 Pot Room dust collector and that it took 1.5-2 hours per bag change, for a 

downtime percentage of 0.4% (MCW 1955r). Occasionally a bag would break, obviously resulting in 

even higher dust levels. For these reasons, respirators were supposed to be worn but were not 

always (e.g., AEC 1954e). There was also a possibility of increased external dose from dealing with 

the collectors. For example, in 1956 a maintenance worker received an external dose higher than 

usual, as shown by his film badge reading; this was found to be mostly due to the eight hours he 

spent repairing one of the Hoffman collectors that served the recast (YM-5 or billet) area (MCW 

1956j). This collector handled the beta-active black oxide that resulted from the recast activities 

(MCW 1956j).

Dust was a problem even with empty drums. Mallinckrodt reported to AEC in 1950 that used K-65 

drums being returned to Mallinckrodt in ATMX railcars did not always have lids and were not always 

cleaned (MCW 1950n). This resulted in dispersal of the residual dust throughout the car during 

shipment, to such an extent that Mallinckrodt collected 100 pounds of it in one railcar after it was 

unloaded; there was contamination of the receiving dock as well. Dust concentrations in the railcar 

were in excess of 250 times tolerance (i.e., more than 17,500 dpm/m3

) were found in the general area 

of the railcar and about 27 times tolerance in the worker’s breathing zone, but the radon 
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concentrations did not exceed 20% of tolerance and the gamma radiation did not exceed 15% of 

tolerance. AEC appears to have resolved this by directing the drum sender (presumably Lake Ontario 

Ordnance Works) to correct the situation.

It was noted in January 1955 that for a considerable period, radon concentrations of 1-8 x 10-10 Ci/L 

had been detected in the Plant 6 Cloth Storage Room (MCW 1955j). After much detective work, it 

was found that the wall near the M-50 (tank) sump was saturated with “material” that had come from 

the sump and from the Wash Oliver cell (presumably by liquid leakage and/or by deposition via air 

currents). Holes in the wall and ceiling of the Cloth Storage Room created a draft from the sump area 

across the accumulated material and up into the upper area of the room. This situation appears to 

have been remedied once discovered. However, in December 1956, the Cloth Storage Room was 

again studied because of the wide variation in the radon readings taken during weekly checks; Over 1 

August to 31 December 1956, the 18 weekly measurements of radon level varied from less than 1.0 x 

10-12 Ci/L (i.e., nondetectable by Mallinckrodt methods) to 1.11 x 10-10 Ci/L, with the average being 0.5 

x 10-10 Ci/L and the median being 0.265 10-10 Ci/L (MCW 1957). It was found that there were two 

major radium-containing deposits in the floor and a wall of the room. While the highest gamma 

reading was only 2 mR/hour, in accessible areas, it was thought that the reading would be higher in 

the nonaccessible areas behind equipment. Thus the migration of material from adjacent areas 

appears to have been a continuing phenomenon.

It was recognized that radon levels at the K-65 and ore storage area at SLAPS were high also; along 

with the gamma problem, the radon problem necessitated restrictions on the time a worker could 

spend there (MCW 1949g). Mallinckrodt was supposed to be taking radon samples at SLAPS, but 

after one set had been taken someone mistakenly stopped taking them (AEC 1949j). An AEC safety 

representative then averaged the sixty-plus samples taken up to then to produce results that could be 

graphed so as to see the rough falloff with distance (AEC 1949j). 

The ore drum thawing and drying oven – apparently a large walk-in room with steam coils for heating 

– was used to thaw the K-65 drums before processing and even for temporary storage of dozens of 

these drums (MCW 1949o). Not only did radon emanate from the drums during heating, but because 

the drums were apparently opened for the heating, K-65 had been spilled onto the coils and was 

contributing radon on a continuing basis. This could not all be cleaned off by a vacuum cleaner. High 

radon levels were seen outside the oven during the heating, so a fan and stack were installed in the 

back of the thawing oven to draw away the radon from the drums during thawing (MCW 1949o). 

However, this produced no improvement. It was discovered when the heating coils were turned off, 

even when the fan was on, air was drawn in under the oven door and forced out over the top of the 

door, where it eddied in front of the oven before being drawn in again under the door. This 

phenomenon kept radon concentrations up inside and outside the oven. Although the time spent by 

workers in the oven was short and their daily weighted exposure was below the radon tolerance level, 

it was believed that it was not safe to operate at tolerance levels for extended periods (MCW 1949o). 

Thus the thawing oven ventilation had to be revised and the coils taken apart for cleaning.

As noted earlier, in 1955-1957 Mallinckrodt processed AM-7 residues (see Table A-4) and produced 

3600 gallons of a concentrated thorium nitrate solution that was sent to Mound (AEC 1967; DOE 

2002). AEC (1967) states that Mound purified and concentrated approximately a kilogram of thorium230 from this material, but DOE (2002) states that although 500 grams were produced and an 

additional 500 grams ordered, the latter were apparently never produced. The favorable to claimant

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 10.4 kilograms of total thorium, and 0.15 kilogram of uranium. This is 
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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.

When AEC asked Mallinckrodt to establish a pilot plant to process more of the thorium-bearing 

residue, Mallinckrodt refused, on the basis that the health hazards of thorium were not well 

understood (AEC 1955c). One concern was that because the process involved only solutions (i.e., 

the liquid form), exposures would mainly arise from accidental releases or leaks, which would not be 

easily detectable by ordinary industrial hygiene measures; hence Mallinckrodt wanted some way to 

detect exposure via urinary or other biological measurements (AEC 1955c). AEC agreed to study the 

biological half-life of thorium in the lung and bone following inhalation of thorium fluoride and nitrate 

salts; they were also to study the urinary clearance rate (AEC 1955c). Mallinckrodt approached the 

Los Alamos site because it would be receiving the thorium, but could elicit only the information that 

thorium could be considered equivalent to plutonium on a curie basis (ORAU 1991). AEC agreed that 

the pilot plant would be built and operated as if it were to process plutonium (AEC 1955c) and the pilot 

plant began operation at MEP on 25 February 1956 (ORAU 1991). Because the health hazards were 

on a “speculative level,” Mallinckrodt restricted the number of workers on the project and took strict 

precautions as to handling, contamination control, and access (ORAU 1991). However, there does 

not appear to be any information available as to how the urinalyses were analyzed. 

The “breathing zone” operations in the thorium processing were the removal of the raffinate cake 

(formed in the processing) in dumpsters and the opening of the contactor tank to adjust the 

concentration of the process material and to add laboratory waste materials (AEC 1955d). While AEC 

(1955d) found that exposures of the Plant 7E workers were always below the MAC (note that since 

gross alpha was measured, the thorium and its alpha-emitting daughters were in fact measured in air 

sampling), it was also pointed out that the pilot plant was then in the startup phase of operations and 

that thus the 1955 air survey was possibly not representative of mature operating conditions (AEC 

1955d).

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. Assuming that the waste was all produced in 1942 and sat in 

containers for 15 years, until 1957; this will maximize the radium and radon content. As Table A-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.

Regarding the protactinium processing of residues, there is no suggestion in any reference that the 

Sperry cake used had been processed in any way at the St. Louis downtown site; 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.

AEC (1960) states that the SLAPS waste, also called the airport residue, was evaluated for sale or 

disposal in about 1960 and was sampled even before that. The pitchblende (K-65) residues were 

auger-sampled in June 1953; barium sulfate cake, Colorado raffinates (from domestic ores), and 

miscellaneous ores were also considered. See Table A-4 for quantitative information.

It is not clear whether the storage workers and guards were always Mallinckrodt employees or not, but 

Mallinckrodt workers appeared to have been doing the residue delivery and placement work even in 
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1946-1953 (MCW 1949d; MCW 1949p) and there are mentions of “airport” in urinalysis records. 

ORAU (1989) suggests that Ledoux Company laboratory workers also worked there, probably doing 

sampling. Material was taken to and from the Mallinckrodt St. Louis site by truck (Mallinckrodt 1994; 

AEC 1947a; MCW 1949p). Guards were maintained at the site from 1946 to 1951 (AEC 1959); after 

that, it appears that guards were present only for routine patrols and actual worker entry into the site.

At the time that planning for SLAPS decontamination was being planned in about 1964, the residues 

still stored on the site were the pitchblende raffinate (AM-7), Colorado ore residues, leached and 

unleached barium sulfate cakes (AJ-4), miscellaneous residues (but not K-65), and the C-oxides 

(AEC 1964). The C-oxides were then being shipped to Fernald for reprocessing and the rest were to 

be offered for sale. In addition, there were several hundred tons of contaminated metal and debris at 

the site (AEC 1964).

5.2.4

As noted earlier, other feeds besides ore were used from the mid-1950’s on. The drums of these 

alternative feeds were usually sampled (e.g., for oil analysis) in the Plant 6 warehouse by either 

Analytical Laboratory workers (MCW 1956h) or warehouse workers. The process as performed by 

the laboratory workers was described in December 1956 as follows (MCW 1956h). Three drums were

opened and a small amount was scooped out of each and put into a bottle, which was labeled and 

taken to the laboratory. Air samples of this activity were taken with the warehouse exhaust fan 

running and the large doors (panels) to the outside open. Air movement was found to be almost nil 

and the dust created by the sampling lingered in the air for an extended period. The average airborne

level during sampling was 18,000 alpha dpm per cubic meter. A recheck of sampling was done under 

the same conditions but from the residue drum where the samples taken at the origination point were 

put after analysis there; this drum held the samples from the entire lot shipped and so could be 

assumed to be representative of the lot. Readings during the residue drum sampling showed an 

average level of 8530 dpm per cubic meter. Using the all-purpose hood in the Old Boiler House 

showed levels of 3460 and 584 dpm per cubic meter respectively for the two types of drums. Since 

the residue drum was at most half full and was a single dust source, it was clearly preferable to 

sample that way. It is not clear whether the change to sampling the residue drum and using the Old 

Boiler House hood was made, but it is likely that it was.

Sampling and Laboratory Activities for All Forms

In 1955 feed material had come from or was expected to come from Beaver Lodge, South Africa, and 

Portugal (the latter as a concentrate) (MCW 1955m). A problem identified with this material was that 

it contained scrap iron, cake materials, and large pieces of uranium. Obtaining a representative 

sample of the feed would thus require preliminary separation of the extraneous material. It was 

necessary to process this feed and more like it at the rate of 200 drums per day, hence construction of 

a suitably ventilated sampling station was not an option in the short term. It appears that at least in 

the short term Mallinckrodt sampled these drums with an auger in the ventilated sampling bays of the 

Plant 6 warehouse.

In the early days of the project, the shotgun laboratory in Building 55 of Plant 2 used a 100-mg RaBe

source to test UO3 samples using neutron absorption techniques; hence it had to be designated as a 

special restricted area for access control purposes (Mason 1977). It appears that the RaBe source 

was not used after some early point, since (MED 1944m) and other documents state that such 

samples were prepared at Mallinckrodt and then sent elsewhere for analysis. It may therefore be 

assumed that use of the RaBe source for this purpose had ended by September 1944 (the date of 

MED 1944m).
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MED (1944m) stated that UO2 shotgun samples (possibly the reference meant UO3) were prepared by 

dissolving 25 lbs of UO2 in acid and extracting uranium nitrate hexahydrate in a four-step ether 

process; the remaining water layer, which also contained nearly all of the Th-234 and Pa-234 (UX1 

and UX2), was run through a separating funnel, then evaporated to dryness in a dish, first by the 

chemist using a handheld blast burner, then in a muffle furnace. The sample remaining, which 

averaged 3-5 grams, was ground by hand in a mortar and poured into a bottle. The accumulation of 

sample bottles, usually three to twelve per day, was sent away daily for analysis. In 1954, the 

shotgun samples were made up in the Dry Material Sample Room (or Powder Sample Room), 

apparently because of equipment availability, but the work was performed by the Shotgun Sample 

Lab personnel and not the Dry Material Sample Room personnel (MCW 1954b). There was also a 

Metal Sample Room (MCW 1954b), presumably for assaying the finished metal forms.

Prior to August 1955, an average of two Mallinckrodt UO3 samples were run daily in the Shotgun 

Laboratory and film badge readings were under the applicable limits (MCW 1955e); usually only one 

worker was needed for this. However, from August 1955 into October 1955, a campaign was 

performed to run UO3 samples from other sites as well; these other-site samples constituted more 

than half of the total and were mostly from Fernald but also from Beaver Lodge, Port Hope, “Canada,”

and “Pilot Plant” (apparently not a Mallinckrodt pilot plant). The workload increased by a factor of 

three and temporary workers were hired for a few months to do the work. With one Mallinckrodt 

worker and one temporary per shift to do two Mallinckrodt samples and four other-site samples per 

day, film badge readings went up significantly – one badge read 5810 mrep in one week in August 

and another read 2040 mrep in September, mostly beta. The hot plate hood, the muffle furnace, and 

the evaporating dish in the cooling and scrapedown hood were found to be contaminated with 

material producing mostly beta radiation, while a can of pellets under the sample press was giving off 

a significant gamma dose rate.

The situation was studied (MCW 1955e). It was found that there was no significant radiation reading 

in the laboratory assay of UO3 until the laboratory crucible containing the sample was placed on the 

hot plate; as the sample was evaporated, the reading rose rapidly. The source of the radiation was 

the dried residues on the crucible. For the Mallinckrodt-produced samples, the average reading of 10 

samples was 1500 mrep/hour at 1” above the top of the crucible and 1000 mrep/hour at 2”; the 

respective averages for the four “Pilot Plant” samples checked were 1800 and 1200 and for the ten 

Fernald samples they were >12,500 and 12,000 mrep/hour respectively. (Apparently the maximum 

that the Zeus instrument could read was 12, 500 mrep/hour, so the Fernald 1” readings were 

undoubtedly much higher.) A film badge set at 2 feet over the scale on the main work bench and 

about 5 feet from the boildown and crucible cooling hoods for a week read 1000 mrep beta and 50 

mR gamma. A time-and-motion study showed that a laboratory worker handled the sample in some 

way for a total of 11 minutes after boildown, i.e., after significant readings could be expected to 

appear.

The Shotgun Laboratory was decontaminated by the Mallinckrodt decontamination group, with the 

contaminated items either cleaned or disposed of; a Lucite shield was provided to put in front of the 

crucibles during cooling and a steel box was provided to put the crucibles in during cooling and 

storage (MCW 1955e). Respirators were required to be worn during the scraping, grinding, and 

palletizing of the samples. A film badge placed in the same area as the first one had been read 1500 

mrep/hour beta and 95 mR/hour gamma after two weeks, indicating that the cleanup and other 

corrective measures had been only partly successful.

K-65 sludge was sampled to determine its radium content 24 hours after it was produced, put into 

metal drums and stored in an adjacent warehouse (AEC 1949f). The sampling appears to have been 

done in the warehouse by the warehouse workers in the early years, but in a sampling room in the 
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Scalehouse in later years; the assay was done in the Ledoux Lab by Ledoux company workers (MCW 

1954a, ORAU 1989a).

In March 1955, a survey of the Scalehouse Sampling Room was made to evaluate dust, radon, and 

external exposure hazards from sampling K-65 drums (MCW 1955n). At this time, drums of K-65 

were sampled once a day. Approximately 24 drums were moved, four at a time on a pallet, by 

forktruck from the storage bay to one of the K-65 sampling bays. Two samples were taken from each 

drum using a thief sampler inserted by means of the sampling holes (as described above); after the 

sample was deposited in a pan, the thief was cleaned by scraping it and knocking it against the 

sample rack, putting dust into the air. The pans of material were accumulated inside the sampling 

area until all the drums had been sampled, after which the pans were taken into the Scalehouse area 

and put on a rack to await transport to the Ledoux Laboratory. The operator then cleaned the floor of 

the sample room. It was found that during the 80 minutes of the sampling work period, the total 

external exposure to an operator from sampling a lot of 24 drums was 3.4 mR gamma and 3.8 mrep 

beta; the average total beta-gamma dose rate was 5.5 mrep/hour, the peak radon level was less than 

0.5 x 10-10 Ci/L and the average was less than 0.25 x 10-10 Ci/L; and the average dust level was 1870 

dpm per cubic meter, with both the general area and the operator breathing zone concentration during 

the actual sampling being about 2300 dpm per cubic meter. It was also found that air was frequently 

forced into the Sampling Room through the sampling holes in the floor, especially with a north or west 

wind. This air carried with it dust from its passage across the stored drums of K-65 and thus 

contributed strongly to the dust concentration in the room despite the nominal adequate exhaust flow, 

which was exhausted at floor level to draw down the dust from the sampling pans. It was 

recommended that respirators be worn, which apparently was not being done previously.

In 1953, the Ledoux Lab technician spent three whole days and two half days in the Ledoux Lab and 

two half days in the shotgun (assay) lab (MCW 1953b). At this time, the K-65 (radium-bearing) 

residue testing work in the Ledoux Lab consisted of compositing residue, drying it in an oven, 

blending it in a blender, redrying it in a dry box, mixing it in a tumbler, and then presumably testing it. 

The sources of dust in this operation were identified as using the grinding hood, when dumping pans 

of K-65 for grinding; weighing out (since there was some spillage and some need to scrape off the 

bottoms of the pans); cutting the sample (ground K-65 was cut through a riffle); screening (on the “Rotap”); and using the jaw crusher (for AM-7 residue only), for which there was no hood (MCW 1954a). 

The sink used for washing pans had no sump, so the wash water went to the sewage; splashing from 

the sink was viewed as a source of the contamination on the floor (MCW 1954a). At one point, the 

exhaust fan door was locked closed until the chain assembly could be repaired, hence ventilation was 

impaired (MCW 1954a). The cabinets’ interiors were found to be very dusty and the hood bases and 

walls, which were made of transite, had visible signs of absorbing K-65 (MCW 1954a). There was an 

associated Furnace Room, presumably for drying, in which respirator wearing was recommended 

(MCW 1953b) and the K-65 was said to be weighed on a pan on a scale “through the hole in the 

Furnace Room wall” (MCW 1953b). However, the heat from the ovens (furnaces) created enough 

draft to draw the radon up out of the lab and out over the roof (MCW 1954a). All of this information 

suggests a significant amount of manual handling of the residue and uncertain control of 

contamination, even after a new laboratory was built in 1952 and improvements were made in 1953 

(MCW 1953a).

When the “U-Con” process was set up to recover uranium as a slurry from reject MgF2 material, the 

U-Con began to be sampled in the Slag Separation area of Plant 7 and the samples processed in a 

minimal preliminary fashion in the Analytical and LeDoux Laboratories (MCW 1955L). Four samples 

per day were taken and transported together in open stainless steel pans in a coaster wagon to the 

Shotgun Laboratory. They were processed in a hood with a left-right sliding panel, the purpose of 

which was to keep dust out of the laboratory technician’s breathing area. The technician reached 
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around the panel and used a small hand scoop to remove some of the U-Con and place it in a grinder 

in the hood. After the sample was ground the grinder pan was emptied back into the sample pan. 

Then the sliding panel was moved to allow the dust in the grinder to be brushed out into the grinder 

pan. Finally the panel was slid back and the grinder pan was re-emptied back into the sample pan 

and the sample pan was removed from the hood. The grinding and grinder brushing took a total of 

8.7 minutes and was repeated for each of the four samples per day. With preparation work, the 

technician thus spent about an hour per day in the Shotgun Laboratory. A dust study showed the 

average breathing zone concentration to be 20 dpm per cubic meter and the general area 

concentration to be 23 dpm per cubic meter over the entire four-sample operation, indicating that the 

open pans were more of a problem than the grinding itself, because of the hood.

A check of the Shotgun grinding hood was also performed. This showed that the flow volume was 

234 cubic feet per minute and the air velocity averaged 75 linear feet per minute over the face of the 

hood (MCW 1955L), only somewhat under the 100 linear feet per minute that might be typical of such 

a hood today.

5.2.5

A respirator decontamination facility was added, presumably in Plant 6, in April 1950 (MCW 1950e). It 

is unclear where this was done previously.

Safety, Health, Decontamination, and Laundry Facilities

Although most contaminated equipment could be repaired in situ or in the Mallinckrodt repair shop – 

which was not dedicated to AEC work but served the entire Mallinckrodt general site – some 

equipment had to be sent out to other companies’ facilities for repair (AEC 1948h). AEC approved 

this practice based on its requirement that all items should be decontaminated if possible to “zero 

wipe” (i.e., no detectable removable contamination) and on the assumption that any resulting 

contamination on floors and machines would not be significant. AEC also stated that air samples 

should be taken of representative repair operations in the Mallinckrodt shop and that areas 

immediately adjacent to the Mallinckrodt shop should be monitored for contamination in order to verify 

that surface contamination was not being tracked out (AEC 1948h); it is not clear if this was done or 

not.

MED (1942) stated that both contaminated and clean clothing were laundered on the premises, in 

separate laundries; however, AEC (1947a) stated that contaminated protective clothing from all areas 

was sent to public laundries, contrary to other AEC sites’ practice. Use of public laundries appears to 

have ceased at AEC direction (as suggested by AEC 1947b). An onsite laundry with “clean side, dirty 

side” flow was installed some time prior to April 1950 (MCW 1950e).

5.3 INTERNAL DOSE CONSIDERATIONS

The primary route of internal exposure was via inhalation of airborne particulates, although radon was 

a considerable problem in some areas. It was noted in December 1942 (MED 1942) that the primary 

precautions taken against radiological hazards were against inhalation of the dust. 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.
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A number of parameters related to internal exposures were measured, but very little internal data is 

reported in units of dose. Measurements of airborne concentrations were used as indicators of 

potential exposures and of the need for changes in processes and equipment (as suggested by AEC 

1950a). Uranium 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, although some dose calculations were apparently done by AEC for study or 

verification purposes. In addition, as discussed below, the particle size and solubility of the various 

uranium forms were still being investigated experimentally by researchers. This technical basis 

document presents potential internal exposure information in terms of airborne concentration data 

(and the derived inhaled amounts) and surrogate urinalysis results.

5.3.1

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; Air 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.

A study of air dust particle sizes in the UO3-to-UO2 area was done in 1948 for AEC by the University 

of Rochester (1948b). Both filter paper dust samples and cascade impactor samples were taken at 

the tumbling area (where the newly formed and cooled UO2 was place in tumbling drums and rotated 

on a tumbler to break up clumps); at the skip hoist (where the tumbling drums were opened and 

transferred to the hopper of the Fiberpack loader); at the Fiberpack loader (where a vibrating chute 

fed the material into 75-lb containers and the containers sealed on a roller conveyor); and at the 

warehouse (where the UO2 containers were stored). The sampling rate of the cascade impactor was 

14 liters per minute; it is unclear what the filter paper sampling rate was. All samples were analyzed 

for uranium by the fluorometric method and the results were found to be well within the accuracy 

range of the method. Particle size distributions showed normal curves and to be in excellent 

agreement with the theoretical curves. The results are shown in Table A-7. 

Also shown in Table A-7 are some measurements reported by Sanders (1975). 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 lung-inhalation range, while 

reduction operations (UF4 to metal) have the smallest percentage. 

Eisenbud (1958) reported that “It has been shown that in these [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

g/cc, so that a 1-2 μ particle would behave as though it were a 3-6 μ particle of unit density. An AEC 
Document No. ORAUT-TKBS-0005 Revision No. 02 Effective Date: 06/14/2007 Page 58 of 225

report on uranium mills stated that ore concentrates (the form that arrived 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 was 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 two-stage setup: following a 

cyclone separator, the first stage collected 100% of particles ≥ 10 μm aerodynamic equivalent 

diameter (AED), 75% of those = 5 μm, 50% = 3.5 μm, 25% = 2.5 μm, and 0% of those ≤ 2 μm, while 

the second stage, a filter, collected the rest of the particles. They found that less than 15% of the 

airborne material collected on the second (the respirable-particle) filter. Schwendiman et al. (1975) 

noted that this agreed 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 all types of particle sizes at the Mallinckrodt 

plants and how the information would apply to each worker’s breathing zone over time, the ICRP 66 

default deposition parameters (ICRP 1994) should be used to estimate internal doses for Mallinckrodt 

workers.

Regarding the isotopic and chemical composition of the airborne dusts, there is only a little 

Mallinckrodt-specific information. In 1947 AEC had the ore and residue dust studied by the University 

of Rochester, which investigated the uranium and radium content of filter paper after use in air 

sampling (Rochester 1948a); it was stressed in the University of Rochester’s report on the results that 

the samples were taken not to determine health hazards directly (i.e., by inferring the air concentration 

from worker positions) but rather to obtain a heavy concentration of each material so as to compare 

the relative concentrations in the dust. Thus each concentration was taken at the maximum emission 

point (in fact the centrifuge was partially opened for the sample-taking), without reference to actual 

worker position or breathing zone. The precision of the measurements was said to be better than +/-

10% and each figure in University of Rochester’s report represented at least three analyses. The 

U:Ra activity ratio was approximately 2 in the ore; 0.02 in the gangue (GLC or K-65) residue from the 

Scalehouse and 0.3 to 0.4 in the Centrifuge area where the liquid and solid residue (Ra-bearing) 

fractions were separated, indicating removal of most of the uranium from the K-65. The U:Ra ratio of 

2 indicates that U-238, U-234 and Ra-226 were approximately in equilibrium in the ore.

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 (Wilde 1975). 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 
Document No. ORAUT-TKBS-0005 Revision No. 02 Effective Date: 06/14/2007 Page 59 of 225

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 the Th-230 and Ra-226

concentrations were reduced in the processing (Wilde 1975). However, as discussed in the 

paragraph above, at least the one study of what was probably Belgian Congo ore at Mallinckrodt 

indicated secular equilibrium with Ra-226. 

AEC began taking air dust samples for particulate alpha emitters in 1943 and Mallinckrodt appears to 

have been taking dust samples and sending them to NYOO for evaluation from about 1943 on (MED 

1944j). The various AEC air dust study reports state that an established AEC protocol was followed, 

but the documentation of the protocol is lacking. However, some information is known about 

Mallinckrodt’s and AEC’s methods. MED (1943d) stated that the sample rate for measurements in the 

second quarter of 1943 was 3 ft3

/min, with either 30 or 120 ft3 sampled, depending on the area. 

Mallinckrodt (MCW 1946g) stated that dust was measured with a “precipitron and ion-meter.” For the 

data reported by MCW (1949d), dust samples were collected on 1-1/8” 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%. Data from 1946 lectures by A. A. Jarrett 

of Clinton Laboratories were used in the error calculations. AEC (1949g) also states that Mallinckrodt 

was using a flow rate of 0.5 ft3

/min, with a Fisher and Porter rotometer. AEC (1950a) stated that it 

was using 1-1/8” Whatman filter paper disks and a handheld air sampler with a collection rate of 15-20 

linear feet per minute (lfm) to take its samples at Mallinckrodt.

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 (usually

Whatman #41) and counted on an alpha scintillation counter (e.g., the laundry samples of Utnage 

1958b). 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). In 1958 at the Oak Ridge Gaseous Diffusion Plant, the Whatman #41 paper was 

used with fixed counting equipment, for a counting and collection efficiency of 30% (Becher 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 measurement techniques were consistent with these 

references.

In 1948, some samples on filter paper were taken at Mallinckrodt, conveyed by an AEC inspector to 

Clinton Laboratories in Oak Ridge to be counted for alpha and beta-gamma activity, and then returned 

by the inspector to Mallinckrodt to be counted for alpha activity there, apparently as a comparison test 

(AEC 1948a). The samples were taken in the ore milling area, where ore drums were opened and the 

ore was ground to the desired consistency for later digestion; thus it can be inferred that the samples 

showed uranium in equilibrium with its daughters down to radon (as suggested by the results of 

Rochester 1948a – see table above). The counting equipment was said to be a parallel plate counter 

at Mallinckrodt, but was unspecified for Oak Ridge. For both the Oak Ridge and Mallinckrodt alpha 

counting conditions (with identical geometries), the absorption loss was assumed to be 30% and the 

geometry factor to be 52% (with the extra 2% above the 50% being ascribed to backscatter of the 

alpha particles). For the Oak Ridge beta-gamma counting conditions, the absorption loss was 

assumed to be 30% and the geometry factor to be 10%; the beta-gamma counting apparatus at 

Mallinckrodt was being repaired, so no beta-gamma measurements were made there. The (alpha?) 

background at Oak Ridge was 4 cpm, while the background at Mallinckrodt was 20 cpm. The 
Document No. ORAUT-TKBS-0005 Revision No. 02 Effective Date: 06/14/2007 Page 60 of 225

conversion factor for alpha was given as 2.75 dpm per cpm and for beta-gamma as 14.3 dpm per 

cpm. (Although these figures were backward in the text of the reference, the data sheets show the 

correct correspondence.) In 1949, Mallinckrodt (MCW 1950g) stated that it was assuming the counter 

geometry to be 52% and the filter paper absorption to be 30%.

AEC (AEC 1948a) noted that the samples taken could not be read at Mallinckrodt for several days 

after they were brought back from Oak Ridge because the parallel plate apparatus and the counting 

room had to be decontaminated and the background reduced from 110 cpm to 20 cpm. The 

agreement for the alpha samples was fairly good, in general mostly within 25% and in many cases 

much better. For a subsequent set of samples counted at Mallinckrodt (with the background now 

down to 6 cpm), the same AEC inspector reported (AEC 1948b) that the formula used to convert from 

alpha cpm to microcuries per cubic meter was as follows, with the 0.3 and the 0.52 being the 

absorption and geometry factors respectively and the 0.5 being the intake rate of the sampling pump:

# μCi/m3 = 

 (1 - 0.3) × .52 (c/d) × 3.7 × 1010 (d/sec/Ci) × 60 (sec/min) × 0.5 (ft3

/min) × 1 × 10-6 (Ci/μCi)

 # c/min x 35.3 (ft3/m3) x sampling time (min)

A report issued about the same time by AEC to Mallinckrodt gives the results of dust samples taken in 

an unspecified Mallinckrodt location and counted apparently at NYOO (AEC 1949i). Counts are given 

in dpm per cubic meter, with alpha counts given for all samples and both alpha and beta-gamma 

counts given for about half the samples. These latter show alpha-to-beta-gamma ratios ranging from 

0.98 to 6.44, with most values in the range 2.5-3.4.

John Harley, former director of HASL, told an epidemiological studies group (ORAU 1983b) that the 

equipment used to do the air dust surveys was not very good. He said that HASL had confidence in 

readings greater than or equal to 25 ug/m3

, but in the range of 10 ug/m3, the results were thought to 

be “shaky.” He added that although most of the production sites did not do breathing zone sampling 

(only general air sampling) and some even resisted it, Mallinckrodt did do breathing zone sampling 

(e.g., MCW 1949d).

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. AEC was also 

interested in the feasibility of using the levels of urinary excretion of uranium as a control metric in 

production and processing facilities (AEC 1949d); to do this, it was necessary to correlate the urinary 

measurements with the dust exposures, which could not be done if only general, instantaneous dust 

levels were used. The time-weighted averages for Mallinckrodt are documented in a series of AEC air 

dust study reports (e.g., AEC 1954b) and in some Mallinckrodt reports (e.g., MCW 1949d).

5.3.2

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 

Airborne Dust Levels
Document No. ORAUT-TKBS-0005 Revision No. 02 Effective Date: 06/14/2007 Page 61 of 225

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). 

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. AEC 

also stated many employees were exposed to elevated dust levels for years (AEC 1949b); AEC 

estimated that the inhalation of many Plant 4 operators involved in UF4 production was 27 grams or 

more of uranium, as compared to what it termed a “life tolerance” of 6 grams, and that an additional 

two years of exposure would add about 2 more grams (MCW 1949b). 

The high levels were of concern to both MED and Mallinckrodt. MED (1944h), in transmitting to 

Mallinckrodt the results of dust samples taken in June 1944 in the bomb and furnace areas, noted that 

the results were high and that either ventilation in those areas should be improved or respirator use 

should be required. Mallinckrodt too was concerned by the high levels, for example the fact that dust 

concentrations around the various crushing, grinding, and packaging operations consistently 

exceeded the tolerance levels by a factor of 1-300 (MED 1944k). As a result, Mallinckrodt hired a 

consulting engineering firm to put together a re-design proposal for dust control (MED 1944k).

MED (1944n) stated the following. In September 1944, MED and Mallinckrodt representatives met to 

discuss proposals for ventilation improvements in the green salt and metal production areas. It was 

decided that the ventilation in the green salt furnace room did not need improvement because there 

were 30 air changes per hour and because the “smell of HF [hydrogen fluoride] was faint.” But on the 

other hand, the green salt unloading, grinding, milling, and blending operations were so dusty that it 

was decided to do them henceforth under hoods in a continuous sequence, instead of as separate 

operations, and the blending would be done mechanically instead of manually. Also, the bomb 

loading, jolting, unloading, and charge blending and the biscuit chipping, all of which were carried out 

in close proximity, would be provided with separate, localized ventilation. Finally, the green salt 

blending operation would be redesigned to use a tumbler-type mixer instead of manual mixing, after 

which time the blender table would no longer be used. Thus in about 1944-1946, significant

incremental improvements in dust control were made that reduced initial production-level dust 

concentrations in the green salt area. 

Respirators were said to be required “for practically all plant operations” in 1946-1948 (Mason 1958a) 

but respirators were not used consistently (see Section 5.3.3 below).

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 1950, for example, it was reported that the air supply system in the Ore Room was 

such that there was a significant outward flow through the drum opening/closing area shield (MCW 

1950u); in 1953, an AEC inspector reported that there were many small openings between the 

operating area and 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 that as cans of UF4 were transferred from 
Document No. ORAUT-TKBS-0005 Revision No. 02 Effective Date: 06/14/2007 Page 62 of 225

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 1948-1950, tighter administrative controls were instituted and 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 uraniumrefining 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 A-8 and A-9 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 or representative 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 

1950c; MCW 1955d); also, beta-gamma activity analysis was not routinely done (MCW 1955d). Thus 

Tables A-8 through A-16 must be assumed to be gross alpha measurements. There were no 

continuous air monitors at the Mallinckrodt site, at first because there was no commercial monitor 

available (MCW 1950e) and later because the results were thought to be of doubtful value (MCW 

1955d).

Table A-10 presents some early air sampling data taken by MED. The samples represent single point 

measurements and no correction was applied for exposure time. Thus unless some assumptions are 

made as to time spent in the given area performing the given operation, only a bounding or 

conservative dose calculation can be performed with this data. However, there are some afteroperation or “not in operation” measurements among the data that could help to characterize an 

average level, if needed.

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 favorable to claimants, 

in general. The AEC data are 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 comparative form in Table A-11 and in comprehensive form in Tables A-12 through A-16.

Table A-11 presents the results of airborne dust surveys made in Plant 4 by, respectively, AECNYOO’s Medical Division in May 1948 and Mallinckrodt safety officials in September-October 1948. 

These results are given as time-weighted daily average radioactivity (particulate) concentrations

(called DWE levels by AEC) in air by plant. Table A-11 shows that AEC and Mallinckrodt’s data were 

in general agreement (AEC (1949b) termed it “excellent agreement”), although there were some 

differences. Mallinckrodt (MCW 1949d) stated that while the Mallinckrodt and AEC results did differ 

significantly in some instances, it was thought that none of the differences was highly significant 
Document No. ORAUT-TKBS-0005 Revision No. 02 Effective Date: 06/14/2007 Page 63 of 225

except in the case of the office workers; Mallinckrodt (MCW 1949d) also noted that the (overall) lower 

results of the Mallinckrodt studies were due to Mallinckrodt’s taking more control samples and 

counting low-activity samples for longer times to reduce statistical error.

One notable difference is for “Cage handling,” where the AEC-measured level is given as 2.7 times 

the tolerance level and the Mallinckrodt-measured level is given as 52 times tolerance, or 189 and 

3640 dpm/m3 respectively. The ratio of these two values is about 20. In the original report in which 

AEC reported its values ( AEC 1848l), the job description and summary listing of job categories and 

DWEs does not include a mention of the cage handling function; there is only a mention of the “Saw 

man” as taking billets to the cage area. However, in the original report in which Mallinckrodt reported 

its values and where the comparison table (assembled by Mallinckrodt) first appears (MCW 1949c), 

there is a category called “Cage man.” Thus it appears that the job category of cage man did not exist 

until mid-1948 or after, i.e., after the AEC study was made. Mallinckrodt seems to have put together 

an estimate for the “Cage handling” category from the AEC data as some sort of composite of the 

values for two or more of the other categories, probably including the slag handler (1.6 times the 

tolerance level). However, given the general functions of the cage man as later described in other 

dust studies, the cage man’s exposure was more likely to be similar to that of the saw man (15.8 

times tolerance according to AEC) and the chipper (26.8 times tolerance according to AEC).

Further, an inspection of the detailed job sheet for “Cage man” in the Mallinckrodt report shows that 

about two-thirds of his dust exposure was due to a single activity, “Dumping D-7” (MCW 1949c). 

What D-7 was is not defined in any available Mallinckrodt document but it appears to be a form of slag 

dust. 

Dumping D-7 is again mentioned in the Mallinckrodt dust study done in December 1949, but the 

average levels reported here were reduced by a factor of about 1000 from those of the earlier 

Mallinckrodt study. This reduction is most likely attributable to administrative measures taken in 1949 

to reduce dust levels in the highest-level activities, but it could also have been due to malfunction of 

the sampling apparatus in the earlier study, as explained below.

Also, the high value listed in the 1948 Mallinckrodt study (MCW 1949c) for “Dumping D-7” appears to 

have not been included in the calculation of the DWE by Mallinckrodt. This may have been due to a 

belief that the sampler malfunctioned in taking that sample. It should be noted that the average level 

in the next highest area of Plant 4 as measured in this study was less than half this value and that the 

known highly dusty operations reported on by Mallinckrodt (MCW 1949d) for Plant 6 for October to 

November 1948, such as ore room, pot room work, and UO2 handling activities, typically had highest

sample levels that were lower by 50% or more than the MCW (1949c) “Dumping D-7” value. (The pot 

room and UO2 handling had only single samples.) Because Mallinckrodt did not include this value in 

the calculation of the average, probably because they thought it suspect, it is possible that the other 

D-7 measurements apparently taken at the same time may also have been higher than the true 

values. Thus the value for “Cage handling” attributed to AEC by Mallinckrodt may in fact be more 

correct.

Thus the AEC-Mallinckrodt discrepancy in the “Cage handling” category may be due to the job’s being 

new with the Mallinckrodt study, to a mis-estimation of the cage man’s exposure from the AEC data, 

or to a malfunction of the sampler.

An additional note about the dust levels measured in 1948-1949 and their applicability to previous 

years is in order. It is clear from the inclusion in Table 1 of Mason (1958a) of 1946 through 1950 

figures (repeated in this technical basis document as Table A-8) that 1946 and 1947 data were

available that Mason thought appropriate for comparison to 1948, 1949, and 1950 data, or at least 
Document No. ORAUT-TKBS-0005 Revision No. 02 Effective Date: 06/14/2007 Page 64 of 225

that Mason believed (by his repetition or plateauing of the same values throughout 1946-1948) that 

the 1948 data were representative of the 1946 and 1947 operations as well. This can be inferred for 

Plant 4 as well from Table 2 of Mason (1958a) (repeated in this technical basis document as Table A8). Thus although the report argues from the paper’s Table 1 that significant improvements were 

made by late 1948 through 1950 that reduced dust levels considerably, this is irrelevant to a 

significant conclusion that can be drawn from Mason’s tables: that the 1948 data was representative 

of the 1946-1947 data.

This is also borne out by other information from Mason (1958a). This paper states that the dust 

concentrations “in some operations” in 1942-1943 “were considerably higher than present standards.” 

However, the paper was likely written in 1958 (the conference at which the paper was given took 

place in October 1958), so the “present standards” were those of the late 1950s. Mason might 

arguably have made the same comment about 1946 through1948, since he also stated of the 1946

through 1948 operations in Plant 6 that “during initial operations dust control was minimal, and it can 

be seen that air concentrations were high during the period 1946 through 1948, when respirators were 

required for practically all plant operations.” While Mason stated that the new health program was 

authorized in 1947 and got under way in 1948, he also said that one of the first projects of the new 

Mallinckrodt Health Department was “a thorough analysis of the dust data already accumulated and 

the immediate collection of additional data to enable an estimation of dust exposure already received 

by operation and maintenance personnel.” Thus even in 1948, apparently no immediate engineered 

improvements were undertaken. Mason went on to say that only in 1949, under the new health 

program, “immediate steps were taken to install good ventilation and dust control and to initiate 

process improvements.” This implies that the point at which dust levels began to go down 

significantly in Plant 6 was in 1949, as is also shown from Mason’s (1958a) Table 1. Regarding Plant 

4, Mason stated that mechanization was installed in 1948 and 1949, but since he gave the same 

figures for 1943 through1947 as for 1948, it must be concluded that either he considered the 1948 

values to bound the 1943 through1947 ones or the values he used for 1943 through1948 (from 

whichever year) corresponded to those that were representative of the entire period.

Mason (1958a) referred to his Tables 3 and 4, showing Plants 7 and 6E values respectively (repeated 

in this technical basis document as Table A-9), as showing an improvement of “a factor of 60 

compared to 1943 figures and by a factor of 8 compared to 1950 figures” for Plant 7 versus Plant 6 

and a reduction factor “of about 20 compared to 1943 figures and 8 compared to 1950 figures” for 

Plant 6E versus Plant 4. These explicit comparisons too suggest that Mason had information as to 

the early (1943) dust levels and that his tables reflect that. Mason also plotted his data from 1943 on, 

again typically showing a plateau for the years 1943-1947.

Mason (1958a) noted that “no regular dust sampling program was in effect during 1943 through 1947, 

but sufficient samples were collected to show that airborne uranium concentrations were high by 

present standards; concentration of 50 to 100 times the present MAC [70 dpm/m3] were not 

uncommon, and some operations produced concentrations up to 1000 MAC for a few minutes.” 

However, such levels were also measured in the later dust studies, e.g., near the end of the paper, 

Mason commented that it was “evident that personnel did work in fairly high concentrations in the 

early days of the operation, and that the exposures received depended partly on the effectiveness of 

the respirator program.” However, he appears to consider the early days to be 1942 through 1947 or 

(by his tables) 1942 through 1948.

This does not mean that the dust exposures in about 1949 through 1951 were not also high (above 

the preferred level), although not usually as high as in 1948, as was shown by the following table of 

data. Some dust sample measurement data from 1948 and 1950, in multiples of the preferred level 

(PL), which was the same as the MAC, are shown below. 
Document No. ORAUT-TKBS-0005 Revision No. 02 Effective Date: 06/14/2007 Page 65 of 225

Survey date May 1948 Nov 1948 Jul-Aug1950

Survey performed by AEC Mall Both

General information about the studies 

Number of workers in areas studied 170 279 406

Percent at less than the PL* 31 43 61

Percent at 1-3 PL 20 32 31

Percent at 3-5 PL 23 3 5

Percent at 5-10 PL 13 13 0.7

Percent at 10-15 PL 2 0 0

Percent over 15 PL 12 9 2.5

General Categories (in multiples of the PL)

Average exposure 53 12 2.1

High exposure 660 195 20** 

Ore Room operations 71 195 4.9

Digest/Feinc/feed makeup/C-3, centrifuge operations 3.4 - 12 9.8 1.1 - 2.2

Pot Room operations 460 111 4.8

Raffinate and recovery operations 5.2 3.9 1.8 - 2.3

Rockwell (brown) furnace operator 350 76/45 20/3.1

Brown oxide packer 560 161 5.2

Warehouse worker 3 1.4

General maintenance and mechanics 1.8 1.3

Boiler House 0.63 0.62

Digest pilot plant 3.5 1.3 1.5

Experimental continuous furnace (pilot plant) Not in operation yet Not in operation yet 122

Ledoux Lab K-65 sampler 30 20

Shotgun Lab 0.34 3.4

Production offices 1.4 0.2

MCW offices 0.7 0.08 0.013

AEC offices 0.5 0.11 0.013

Guards 1.1 0.5

Health office 0.2 0.1 0.16

Other Information

Ore Room: cleaning ore drums, dpm/m3 Range 1710-127000,

avg 64400, for 30 min ---

Range 253-496,

avg 374, for 64 min

Ore Room: general air, operating, dpm/m3 Range 275-8920,

avg 2590, for 125 min ---

Range 700-5750,

avg 1954, for 20 min

Figures are from AEC (1948l), MCW (1949d), and AEC (1950o). Dashes in the table above indicate the range given in AEC (1950o), while slashes indicate two 

types of workers of the same title, e.g., day shift versus night shift workers.

* The PL is the preferred level, equal to 70 dpm/m3

. All dust sample measurements are in multiples of the preferred level (PL, or 70 dpm/m3), except as 

indicated.

** This figure does not include the exposure of the three Experimental Continuous Furnace workers’ exposure at 122 PL, although their exposures were 

included in the calculation of the percent over 15 PL. 

In 1948 there were areas where the exposures were indeed “50 to 100 times the present MAC [i.e., 

50-100 times the preferred level of 70 dpm/m3

, or 3500-7000 dpm/m3]” and that some operations, 

e.g., the ore room drum cleaning, clearly did produce concentrations up to 1000 MAC (70,000 

dpm/m3

) for periods of minutes at a time. Thus it appears that the 1948 figures do fit Mason’s 

description of what “high” was.

Some comparisons were made between spot dust levels measured in 1943-1946 (Table A-10) and 

the corresponding ones measured from 1947 on, with the point being to verify that for the presumably 

peak or near-peak points of individual operations, the 1943-1946 dust levels were generally no 

greater and were usually less than they were later, after Plant 6 was in full operation. This too 

supports the conclusion that the 1948 figures, prior to installation of significant upgrades in Plant 6, 

would bound the experience in the earlier years.

Data similar to that in the table above can be shown to exist for Plant 4 (i.e., for 1948 and before), so 

that the “50-100 times” quotation appears to apply for Plant 4 as well. Also, in AEC (1949c), a 

memorandum from early 1949, an AEC official summarized the hazards in Plant 4, which he said 

were to be solved by the completed designs for Plant 6E, including dust levels of up to 186 times the 

preferred level. This suggests that the 1948 figures for Plant 4 are also indicative of the exposure 
Document No. ORAUT-TKBS-0005 Revision No. 02 Effective Date: 06/14/2007 Page 66 of 225

levels of earlier years. The spot dust levels for the early years, when compared to those of about 

1947-1948, also appear to show that the 1947-1948 levels bound the early years.

It can be concluded that Mason (1958a) and the references supporting the table above show that the 

1948 figures did bound or were representative of the figures of the earlier years. Comparisons with 

the data from earlier years are consistent with this conclusion.

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 A-17; some 

information about the laundry 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 5.3.5 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.

No information is available regarding dust exposures at SLAPS. This was likely due to low radioactive 

dust levels in this open area (i.e., most of the dust collected would have been nonradioactive even in 

the dump pile areas, except possibly when the dumping was actually taking place) and to the 

contained nature of the K-65 residue. Because the SLAPS workers likely were subject to higher dust 

levels in their regular work at the main St. Louis site (e.g., in the warehouse or loading the dumpsters 

with barium sulfate residues or doing guard rounds in the plant), their dust exposures are based on 

their main St. Louis site plant exposure and thus are expected to be conservative.

The reported air concentrations generally pertain to those workers directly and continuously involved 

in uranium refining work. However, Breslin (1958) defined “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.

The case of the maintenance and craft worker requires special consideration. Area mechanics were 

assumed to be dedicated to the particular process area or plant, with their work time being mostly 

spent in and around process areas, and thus their daily weighted average inhalation exposures 

should be considered to be representative of their residence time in the process areas. Similarly, their 

radon exposures can be taken to be those of their process areas. However, it is more difficult to 

determine inhalation and radon exposures for maintenance and craft workers who spent part of their 

time in their shop areas, where contaminated equipment might be brought to them to work on at 

irregular intervals, and part of their time working on equipment in process areas. As Hickey and 

Dupree (1984) pointed out, Mallinckrodt Uranium Division maintenance and craft workers were not 

usually assigned to a particular process (with the exception of area mechanics, which Hickey and 

Dupree do not note) but served the entire division. Not only Breslin (1958, quoted above) but also 
Document No. ORAUT-TKBS-0005 Revision No. 02 Effective Date: 06/14/2007 Page 67 of 225

MCW (1950c) pointed out the variability of exposure of these workers as compared to the process 

area workers.

AEC and Mallinckrodt dust studies from 1948 on, list at least some maintenance workers besides 

area mechanics as a category and tabulate the time spent in various activities and areas in some 

detail (e.g., MCW 1949c; MCW 1949d). However, AEC (1954b) and MCW (1949d) stated that the 

studies they described used only general area samples to estimate the exposures of maintenance 

workers and not breathing zone samples while they were performing specific operations (unlike the 

case with the process workers). MCW (1950c) made a similar point with regard to the AEC-directed 

back-calculation of dust exposures to Plant 4 employees in 1949. Although it is not possible to 

determine what the temporarily elevated levels in the shops and the resulting somewhat higher 

exposures were, it can be concluded from the data in the various dust studies that the shop area dust 

levels were lower than the process area levels, usually decidedly so, and that thus the most significant 

contribution to the exposures of these workers most likely came from their time spent in process 

areas. It is assumed that from 1948 on, this time is factored into the exposures of these workers (as 

part of the daily weighted averaging calculation, Appendix C shows an example of this calculation).

Relatively high potential for dust exposure applied not only to those actually present in the dusty 

buildings and to some extent to those working elsewhere in the plants, but even to those 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 co-located (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 1955d). 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 Weinstein 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.

Table A-18 lists job titles obtained from AEC and Mallinckrodt reports and from film badge and 

urinalysis records; in addition, it gives a geometry factor set appropriate for each job title (for later use 

as described in Section 7), which have likely been superseded by project guidance. Because the 

tables include information of interest for comparisons, they have been retained.

Table A-18 was prepared to be used with Tables A-19 through A-22 and later tables to help determine 

the exposure to an individual worker when bioassay data for the worker was missing or was

conflicting and when comparable worker bioassay data (see Section 6) was insufficient. Tables A-19

through A-22 were derived from condensing information in Tables A-12 through A-16 and from other 

sources; the data they contain thus represents average daily weighted air concentrations in nearly all 

cases. Additional data for individual sampling and laboratory operations is given in Section 5.2.4. All 

of the data in the tables mentioned are based on a natural uranium mixture, with the exception of 

thorium processing, as discussed below.
Document No. ORAUT-TKBS-0005 Revision No. 02 Effective Date: 06/14/2007 Page 68 of 225

The process, in which AM-7 residue was converted into a concentrated thorium nitrate solution, as 

explained in Section 4.7, 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. The processing appears to have been entirely of liquids 

(AEC 1955c). Besides the AEC-measured data given in Table A-16, there are 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 weight in the original residue, 

hence the Th-230/Th-232 mass ratio will be assumed to be 0.116/0.884, or 0.131. This gives a Th232 source to be added to the Th-230 source term (see Table A-6). Although the internal dose per 

curie intake is generally somewhat higher for Th-232 than that for Th-230, the specific activity (Ci per 

g) of Th-230 is over 105 times that of Th-232.

5.3.3

An undated MED reference that is assumed to be of 1942-1944 vintage (MED undated a) stated that 

respirator use was mandatory when the process required use; MED (1942) and MED (undated a)

stated that this was especially necessary for the UO3 and ore grinding and sifting operations 

respectively. An MED reference from May 1945 shows layouts indicating the areas of required 

respirator wear (MED 1945n). A Mallinckrodt manager stated in 1946 (MCW 1946g) that although 

dust levels had been high, exposures were being reduced not only by improving the dust systems but 

also by having workers keep their respirators on all the time that they were working in high-dust areas 

(thus implying that this may not always have been the practice). An AEC inspector noted in 1947 that 

although respirators were supposed to be worn in high-dust areas such as the ore grinding and 

orange production area, they were seen to be worn around the neck or even not at all in areas 

immediately adjacent to the cited areas; these adjacent areas were visibly covered with finely divided 

particles that could clearly be made airborne when swept off the floors and walls by drafts (AEC 

1947a). 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. 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 1955d). The requirements 

for them were spelled out in standard operating procedures (MCW 1955d). Even so, AEC inspectors 

noted instances where visible dust clouds were present or they measured clearly significant dust 

levels, yet the operators were not wearing respirators ((AEC 1954b; AEC 1954c; AEC 1954e; AEC 

1954f; AEC 1954g; AEC 1955e).

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 
Document No. ORAUT-TKBS-0005 Revision No. 02 Effective Date: 06/14/2007 Page 69 of 225

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) stated 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 asserted 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.

MED (1946b) gave acceptable types of respirators for use in uranium processing work, as follows. 

For use in U3O8 dust, the MSA “Comfo” 2101 with the All Dust 2133 filter was acceptable; for use in 

UO2 dust, the MSA “Comfo” 2101 with the Dust & Mist 2130 filter, the MSA Dustfoe 2147 with the 

Dustfoe 2138 filter, or the Willson 770 2119 with the Fume 2123 filter; and for use in UF4 dust, the 

Duper 46 2111 with the 2124 filter or the MSA “Comfo” metal frame 2101 with the Dust & Mist 2139 

filter. A 1955 list of Plant 6 process cells, dust collectors, and tanks, together with the then-current 

uses of the tanks, appears in MCW 1955g; maximum emergency nonsurvey stay times (i.e., for 

urgent access and not inspections) and the general respiratory protection (mask) types are also given 

in this reference. It appears that the stay times are based on external dose rate, not on inhalation 

considerations.

In the absence of any firm figures on respirator use and efficacy at Mallinckrodt, it will be assumed 

that respirators were not reliably used.

5.3.4

As noted earlier, radon levels could be significantly elevated in enclosed areas where material 

containing uranium daughter products was stored. ORAU (1989a) stated that a short burst exposure 

was more typical than a sustained exposure and this appears to be borne out by contemporary 

descriptions of the various operations in which radon might be released. 

Radon

Radon measurements by Mallinckrodt and AEC were reported in units of 10-10 Ci/L. It is clear from 

MCW 1955k (which, despite its lack of company heading and date, is assumed to be a Mallinckrodt 

procedure circa 1955) that for both the “environmental” radon measurements and the breath radon 

measurements, it was just the radon and not the radon daughters in the air that was reported. That is, 

the procedure specifies that the air sample be alpha-counted for 14 hours, i.e., allowing the radon to 

come to equilibrium with its daughters Po-218 and Po-214, and that the first two hours of counting

results are to be discarded, i.e., discarding the contribution of the radon daughters already in the air at 

the time of sampling. The data from 2 to 14 hours is then extrapolated back to zero. As also stated 

by MCW (1955k), the chamber standardization value is given as a theoretical 266 counts per hour per 

pCi of “radon in equilibrium with its daughters,” with half the counts due to the radon and the other half 

due to the two daughters. The radon decays are assigned a counting efficiency of 100% in the 

ionization chamber, but the daughters are assigned an efficiency of only 50% because they deposit 

on the walls of the chamber. Thus in that laboratory and with that type of chamber, the count rate was 

actually 220 counts per hour per pCi of radon in equilibrium with its daughters, for an effective total 

efficiency of 83%. However, because the net counts per hour are divided by the chamber 

standardization value, the result is in pCi of radon alone, not radon and its daughters.

Operations involving radon exposure potential were examined by the Mallinckrodt health physicist and 

supervision to determine if there was any step in which transient exposure to flood (sudden high) 

radon concentrations were a possibility (MCW 1950e). From perhaps the mid-1940’s on, radon 
Document No. ORAUT-TKBS-0005 Revision No. 02 Effective Date: 06/14/2007 Page 70 of 225

samples were taken once a week in high-radon areas (MCW 1950e). This included the outdoor yard 

areas, from which measurements Mallinckrodt concluded that some outdoor area was always 

downwind and that the Plant 6 outdoor background (based on “ordinary” emissions to the outdoor 

plant atmosphere) was on the order of 1 x 10-11 Ci/L (MCW 1949n). AEC (1949g) stated that 

Mallinckrodt had constructed several radon measuring devices for its own use and were in fact in the 

process of constructing one for NYOO. The standard used by Mallinckrodt was made up by the 

National Bureau of Standards (AEC 1949l). AEC (1949g) also reported that the author of the 

reference, an AEC inspector, had visited the airport (SLAPS) with Mallinckrodt health personnel and 

had assisted them in taking radon samples by using an extension pole device designed by 

Mallinckrodt; samples could thus be taken at levels as high as 20 feet above the ground. It seems 

clear that based on (1) the long history of radon sample taking at Mallinckrodt, with the start preceding 

by some years the processing of high-radium and thus high-radon pitchblende ores, and (2) the 

apparent competence of the Mallinckrodt health and safety, and instrument people at developing and 

using radon-measuring instruments, the radon measurements can be regarded as reasonably 

representative of the conditions.

Some early radon measurements from 1945 are shown in Table A-23. The principal locations where 

difficulty was found in 1946 in keeping radon concentrations below tolerance levels were the railcars 

during unloading of ore drums and in the ore and residue storage area (MED 1946c).

MED (1946c) noted that a railcar could be unloaded fast enough that the then-current ventilation 

methods and restrictions spent on time in the railcars were usually sufficient to keep below tolerance 

levels. Still, the radon problem became more acute with the heavy use of pitchblende ore and in 1948 

AEC provided or arranged for new railcars to be used that could accommodate installation and deinstallation of fans at the ends of the cars at the delivery points (MCW 1948h). Presumably the 

openings were covered with panels while the railcar was in transit. In October 1948, Mallinckrodt 

tested the efficiency of the ventilation in one of these new railcars (MCW 1948h). The door was left 

open two feet overnight, and then closed for three hours until the start of the test. The fans were 

started and at subsequent intervals the radon level was measured at two points at the center of the 

car halfway between the door and either end. The result was that the radon level fell from an average 

of 7.03 x 10-10 Ci/L at time zero to about 0.5 x 10-10 Ci/L after about 3.5 hours, reaching the 1.0 x 10-10

Ci/L point some time around 20 minutes, although the levels fluctuated at and between points. While 

the ventilation was seen to be effective, two drawbacks were noted. First, the gamma dose rate at the 

top and side of the car was 20-32 mR/ hr and it took two or three workers about 1.5 hours to install the 

fans, so that the workers who installed the fans received 30-50 mR, according to pocket chamber 

readings. Second, because the total time to unload a car increased to more than two hours with the 

addition of the fan installation step, the car could be on the siding for an extra day before being 

unloaded and thus people passing by and working near the car received extra exposure. Hence 

Mallinckrodt recommended that the fans be left permanently installed in the railcars (MCW 1948h). It 

is not known if AEC allowed this.

The problem was more complicated in the storage area because worker occupancy times were of 

necessity longer. There was no forced-draft ventilation in the ore storage area at that time, only a 

natural exchange of air provided mostly by a four-foot opening on either end of the building. The 

result was that radon samples likely averaged on the order of the maximum permissible concentration 

(1 x 10-10 Ci/L) or higher. These openings appear to have been permanent, since AEC (1947e) states 

that the (each) whole end of the building could be removed, but if not, a large opening still remained.

In the Scalehouse Sampling Room, samples of K-65 residue were taken from open K-65 drums sitting 

in the sampling bays on the level below the Sampling Room (MCW 1948j). This was done with a thief

sampler through holes in the Sampling Room floor. Although there were several small exhaust fans 
Document No. ORAUT-TKBS-0005 Revision No. 02 Effective Date: 06/14/2007 Page 71 of 225

that were supposed to draw the radon away from the sampling workers, the turbulence caused by a 

unit heater resulted in the radon being distributed throughout the room. Also, although most of the 

time the exhaust fans maintained adequate control over the open pans into which the drawn samples 

were deposited, at times the turbulence also drew radon from the thief and the pans as the sample 

was being dumped out. The concentrations found in the Scalehouse itself were very erratic, showing 

that here too the radon was being carried in momentary eddy currents about the room. High 

concentrations usually coincided with the pulling out of the drums on pallets from the storage bins and 

the removal of the drum lids. It was noted that a worker could be present in the Scalehouse and the 

Sampling Room for up to 4 hours on some days.

Besides the ore and residue handling and storage areas, some processing equipment was also

associated with elevated radon levels. In August 1948, a ventilating fan was installed in the wall of 

the K-65 centrifuge room to exhaust thermally hot air (MCW 1948I). However, this created a suction 

from the neighboring Wash Oliver (filter press) room into the centrifuge room, drawing appreciable 

quantities of radon off the Wash Oliver filters and into the centrifuge area. To solve this problem, a 

door was installed between the two rooms. Prior to the installation of the fan, the overall average 

radon concentration is the centrifuge room was 0.83 x 10-10 Ci/L; after the installation of the fan, it was 

1.25 x 10-10 Ci/L; and after the installation of the door it was 0.62 x 10-10 Ci/L. Further ventilation 

improvements were apparently made in the centrifuge area.

A significant short-term exposure to radon would occur when the Feinc Niagara filter press, which was 

used to produce the high-radium K-65 residue, was opened for the washing of the plates (MCW 

1950b). This occurred about 12 times a day when the processing was “operating on digest” because 

the plates could not effectively be washed down by remote control in this regime. It was necessary for 

a worker to open the press and hose down the cake. Approximately once a week, the plates became 

“blinded” (clogged) and had to be manually removed for cleaning with caustic (MCW 1950b). When 

the Niagara was opened, the radon concentration in the near area would suddenly go as high as 34 

times tolerance, then went back to tolerance level over about 5 minutes. As the last of the cake was 

washed into Tank M-289, the radon concentration would double again because of the venting of M289 as the pipe cleared and air from the tank was displaced back to the open Niagara and thus to the 

room. The concentration fell back quickly to nondetectable levels when the line to M-289 was closed. 

Radon levels at the small tank where the plates were cleaned, however, were low or nondetectable.

Radon concentration in enclosed spaces in MED/AEC 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 K65 return (drying) oven, the ore thaw house, and the Wash Oliver cell, as shown by some of the 

results in Tables A-24 and A-25 which cover the years 1947-1949 and 1950-1957 respectively. 

However, Mallinckrodt (MCW 1949m) reported the results of measurements taken in the Ore Room 

and its addition soon after ventilation and remote handling equipment was installed; some of these

results are shown in Table A-24. The improvements significantly reduced radon levels in most areas, 

but there were still some operational problems (see Section 5.2.1). Because as MCW (1949m) states 

the radon hazard was greatest near newly opened drums and because the ventilation system 

apparently was not very effective at taking care of this hazard for close approaches to the drums (i.e., 

the radon die-off to less than 1 x 10-10 Ci/L after a drum was opened took 25 minutes), radon was a 

continuing concern in the ore drum storage and handling areas. The operational problems that forced 

operators to approach open drums appear to have been corrected thereafter.

Most of the information regarding radon levels from 1947 on comes from the weekly “MZ” radon 

reports (Mz was a code for radon) (MCW 1955d; MCW 1947-1957). Additional data are shown in the 
Document No. ORAUT-TKBS-0005 Revision No. 02 Effective Date: 06/14/2007 Page 72 of 225

table below (from AEC 1949b, Table 1), which gives typical radon exposures for different areas at the 

Middlesex facility over the years 1944-1949 (covering the early years of pitchblende use). These 

would be comparable to the railcar unloading and Ore Room operations at Mallinckrodt. 

Middlesex ore storage worker radon exposures, 

January 1944–January 1949.

Area

Level

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

Radon levels are given in multiples of the radon MPC or 

preferred level of 10-10 Ci/L. Use of blower-equipped 

railcars starting in about early 1949 reduced levels to 

below the MPC.

Tables A-24 and A-25 also give information that has been found about the radon levels in the plants

other than Plant 6. However, because in the plant with the greatest exposure potential (Plant 6) and 

in SLAPS the levels were lower than tolerance (1 x 10-10 Ci/L) in all but very enclosed areas and 

because the thorium and radium were removed in the production of UO3, it is likely that the radon 

levels were well below the tolerance level in areas where ore and pre-UO3 residues were not present.

A further item regarding radon content of the residues was noted by AEC (1947e). A certain 

percentage of the radon produced during decay would be adsorbed on the surface of the ore particles 

(the percentage depending on the particle size of the ore, the temperature, the moisture level in the 

ore, etc.). When the ore was digested in acid, the adsorbed radon was released and the gamma 

activity of the ore would drop. But when the radium was precipitated as a sludge, as was done at 

Mallinckrodt, the precipitate would adsorb some of the radon and there would thus be an increase in 

gamma activity as the precipitate “aged.” This meant that the radon content of the digest vent

exhaust, the residue, etc., could be variable depending on process conditions.

ORAU (1989a) estimated the radon exposures of approximately 184 Mallinckrodt Plant 6 workers as 

part of a larger study. They calculated the working level (WL) for each job title using the mean radon 

concentration value from all available surveys in 1946-1957 for each job title’s assigned work area; 

exposure estimates were then calculated in units of working level months (WLMs) for each job title. 

ORAU (1989a) assumed that an equilibrium factor of 0.25 applied for radium and its daughters, based 

on plant and memoranda, other documents, and findings from uranium mine and residential radon 

studies. Their approach of assigning a WLM value for each job title for a single grouped period of 

years (rather than breaking it down by years) was due to the lack of survey data for many of the 

years. The “roving operator” and the production/processing/manufacturing operators were assigned 

the average area radon levels for the entire Plant 6 area and the highest daily work time breakdowns 

of all the production processes. Weekly rotation as practiced at Mallinckrodt was handled by 

assigning an average radon level of all available survey data from those work areas in the rotation 

plan for that job title. Pertinent results are shown below for information. These results are for the 

years of pitchblende processing and so would presumably be conservative for the pre-pitchblende 

years.
Document No. ORAUT-TKBS-0005 Revision No. 02 Effective Date: 06/14/2007 Page 73 of 225

Exposure rates of 184 Mallinckrodt workers, based on surveys done in 1946-1957 (ORAU 1989a).

Job title

No. of 

workers

WLM per 

month worked

Manufacturing operator 3 0.138

Operator, process develop 1 0.138

Production operator 2 0.138

Roving operator 3 0.138

Cleanup operator 8 0.127

Maintenance (general) 1 0.111

K-65/GLC sampler 1 0.102

Pot Room operator 14 0.097

K-65/AJ-4 sampler 4 0.096

Raffinate operator 9 0.092

Barium operator 6 0.091

C-3 (Centrifuge) operator 18 0.084

Feinc operator 12 0.084

Director, technical 1 0.077

Mail boy (office) 2 0.057

Porter, Production 6 0.056

Ore Room operator 16 0.055

Job title

No. of 

workers

WLM per 

month worked

Recovery operator 13 0.052

Guard 3 0.047

Porter, Lab 1 0.044

Digest operator 2 0.042

Pilot Plant operator 4 0.041

Clerk, Manufacturing Office 1 0.040

Clerk, Production Office 8 0.040

Feed operator 2 0.038

Filter press operator 3 0.038

Plant monitor 1 0.030

Health office 2 0.020

Engineer, senior mechanical 1 0.018

Furnace operator 12 0.010

Industrial relations (office) 1 0.010

Ether House operator 17 0.008

Machinist (instrument) 2 0.008

Chemist 4 0.003

Note that for use in dose reconstruction, radon concentrations reported in units of Ci/L must 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. Some reported results are given 

in Tables A-23 through A-25 of this technical basis document in the original units.

5.3.5

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. Little surface 

contamination measurement data appears to have survived, presumably because it could vary so 

much and because excessive levels, when found, were cleaned up but not necessarily documented.

Surface Contamination

Mallinckrodt (MCW 1958; MCW 1959) gave the results of a plant-wide surface contamination and 

external dose rate study that Mallinckrodt did after some preliminary 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 considered 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.
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Data from this Mallinckrodt study (i.e., the study partly documented in MCW 1958 and MCW 1959, 

quoted by Utnage (1958a)), also 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 transferable activity was measured with 

a Samson alpha survey meter, with an overall counting and efficiency factor of 20%. The data, shown 

in the table below, 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 1,000 

dpm/100 cm2 of surface contamination. 

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

Other fixed surface contamination data are given below 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 recommended AEC limits on hand contamination and on ingestion as a result of 

smoking.

A 1947 AEC letter to Mallinckrodt (AEC 1947b) stated that surfaces in the lunchroom should show 

less than 100 dpm alpha on a smear test and less than 500 dpm alpha per 100 sq cm (fixed plus 
Document No. ORAUT-TKBS-0005 Revision No. 02 Effective Date: 06/14/2007 Page 75 of 225

Contamination levels on various surfaces and pieces of 

equipment in the laundry.

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

removable); the maximum hand count should be equivalent to no more than 0.35 mg of uranium on 

either side of the hand. In 1949, dust samples taken in the lunchroom at various times over one lunch 

period, showed that the level ranged from 6 dpm per cubic meter with only 7 workers present to 57 

dpm per cubic meter when the room was full; the level was 12 dpm per cubic meter at the start of the 

post-lunch cleaning and then went up to as high as 65 dpm per cubic meter following the cleaning due 

to dust suspended in the air (MCW 1949d). The same sampling study showed that levels in a 

conference room in association with a service group (operators’) meeting were 4, 22, and 30 dpm per 

cubic meter respectively before, during, and just after the meeting. In both the lunchroom and 

conference room, the workers wore cover clothing over their work clothing. In 1951 it was reported 

that the average dust level in the Plant 6 lunchroom (MCW 1951b) was 1.3 dpm/m3

. Thus although it 

was stated in MCW (1949d) that the sub-tolerance levels reported were acceptable, clearly the 

lunchroom levels were reduced over time by better clothing change and contamination control 

practices.

Smoking areas were typically adjacent to the working areas. In 1951, in response to AEC’s pointing 

out that Plant 6 workers entering the smoking room habitually slapped their dusty gloves and brushed

material from their clothing after they entered the smoking room, Mallinckrodt agreed to provide 

improved ventilation in the smoking room and to put a small vacuum cleaner inside the smoking room 

for workers to use to vacuum their gloves and clothing after they entered but before they smoked 

(MCW 1951d). Mallinckrodt also agreed to have greater enforcement by supervisors of the 

requirement that gloves be removed before smoking (MCW 1951d).

These findings may explain the observation by AEC (1951a) that in a majority of the jobs evaluated at 

Mallinckrodt, the contribution to radioactive dust levels from time spent in the smoking (cigarette 

break) areas was greater than the contribution from the time spent in the operations area. In 1951 it 

was reported by Mallinckrodt that air samples taken in the smoking room in Plant 6 were always 

above tolerance, even after UO2 production had moved to Plant 7 (MCW 1951b); the average dust 

level was 116 dpm/m3

, versus 1.3, 4.2, and 66 for the lunchroom, locker room, and change room 

respectively, and the typical daily individual occupancy of the smoking room was 60 minutes. This all 

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.

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%. 
Document No. ORAUT-TKBS-0005 Revision No. 02 Effective Date: 06/14/2007 Page 76 of 225

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.). This 

suggests that ingestion is a negligible source of internal dose during operations, compared to 

inhalation.

Regarding clothing, Table A-27 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 almost 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). Gloves were discarded when soiled and were treated as 

contaminated clothing (MCW 1950e).

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 to 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 1949b). AEC advised that strippable coatings would eventually 

need to be used (AEC 1949b), 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.
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Monitoring for surface contamination was done using portable alpha and gamma-beta survey meters 

and smear tests (MCW 1950e). Tools to be moved from contaminated areas were to be cleaned to a 

stated level (MCW 1950e), given in 1947 as 100 cpm, presumably beta-gamma (AEC 1947b). If the 

article could not be decontaminated, it was to be securely wrapped in paper or placed in a clean 

container and marked with handling instructions. However, these checking and control requirements 

did not apply to tools belonging to outside contractors, who presumably could take away tools 

contaminated to any level (MCW 1950e). Alpha hand counting devices were not used after some 

time prior to 1950, supposedly because AEC-NYOO requested that their used be discontinued (MCW 

1950e). As of at least 1950, hands and feet were not checked when workers exited a process area to 

go to the lunchroom to eat or smoke (MCW 1950e; AEC 1947a), but they had to go through the 

personnel decontamination cleanup in the change room area, with a strict “clean side, dirty side” 

regime maintained (MCW 1950e) (see Section 5.5.3). Visitors were not checked for contamination 

before leaving the plant, at least through about 1950 (MCW 1950e). 

5.3.6

Mallinckrodt uranium processing workers were given a pre-employment physical that included an 

initial urinalysis and a blood count (MED 1942) and they were given an annual physical that included 

a routine urinalysis and a blood count (MED 1942 (which says it was a “followup” examination); MCW 

1955d; Mason 1958a). From about the summer of 1948 on (MCW 1950c), 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 1955d; Mason 1958a); after this, time the frequency was no more than 

semiannual (MCW 1955d). 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 Mallinckrodt radioactivity urinalyses were first performed by the University of Rochester under 

contract to AEC-NYOO up to about 1948. At one point, according to ORAU (1983b), there were some 

problems with the running of standards because there had been uranium plateout on a table. This 

was corrected. In 1948, Barnes Hospital began to do the urinalyses for Mallinckrodt. However, an 

AEC health official stated (AEC 1948) that while it had been his understanding that the analyses were 

being done at Barnes Hospital (at Washington University), 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 have been

potentially contaminated (i.e., some of the uranium content may have come from the laboratory 

analyzing them).

Apparently Barnes Hospital resumed doing the urinalyses (AEC 1949g, MCW 1950e). However, in 

1949, AEC compared the Mallinckrodt analyses against those for other sites handling similar material 

and concluded that the results were consistently inexplicably high (MCW 1950e), although 

Mallinckrodt thought the agreement was acceptable (MCW 1949j). AEC then sent Mallinckrodt some 

spiked samples and also had an independent Mallinckrodt chemist prepare a stock solution of known 

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 to standard ones. The conclusion was that the 

samples were indeed reading too high at Barnes. (It is not clear if the spiked sample comparison 

referred to in MCW 1949j is this same set of samples, but if so, the data can be found in MCW 1949j, 
Document No. ORAUT-TKBS-0005 Revision No. 02 Effective Date: 06/14/2007 Page 78 of 225

along with a statement that the reliability of the Mallinckrodt sample testing is +/- 15%). 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 were of doubtful value (AEC 1950l). Still, AEC 

(1950i) 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 (1950q) 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. 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 

1950l).

It is not clear who did the urinalyses from 1950 to 1954; although MCW (1950m) suggests this was no 

longer being done at Barnes Hospital but at AEC-NYOO, presumably at HASL. Also, in September 

1951 AEC-NYOO was clearly performing them, but directed AEC St. Louis area representative not to 

have any more routine samples sent during the last quarter of 1951 because of an “unusual work 

load” that its chemical laboratory had (AEC 1951c); emergency samples could still be sent, however. 

In 1954 AEC gave Mallinckrodt permission to perform their own urinalyses (MCW 1954e, ORAU 

1983b), 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 AECNYOO (HASL) must have resumed performing the analyses, because a 1955 Mallinckrodt report 

describing its health program stated that NYOO was doing so (MCW 1955d). This report stated 

NYOO was analyzing about 2500 Mallinckrodt urine samples a year and urine samples that were 

taken were split, with half going to AEC-NYOO for the radiological analysis and half to Barnes 

Hospital for the medical analysis.

Referring to urine samples sent to the University of Rochester from any AEC site, ORAU (1983b) 

stated that samples were collected from workers in four-ounce glass bottles with bakelite caps and 

shipped to HASL. After 1949, only samples taken within one hour of the start of the weekly shift were 

accepted (ORAU 1983b).

The radiological analysis was apparently only for uranium content (referred to as “X in urine” or 

“uranium-in-urine”). However, one set of urinalysis records shows for a single individual and for 

periods from March to May 1956 the notation “Shoot for radium.” Thus it appears that radium in urine 

could be analyzed as a special case. It is not clear whether the thorium (ionium) processing workers 

in Plant 7E had special urinalyses for thorium or not. However, this is a possibility, as suggested by 

Mallinckrodt’s health and safety director’s obtaining a copy of a Los Alamos Scientific Laboratory 

procedure on urinalysis for plutonium (LASL 1955) and the director’s noting on his copy that he was 

told that urinalysis for thorium could be done by the same method.

ORAU (1983b) stated that all urinalyses done at the University of Rochester used the fluorometric 

method; the urine was dried in a platinum dish, then fluxed with either sodium fluoride or a lithiumcalcium fluoride mix and counted. 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. AEC (1949l) states that Barnes Hospital (at 
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Washington University in St. Louis) used “fluorometric (photoelectric ultraviolet)” methods in 

performing urinalyses for Mallinckrodt.

ORAU (1983b) states the following about samples analyzed at the University of Rochester. These 

samples were run in triplicate, with the results usually being within ± 2 μg/L of one another; if this 

turned out not to be the case, it indicated that there had been poor fusion of the flux and the samples 

were re-fused and re-run, which usually corrected the discrepancy. The value recorded was the 

median value of the three. For insoluble uranium, it was considered that 30 μg/L in the urine 

corresponded to an air concentration of 50 ug/m3

. The analysts felt confident of readings greater than 

or equal to 5 μg/L. However, if more confidence was desired for lower-level samples (e.g., for special 

projects), the urine was concentrated either by ion exchange or by extraction, or more aliquots were 

run.

The uranium fusion photofluorimetry urinalyses performed by the University of Rochester and the 

AEC NYOO were similar to those performed at other AEC facilities. Based on the lack of clear 

information regarding a detection threshold for Mallinckrodt urinalyses, the default detection threshold 

for uranium urinalysis may be assumed to be 10 μg/L based on a reported sensitivity of 5 to 10 μg/L 

for uranium fluorimetry urinalysis in the early years, as was reported for other AEC facilities (Wilson 

1958).

Because of the questions regarding the validity of the samples, the apparent variations in sample 

analysis methods, and even who was doing the analyses, the Mallinckrodt urinalysis data should be 

used with caution, at least when the data were taken by Barnes prior to about 1951. However, it 

appears that the errors, if any, are in the conservative (high) direction and thus are favorable to 

claimants. 

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 the Center for Epidemiological Research’s (CER) computer database 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) [ORAU 2005]. Some bioassay records are also available in the Project 

site research database (SRDB).

5.3.7

Breath radon measurements began to be made in June 1945 (as shown by, e.g., MED 1945e and 

MCW 1946g), but only for workers who worked in areas where there was a potential for radium intake. 

MCW (1950I) gives a list of worker types or occupational areas, along with a rating of the need for 

workers of each type or in each area to have their breath sampled; most of the Manufacturing 

(process), Stores (warehouse), and Maintenance workers were in the “Definite” category, while most 

of the Power Plant, Occupancy, Plant Protection (guard), Research, Laboratory, and Laundry workers 

were in the “Possible” category and most of the Health & Safety, Office & Administration, and Plant 4 

workers had most of their people in the “Improbable” category.

Information and Available Data Regarding Other Types of Bioassay

The Massachusetts Institute of Technology (MIT) or the National Bureau of Standards (NBS) did the 

breath radon analyses for MED/AEC until 1951, when HASL obtained an ion chamber of the type 

needed to perform the analyses (ORAU (1983b). Initially, samples were taken in the infirmary of the 

MED/AEC site because it was thought that this area was removed from the production areas; later a 

respirator with bottled air was used to collect samples in order to control for radon in the ambient air 

(ORAU 1983b). However, the first breath radon measurements were thought to be worthless by MED 
Document No. ORAUT-TKBS-0005 Revision No. 02 Effective Date: 06/14/2007 Page 80 of 225

because the MED expert concluded that the samples were contaminated at the time they were taken 

(MED 1945e). As a result, MED directed Mallinckrodt to have breath radon samples taken only in a 

room away from the plant and only when the employee had not worked for at least 48 hours and while 

he was still wearing his street clothes (MED 1945e).

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 1949g; AEC 1950a; MCW 1950g). The 

samples were measured at the NBS (AEC 1949g) or at HASL (AEC 1950a, MCW 1955d) by an 

“automatically recording pulse-counting device” (AEC 1950a). If a sample was over the tolerance 

level of 1 x 10-12 Ci/L, then a recheck (repeat sample) was made immediately (AEC 1949g). The 

lower limit (of detection) at HASL was 0.1 pCi/L and there was confidence in readings of greater than 

or equal to 0.5 pCi/L (ORAU 1983b). In 1946, it was reported that of 40 workers tested, the maximum 

level was right at the tolerance level of 1 x 10-12 Ci/L (MCW 1946g).

In early 1950 AEC apparently became concerned about the high background that seemed to be 

present where the samples were being taken (MCW 1950g). 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 

sampling area (AEC 1950k). 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 1950m); if the breath radon samples were also taken there, that could 

explain the relatively high background radon. In early 1951, Mallinckrodt (MCW 1951g) reported the 

results of a further test done at the behest of AEC in which they took breath radon samples from the 

same four workers in the usual location (background of 0.8 pCi/L) and in a lower-background area 

(background of 0.2 pCi/L). Mallinckrodt admitted that although they had been told earlier by the 

(National) Bureau of Standards that the test area background did not matter as long as it was on the 

same order as the sample, AEC had turned out to be correct in its concerns about the background: 

the differences between samples taken in the lower-background area and the samples taken in the 

higher-background area differed by almost exactly the difference between the two backgrounds 

(MCW 1951g). Mallinckrodt apparently corrected this problem by then using the special respiratortesters provided by AEC ((MCW 1951e); they may also have moved the breath radon sampling 

location to a lower-background area. Note that Srivastava et al., in their study of uranium miners

(Srivastava 1986), found that post-shift measurements (i.e., after breathing ambient radon in the 

workplace) were 2 to 4 times higher than pre-shift measurements, which confirms the confounding 

influence of background radon.

A baseline (start of employment) breath radon test was not administered; the routine test during 

operations was thought to be sufficient (MCW 1950w). In 1949, the nominal breath radon sampling 

frequency was quarterly (AEC 1949g), which seems to have been the typical frequency for eligible 

workers (ORAU 1983b). 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 1955d). It was not always possible to obtain a termination breath radon sample because some 

workers left without notice or because the sample had to be taken after 48 hours of non-exposure, 

i.e., usually on a Monday (MCW 1950w). Mallinckrodt thought that about 80% of terminating 

employees had termination medical examinations (MCW 1950w), but it is not clear if this was also 

true of the breath radon analyses for the relevant workers. Breath radon records are available in 

scattered form in reconstruction project files; possibly all of it is found in the large urinalysis research 

file mentioned in Section 5.3.6 above.
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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). ORAU (1989b) quoted a report by the MIT Radioactivity Center that stated that their 

lower limit of detection for breath radon was 0.006 pCi/L and that this corresponded to a minimum 

detectable body burden of radium-226 of 0.00059 μCi. ORAU (1989b) also quotes the MIT report as 

stating that assuming that 65% of the radon generated by the body burden of radium was exhaled, 1 

pCi/L of radon in the breath corresponded to approximately 0.1 uCi of Ra-226 in the skeleton, which 

ORAU (1989b) verified in its own calculation. This can be compared to the assumptions made by 

Srivastava et al. (Srivastava 1986), who took breath samples of uranium miners using a controlled 

radon intake respirator apparatus: the equilibrium percentages of the radium-226 body burden were 

taken to be 33% in the lungs, 39% in cortical bone, 14% in trabecular bone, and 14% in other soft 

tissues and the corresponding radon escape fractions were taken to be 1.0 for lungs and other soft 

tissues and 0.7 for bone. Srivastava et al. thus concluded that 84% of the radon produced by decay 

of the Ra-226 was exhaled, a figure applicable to those working in the mines for several years (i.e., to 

workers chronically exposed for an extended period).

ORAU (1989a) reported the results of a study of ambient radon exposures to some 184 Mallinckrodt 

workers. In addition to the lung doses from the ambient radon exposures, they attempted to estimate 

the contribution from breathing radium dust and used breath radon data and the associated skeletal 

radium burden as an indicator (as per ORAU (1989b). These estimates are given in ORAU 1989a, 

and as with the ambient radon exposures, the subjects are listed only by their ORAU study 

identification numbers.

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 and individual whole body and lung count data (if any) appear to be unavailable. 

5.4 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 1949b; AEC 1951b)). ORAU 

(1980b) stated there did not appear to be any circumstance that would produce an exposure of 3 rem 

or greater in less than several hours and they found no incident in which an acute exposure rate of 

greater than or equal to 3 rem/hour was produced. The tables below are an adaptation of Table I and 

a summary of information given in Tables II of ORAU 1980b; the first shows the total annual gamma 

dose by dose-level and the second shows the ten highest gamma doses to workers, both from 1947 

through June 1957.

Information regarding dose rates for ore and residue drums is given in Table A-32 and regarding dose 

rates for various forms and areas in Table A-33. There is little information about conditions in Plants 1 

and 2 during the wartime startup; however, some dose rate measurements from 1943 and 1944 

appear to have been taken, as shown in Table A-33. As noted previously, film badging did not start 

until late 1945, when Plants 1 and 2 were in the process of largely shutting down. Doses might have 

been somewhat higher than in Plants 4 and 6 due to possible greater manual involvement and 

probably somewhat greater bodily proximity to sources, but on the other hand the quantities involved 

were much lower. The dose rates given in Table A-33 suggest doses were likely to have been higher 

in later years due probably to the greater volume of uranium processed and to the use of pitchblende 

ores.
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Annual gamma exposures, 1947 – June 1957 (ORAU 1980b).

Year

Workers

monitored

Annual gamma exposure, R

0 - 1 1 – 5 5 - 10 10 - 15 >/= 15

1947 253 70 131 27 18 7

1948 366 120 171 47 19 9

1949 554 370 141 41 2 0

1950 615 475 133 7 0 0

1951 694 512 171 11 0 0

1952 757 659 88 10 0 0

1953 763 619 142 2 0 0

1954 756 566 188 2 0 0

1955 871 766 105 0 0 0

1956 958 944 14 0 0 0

Ten highest gamma doses to workers, 1947 – June 

1957 (ORAU 1980B). 

Year

Annual dose, R Weekly dose, R/week

Min Avg Max Min Avg Max

1947 14.4 16.1 23.5 0.28 0.31 0.45

1948 14.9 17.0 20.3 0.25 0.33 0.39

1949 7.7 9.0 13.3 0.15 0.17 0.26

1950 4.5 5.4 7.1 0.09 0.11 0.14

1951 5.0 5.9 7.1 0.10 0.11 0.14

1952 5.1 5.9 6.6 0.10 0.12 0.13

1953 4.0 4.6 5.7 0.08 0.09 0.11

1954 3.9 4.4 5.1 0.08 0.09 0.10

1955 3.9 4.4 5.1 0.08 0.09 0.10

1956 1.1 1.4 1.9 0.04 0.06 0.07

It should be noted that the era of pitchblende use (early 1945 on) was mostly covered by film badge 

monitoring, so the doses characteristic of this work are known. Thus it is considered to be favorable 

to claimants to assume the doses at Plants 1 and 2, for the same type of work, were similar to those 

at Plant 4 and Plant 6 around 1948.

According to MCW (1955f), 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. Although these 

reports do not appear to be available, Table A-33 shows some area film badge monitoring results that 

show what conditions were and of course film badges were consistently worn in the later years. A 

1955 list of Plant 6 process cells, dust collectors, and tanks, together with the then-current uses of the 

tanks, appears in MCW 1955g; maximum emergency non-survey stay-times (i.e., for urgent access 

and not inspections) are also given. It appears that the stay-times were based on external dose rates. 

In general, the operator stay-times were half the maintenance stay-times, presumably because of the 

need for more frequent access by the operators.

See also Sections 5.1 – 5.2 above for some addition external dose information associated with 

particular handling activities or operations.

5.4.1

After high-grade pitchblende ores began to be used, refinery workers were exposed to photons from 

radionuclides in equilibrium with U-238 and U-235. Ra-226, through its Pb-214 and Bi-214 daughters, 

contributed 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 

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

In 1946 the principal large-scale source of gamma radiation was said to be the drums of ore as they 

were stacked in the receiving warehouse (MED 1946c). 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). 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. Dose 

rates from wastes could run even higher than those for the ore. 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) or up to 275 mR/hr at contact (MCW 

1949g). AEC (1948d) gave the gamma contact dose rate with the (Ra-containing) Feinc filtrate 

residue under equilibrium conditions as over 300 mR/hr; however, they stated, they had no way of 

knowing how close to equilibrium it was. The dose rate depended strongly on how “aged” the material 

was: sample pans filled with fresh K-65 read 9 mR/hr, while pans held over a day read 20 mR/hr 

(MCW 1948j).

A study done by AEC of the dose rates from four drums of “aged” K-65 residue grouped on a pallet 

showed the dose rate at 6 feet from the grouping was 28 mR/hr, of which the back two contributed 

16% and the front two 86% (AEC 1947c). Furthermore, it was found that the dose rate varied 

approximately as the square of the distance over the range of three to twelve feet from the center of 

the grouping. With lead shielding interposed, it was found that the first half-value layer (HVL) was 1/4” 

and the second was 3/8.”

As noted in Sections 4.7 and 5.2.3, K-65 was brought back in drums from SLAPS and reprocessed 

starting in early 1948. In January 1949, the gamma dose to the hands of operators opening these 

drums was studied by placing film badges on their wrists (MCW 1949E). It was found that the gamma 

dose varied from 3.3 to 5.0 mR per drum opened, with an average of 4.0 mR; for the typical situation 

of 44 drums opened per 24-hour day, six days per week, by three operators, the total weekly 

exposure per operator was 352 mR. It was estimated that this could increase by up to a factor of 2 

due to drums that might require more time than average to open, variations among operators in 

opening the drums, and in the number of drums opened per shift. Hence the maximum wrist dose 

was estimated to be 700 mR per week per operator. At the time, the individual whole-body gamma 

dose to the operators was 300-500 mR per week; it was thought that 33-67% of this dose was due to 

the drum opening (MCW 1949e).

Additional specific information regarding gamma doses and dose rates in the ore, refinery, and metal 

processing areas are shown in Tables A-32 and A-33: dose rates from drums and railcars are shown 

in Table A-32 and for various plant locations and tasks in Table A-33. Additional gamma-beta dose 

data for ore storage at the Middlesex plant during 1944-1949 is shown below (AEC 1949b, Figure 3); 

this should be comparable to Mallinckrodt experience because presumably the ore was being routed 

through Middlesex to be distributed to the processing sites. AEC (1949b) states that the Middlesex 

doses were about 65% gamma.
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Middlesex ore storage worker dose, 1944-1949 (AEC 1949b, 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

It should be noted that operations that were particularly manual were (1) the various dumping, 

scooping, and scraping operations in which feed, UO2, UO3, UF4, and dust were handled or crucibles 

and furnaces were cleaned; (2) the “plowing” (scraping) of the centrifuges; and (3) the scraping of 

cake off the Feinc filter cloths (this was the pitchblende cake during the pitchblende years) and the 

changing of these cloths. Thus significant external dose reduction usually followed any 

mechanization or improvement of these processes.

Because the gamma dose arose mainly from the radium and its daughters, the gamma dose was 

usually significant only in those areas where the source material had not yet had the radium 

separated or where radium-bearing residues were present. This meant that the gamma doses tended 

to be highest in Plant 2 and later in Plant 6 (AEC 1949b), especially around ore drums and storage 

areas for the radium-bearing residue, K-65. Thus some shielding had been designed into Plant 6 and 

more was added in 1948 in some areas (AEC 1949b, MCW 1950e). Still, due to the high doses 

Mallinckrodt found it necessary in mid-1949 to establish additional restrictions and rotation 

requirements on warehouse workers, who moved stored pitchblende ore drums from the airport waste 

storage site (apparently also an overflow ore storage site); brought “sand” (residue) drums back from 

the airport; unloaded the incoming ore drums from railcars; weighed the ore drums coming in from the 

railcars and airport site; and loaded K-65 residue drums into the railcars (MCW 1949p). It was 

estimated, e.g., that the following were typical weekly activities: three trips for retrieving ore drums 

from the airport pad, three or four drum weighing sessions, one trip for retrieving residue drums, two 

K-65 railcar loadings, and two ore railcar unloadings (MCW 1949p). It was required that pocket 

chambers be used when retrieving K-65 drums from the airport pad (MCW 1949p).

The highest principal Feinc cloth operators’ dose had been found to be an average of about 465 

mrep/week (total beta and gamma) in 1949 and 1950 and the second-highest dose was similar (MCW 

1951g). In January 1951, a study of beta and gamma exposures associated with preparing, repairing, 

cleaning, and changing cloths for the Feinc, C-3, and Recovery filters (MCW 1951a) found that 

operators received an average of 294 mrep per day and 1470 mrep per week gamma and an average 

of 82 mrep per day and 408 mrep per week beta. (See Table A-33 for some of the measured dose 

rates.) The “per day” figures were an effective daily dose since the operation was not actually 

performed daily. These figures were comparable to those found in a previous 1949 study; although 

some of the tasks the cloth operators were doing in 1949 had been given to the area operators to do 

(MCW 1951a).

Mallinckrodt (MCW 1950t) summarized the gamma dose data for all high-dose workers as in the table 

below. Although it would appear from the disparity between the two averages below that a few 

individuals were getting significantly higher doses than the rest, Mallinckrodt contended that the doses 

were fairly evenly distributed among workers in the five rotation groups and thus that all dose 

reduction that could be accomplished using rotation had been accomplished. For comparison, group 

data reported by AEC (AEC 1951b) for two periods in 1948-1949 and 1950 are shown.
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Comparison of gamma and beta doses, 1946-1950.

MCW 1950t

Date

Avg of high 

badges, mR or 

mrep

Avg of 90th percentile 

badges, mR or mrep

Gamma only 1946 780 250

1947 860 290

1948 590 250

1949 410 170

1950 (1st quarter) 320 ---

Gamma plus beta 1950 (1st quarter) 700 235

AEC 1951b

Period

Number over 30 

mrep/week

Number over 150 

mrep/week

Number over 

300 

mrep/week

Gamma only, Plant 6 11/1/48 – 1/24/49 (267) 103 47 17

1/2/50 – 6/19/50 (314) 124 47 11

Beta only, Plant 4 11/1/48 – 1/24/49 (91) 83 35 32

1/2/50 – 6/19/50 (89) 65 29 12

The AEC 1951b figures are weekly averages. The number in parentheses after the period is the total number of badges worn per 

week, on average. The beta figure for “over 300 mrep/week for 11/1/48-1/24/49 included 8 over 700 mrep/week, 5 over 1000 

mrep/week, and 2 over 1500 mrep/week; the corresponding inclusions for 1/2/50-6/19/50 were 2, 1, and zero mrep/week.

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 A-4 and A-6).

Doses registered on film badges worn by people not working directly with the uranium and equipment, 

such as guards and office workers, were 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 

radium content in any uranium product or residue.

A 1958 AEC report on uranium mills gave 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 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 1948e). AEC (1948d) gave the beta contact dose rate with 

the (radium-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.

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) pointed out that 90% of equilibrium beta 

activity was restored by 90 days after vacuum casting. Eisenbud (1975) reported 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 had 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 
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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 to extract the 

uranium from uranyl nitrate, Th-234 and Pa-234m (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. MED (1942) stated that 1942 

measurements indicated that the intensity was low and that no precautions needed to be taken for 

disposal; however, MED/AEC appears to have been more concerned about this later on. Another 

source of the tails was the UO2-derived shotgun sample, which could have the Th-234 and Pa-234 

concentrated to 30-300 times their activity in normal uranium metal in equilibrium, depending on how 

long the UO2 had stood between production and sampling (MED 1944m). The fourth and final ether 

extraction performed in processing a shotgun sample produced a liquor so concentrated with these 

beta emitters that it was said the tolerance dose of beta radiation could be reached by keeping the 

hands above the liquor for 10 minutes per day. Besides that, the chemist handled the sample for 5-10 

minutes from removal from the furnace to bottling, wearing no gloves and directly touching the 

containers. MED (1944m) advised changing from rubber gloves to leather gloves (to increase the 

dose rate reduction from a factor of 2 to a factor of 3) and using crucible tongs (to increase the 

distance to about 10”) for conveying the evaporation dish to the heating areas.

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 1949b; Eisenbud 1975). 

The impurities contained Th-234 and Pa-234, which were concentrated to a significant degree in the 

deposits (AEC 1949b; Eisenbud 1975); this deposit residue could have “up to 1000 times the beta 

activity of natural uranium” (AEC 1949b). 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 1949b), 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). Mallinckrodt (MCW 1949a) observed that 25% of Plant 4 workers received over 500 

mrep/week beta and 3-6 workers per week received 2000 mrep or more; AEC (1949g) also observed 

that the beta values (on film badges) from Plant 4 consistently ranged up to 2.7 rep/week.

Because of the high hand doses, the processing of the derbies and billets was studied in 1948 by 

Mallinckrodt (MCW 1948b). They gave the results to the designer of the new Plant 6E for use in 

redesigning the process to reduce the doses, first by eliminating exposure to large open surfaces on 

uranium forms and second by preventing exposure to and accumulation of scale and powdered 

residues (in which the UX1 and UX2 were concentrated), especially in the recasting or remelting step.

Regarding the processing of residues to concentrate thorium, Table A-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 

Section 5.3.5 and in Table A-27 (from the text and Table 1 of Utnage 1958b). This is mainly beta 

radiation.

MED did some studies to determine the shielding afforded by gloves and clothing. MED (1943c) 

reported that leather gloves of 1 mm thickness cut the beta dose rate by a factor of 2, while 2-mm
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gloves cut the beta dose rate by over a factor of 4. MED (1944i) gave the reduction factor as 2 in a 

1944 study of rubber gloves, with the source being uranium metal in equilibrium. MED (1944a)

reported the results of July-September 1944 measurements of the activity inside 16 pairs of gloves; 

the results showed that contamination did appear to get into gloves used at the recast furnaces but 

not in those used for billet sawing. This implied that some of the recast dose to the hand came from 

contamination inside the gloves. Based on all these results, MED (1944i) gave the time limit for 

handling uranium metal as 4 hours per day with rubber gloves and recommended that the time be 

limited to 2 hours per day with other gloves.

AEC did some further glove and clothing shielding and contamination studies using an 18” x 24” sheet 

of uranium metal in equilibrium with Th-234 and Pa-234 (AEC 1950i), 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 shielded 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. (These figures are included in Table A-27.) The 

Mallinckrodt glove program for contact with radioactive material was said to be sketchy and 

inadequate (MCW 1955d), implying that the use of gloves was not consistent.

AEC (1949g) stated the following regarding pocket meters (pocket chambers). These were typically 

worn two at a time by a worker for an entire work day before recharging. The readings of the two 

were averaged and the averaged reading was tabulated with the film badge results. The pocket 

meters were vibration- and leakage-tested before first use, then calibrated for 100 mR +/- 5 mR fullscale and 50 mR +/- 10 mR half-scale. Mallinckrodt suggested discontinuing the use of the pocket 

meters in 1949 (AEC 1949g, MCW 1949f). In MCW (1949i), they argued that the highest possible 

(gamma) dose rate in their plants was 300 mR/hr, that exposures of greater than 1000 mR per week 

were almost unknown anymore, and that operations were such that extended occupancy of highdose-rate areas was not necessary. They stated they had not found any dose-heavy sources or 

operations that could not have been found by the use of film badges or surveys; also, their latest 

statistical calculations showed a greater than 99.9% probability of a true correlation between film 

badge results and pocket meter readings, but the error at the 90% limits around the regression line 

gave an error of +/- 38% for a dose of 300 mR per week. Thus, they concluded, using both film 

badges and pocket meters constituted an unnecessary double check of dose. AEC at first refused to 

allow them to discontinue using the pocket meters (AEC 1949g), presumably because it was the only 

real-time check and the only immediate post-work check of accumulated dose available, but then 

relented and allowed such use to be discontinued (AEC 1949h; MCW 1950e).

Detector types and measurement methods were not specified in most reports and papers. However, 

it is known that there were numbered, set survey points (“observation stations”) at which detector 

measurements were always made (AEC 1949g). MCW (1946g) stated that external dose rates were 

measured with an “Ion-meter.” AEC (1949g) stated that one Mallinckrodt health physicist preferred 

the Victoreen 247 gamma survey meter and the Landsverk electrometer, while the other preferred the 

Zeus meter (detector) for general use. 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 beta-gamma meter with a thin-wall tube were used. 

AEC (1949a) gave information regarding the Rauland Zeus detector used by Mallinckrodt e.g., the 

window area and the accompanying table of data. However, the conclusion was that the free area of 

the window was only slightly greater than 50%. This information was supplied to Mallinckrodt in the 

context of comparing film measurements and detector measurements, with the conclusion that the 
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assumed effective size of a Mallinckrodt source (such as a tank or contaminated area) might be 

greater than had been assumed.

Counters and meters were maintained and calibrated weekly or monthly on a set schedule by 

Mallinckrodt technicians (MCW 1955d). Because of the lack of space and the corrosive atmosphere 

in the plant, there were no instrument monitors (i.e., instantaneous, continuous, or integrating 

monitors) in any area except for one provided by AEC on an experimental basis (MCW 1950e). 

Hence film badges were placed at selected locations in the process areas to serve as integrating area 

monitors (see Section 5.4.3.6 below).

5.4.2

No neutron exposure measurements are available. However, Dupree-Ellis et al. (2000) deemed 

neutron exposures at Mallinckrodt to be minimal. This conclusion appears to be correct based on 

what is known about the source material and its handling. Nevertheless, neutron production by the 

alpha-neutron reaction and by the early RaBe source used in the Shotgun Laboratory (see Section

5.2.4) was analyzed in the preparation of this technical basis document. Note that this document 

does not currently consider spontaneous fission neutron dose, which might add to dose in areas 

where large quantities of uranium are present.

Neutrons

In the analysis of neutron production by the alpha-neutron reaction, the forms of uranium and thorium 

that would produce neutrons at the highest rates were identified as UF4 and ThF4, the latter being an 

intermediate product in the processing of the thorium-containing waste cake to a thorium nitrate 

solution (see Sections 4.5 and 5.2.3). The uranium oxide forms, soda salt (Na2U2O7), and thorium 

nitrate were also considered, as indicated in Table A-34. Note that as long as there is a reasonable 

amount of target material (fluorine, oxygen, or sodium) available and it is intimately mixed with the 

source material (uranium or thorium or their alpha-emitting daughters), neutron production essentially 

depends on the amount of the source material. 

Regarding the thorium, the total amount of cake processed to obtain the thorium nitrate solution was 

350 tons, or 700,000 lbs (Table A-4). It is not known how much was processed at a time. However, it 

is conservative to assume that the processing was done in the same type of large batch-processing 

tank system used for processing ore (this maximizes the quantity at one time). This is also 

appropriate because the cake was digested and subject to extraction in the same manner as the ore 

and soda salt feed. From Table A-4, at one time a typical monthly processing would include about 

80,000 lbs of Eldorado black oxide, 120,000 lbs of Vitro black oxide, and 60,000 lbs of Vitro soda salt, 

for a total of 260,000 lbs per month or about 8700 lbs/day. Also from Table A-4, K-65 ore was 

processed at another period at the rate of 8000-12,000 lbs/day. Thus it is appropriate to assume that 

a reasonable high average processing volume of feed per day was 10,000 lbs. At this rate, it would 

take 70 workdays to process the cake.

It can be assumed, from the way the ore and other feeds were processed, that one batch was moved 

from the digest tank through etherization, etc., without being mixed with another batch, at least until 

after the ether extraction. Thus it can be concluded that no more than 1/70th of the total Th amount 

was present as ThF4 at any time. This is conservative because it assumes that the amount processed 

in one day was together in one tank or container at a time. It is also conservative because the 

processing is known to have taken up to 18 months, not just 70 workdays, so the batches are likely to 

have been much smaller and the amount of ThF4 present at any time in a container would also have 

been much smaller. This is particularly likely since the ThF4 was said to have been moved to the Hot 

Lab (in Plant 6) in liquid form for final processing to thorium nitrate (AEC 1955c) and thus the

individual transport container quantity was limited by the amount that could be transported as a liquid. 
Document No. ORAUT-TKBS-0005 Revision No. 02 Effective Date: 06/14/2007 Page 89 of 225

Finally, it is conservative because any self-shielding or container shielding has been ignored. The 

spontaneous fission half-times for Th-232 and Th-230 are greater than 1E+10 times greater than their 

physical half-lives (BRH 1970), and so neutron doses from thorium spontaneous fissioning are not 

significant compared to those calculated from the alpha-neutron reaction in materials with low atomic 

numbers. 

Regarding uranium forms, as shown in Table A-4, a fiber or steel drum container of UO3, UO2, or soda 

salt would weigh 75 lbs and a steel drum container of UO2O8 (ore feed) would weigh 100 lb; most of 

this weight would be uranium, so it may be reasonably and conservatively assumed that the entire 

weight is uranium. While a larger volume could be found in, e.g., a digest tank, the liquid and the thick 

tank wall would provide a great deal of shielding. A larger volume could be found in an array of 

containers, but a great deal of self-shielding would be involved and a person would likely not spend a 

great deal of time near an array. Thus it is likely that that dose rate from a single container (being 

temporarily stored, loaded, or transported) will be the typical dose source.

The following conservative assumptions were made regarding the conversion from alpha-neutron 

production rate in the container to an annual dose at the receptor point.

1. Point source geometry used to produce a nominal ambient dose rate, with the distance to the 

receptor point being taken from the center

2. Irradiation geometry of 75% AP and 25% ROT (from Table A-18 of this technical basis 

document for “ionium plant operator”) 

3. A U-238/U-235 breakdown of 99.3% to 0.72% by weight (for the natural uranium forms); a Th230/Th-232 breakdown of 11.6% to 88.4% (for the thorium forms)

4. Neutron energy of 1.5 MeV (1.4-1.6 MeV is the energy range of the neutrons produce by the 

thorium isotopes and 1.0 - 2.0 MeV is the energy range of neutrons produced by the other 

isotopes; also, the neutron flux-to-dose conversion factor varies slowly in this range)

5. Neutron flux-to-dose conversion factor for 1.5 MeV of 1.3 x 10-4 rem/hour per neutron/cm2

-sec

6. 1 hour per day spent by the receptor at 1 foot from the container and 3 hours per day were 

spent at 3 feet, every working day for a year

7. For the purposes of considering the effect of including daughter contributions, for full 

equilibrium of the daughters of natural uranium or the thorium cake mix down to radon (which 

would not be in chemical union with the target and would likely be largely vented), the neutron 

emission rate due to the parent isotope would be multiplied by the following factor: 8.1 for U238, 13.5 for U-235, 2.2 for Th-230, and 20.9 for Th-232. However, this was applied only to 

demonstrate the maxima, since the daughters were not likely to have built up significantly from 

the parent for the uranium forms and were separated before the production of the ThF4 in the 

thorium case

The results are shown in Table A-34. Conversion was made from ambient dose equivalent (H*(10)) to 

organ dose equivalent using the factors given in Appendix B of NIOSH (2000a). Neutron dose 

estimates for facilities that had similar source terms and processes, Fernald (ORAUT 2004) and 

Weldon Spring (ORAUT 2005b) assign uranium-produced neutron doses using a neutron to photon 

ratio method and the method based on actual measurements of dose rates at Fernald and so would 

account for both the alpha-neutron dose and the spontaneous fission neutrons, and so the method in 

these documents is presented in Section 7.5 to account for neutron doses from uranium.
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In the analysis of the RaBe neutron production, the source was taken to contain 100 mg of Ra-226 

(Mason 1977). There is no indication as to how long the source was used, so the period of use may 

be taken to have started in about April 1942 and, as indicated in Section 5.2.4, to have ended in 

September 1944. According to Shleien (1992), Tables 7.5 and 8.4.1, a RaBe source emits neutrons 

of average energy 4.0 MeV at a rate of up to 1.3 x 107 neutrons per sec per curie and the specific 

activity of Ra-226 is 0.0366 TBq per gram, or 0.989 curies per gram. This gives a source strength of 

1.29 x 106 neutrons per sec for the 100-mg source, resulting in whole-body dose rates of 16 mrem/hr 

at one foot and 1.77 mrem/hr at three feet. 

Assuming that a laboratory worker spent one hour per day at one foot from the source and one hour 

per day at three feet from the unshielded source, the worker would receive a dose of 10.7 rem over 

the 29 months, or 4.44 rem per year. Note that these estimates are on the high side since by the 

various descriptions of laboratory work, most of the time in the laboratory would have been spent in 

preparing the samples (e.g., by grinding or chemical additions) and the source would have been 

completely shielded when not in use and likely partially or completely shielded when in use. These 

results are shown in Table A-34, along with the organ doses. Conversion was made from ambient 

dose equivalent (H*(10)) to organ dose equivalent using the factors given in Appendix B of NIOSH 

(2000a).

5.4.3

5.4.3.1

Information and Available Data Regarding Film Badges and Extremity Dosimeters

The same film badge was in use throughout Mallinckrodt uranium operations for AEC (MCW 1961a), 

i.e., from some time in 1945 through the end of operations in 1958. This was a two-element type of 

dosimeter. It is not clear whether the same type of film badge was used during the decontamination 

period (see Section 8.0), but it would appear so.

Information Regarding the Type and Composition of Film Badges Used at 

Mallinckrodt

Mallinckrodt (MCW 1956i) described the badge as follows. The badge frame was made by the A. M.

Samples Machine Company of Knoxville. The front was shaped so that a 1-mm thick cadmium shield 

could be inserted to cover the upper two-thirds of the film, with a similar 1-mm cadmium shield 

covering the back of the film. A large rectangular window covered the badge front to allow the 

identification number (the “health number”) perforation in the cadmium shield to be unblocked. A 

small rectangular window in the lower part of the badge front below the cadmium shield was covered 

only by the identification picture sandwiched between plastic sheets; a similar window in the badge 

back, below the cadmium shield, was covered only by a sheet of clear celluloid. The film was a 

DuPont Type 552 film packet containing two dental-size films wrapped together. The more sensitive 

of the two films was Type 502 and the less sensitive was Type 510. Mallinckrodt (MCW 1956i) 

remarked that the beta and gamma radiation generally encountered at the Mallinckrodt plants had 

less effect on photographic film than the same dose of X-rays that were often monitored with such film 

badges.

Mallinckrodt (1955i) also described the badge as having a metal frame made by the A. M. Samples 

Company and as using DuPont Type 550 x-ray film provided by the Dick X-Ray Company. The 

identification picture was said to be laminated between two pieces of thermo-coated acetate, each 

0.005 inches thick, with the total thickness not exceeding 0.02 inch. MCW (1950m) stated that the 

Mallinckrodt badge had 3.3 mm between the face of the badge and the first surface of the film; the 

average absorber density between the badge face and the film was given as 81 mg/cm2

. 

In an unpublished report the badge was described as follows:
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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. 

According to MCW (1956f), before the badge was issued, the week number was exposed on the 

edges of the film with 70 kV, 10 ma X-rays for 0.33 second. Control badges were exposed to a 10-mg 

radium source in a platinum needle, which was the primary gamma standard and whose intensity was 

taken to be 8.4 mR/hr at one meter. An earlier standard, No. R-515, sent to Mallinckrodt in 1948 by 

an AEC consultant on behalf of AEC, had 9.98 mg of radium, with a platinum-iridium wall thickness of 

0.5 mm, producing 5.4 mR/hr at one meter from the center of the capsule (AEC 1948I).

The primary beta standards were aged metallic uranium slabs about 1/16” thick. Eight-month-old 

uranium was regarded as “aged” enough to have the Th-234 and the Pa-234 in secular equilibrium 

with the U-238; the beta contribution of the Th-230 and its daughters was considered to be negligible. 

The beta intensity at the surface was taken to be 239 mrep/hr (see also Section 5.4.3.2 below). Other 

details of the calibration, such as exposure times and distances, do not appear to be available.

5.4.3.2

The University of Rochester processed the Mallinckrodt film badges from at least 1 July 1947 to 1 

June 1948 (Rochester 1950, MCW 1950k); Rochester presumably also processed the badges before 

that under contract to AEC, as suggested by MCW (1950k). Mallinckrodt processed its own film 

badges from 1 June 1948 to 1 January 1950 (Rochester 1950, MCW 1948g, MCW 1950k). As is 

indicated by the Mallinckrodt health group monthly reports (e.g., MCW 1951b) and the 1955 

Mallinckrodt health program description (MCW 1955d), Mallinckrodt appears to have continued to 

process their own badges even after this period. As the health group monthly reports (AEC 1950m; 

MCW 1951f) indicate, however, Mallinckrodt often had a significant backlog in reading the film 

badges; Mallinckrodt (MCW 1951e) observed that doses were frequently reported a week or more 

after the dose was incurred and thus expedited processing was requested for badges from such 

workers as the burnout and Feinc operators. But it was not unknown for this problem to occur when 

Rochester was processing the badges: results typically took two weeks to be sent back to 

Mallinckrodt and on about 19 August 1946 results were almost two months later in coming back to St. 

Louis from Rochester because of Rochester personnel’s being on duty at the Pacific Crossroads tests 

(MED 1946c). This was one of the reasons that Mallinckrodt took over the processing of its own film 

badges.

Film Badge Technical and Processing Information

No procedures and little other film badge specification data have been found to date. See Section 

5.4.3.1 for information about the film badges themselves. There is also not much information about 

how the film badges were processed by either Rochester or Mallinckrodt. However, MCW (1950r) 

stated that an AEC report (NYOO-57) was an attempt to measure the surface dose rate from uranium 

(presumably metal) received through the stratum corneum of the fingers, at 7 mg/cm2

. According to 

MCW (1950r), this report gave a dose rate of 239 mrep/hr through a depth of 7 mg/cm2 and 178 

mrep/hr through 44 mg/cm2

. Mallinckrodt had been using a value of 265 mrep/hr in their film badge 

measurements, but now agreed to use the NYOO report value of 239 mrep /hr. Mallinckrodt did so 

because the report results were based on a filter closely approximating the dead skin layer; thus this 

basis was more similar to Mallinckrodt’s operating conditions than their old value’s basis. However, 

as MCW (1950r) pointed out, AEC’s tolerance level was based on the “field strength” of the radiation, 
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not on the amount of radiation at any particular depth of any particular material; it was thought that 

any other reference level would be impossible because of the energy distribution of the betas.

There are some documents indicating problems or variations from the norm in badge readings. MED 

(1945l) stated that recently read films worn from 26 November through 4 December 1945 showed 

abnormal density combinations and that it was inconceivable that all of the density was due to 

radiation exposure. It was noted that a Mallinckrodt manager had admitted that the films might have 

been subjected to high temperatures. It was also noted that the problem could not have occurred in 

Rochester (where the films were being read at the time): the Mallinckrodt films of two weeks earlier 

had also shown high exposures and since that time, Rochester had been very careful in handling films 

from Mallinckrodt.

Mallinckrodt (MCW 1948d) reported that there was considerable fluctuation in readings on badges 

worn in the digest and Feinc filter areas; the explanation was given that any interruption in the process 

that caused a holdup of material allowed the material to “age” (i.e., daughter products to build up) and 

thus to exhibit more gamma activity. Breakdowns of equipment, which tended to result in higher 

exposures to operators doing the repairs, were another reason given for the variations.

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; AEC 1950f; 

MCW 1950k; AEC 1950g; AEC 1950i; MCW 1950x). This issue involved the subtraction of the film 

density value under the beta “shield” from the value under the window. Section 7.1 of this document 

states the assumption to be made about this subtraction; however, the references cited may be 

consulted in case of any suspect beta readings corresponding to the 1949-1950 period. In connection 

with this issue, Mallinckrodt (MCW 1950x) undertook an experiment to determine whether the beta 

reading from the open window of the film badge was due to primary (directly emitted) beta radiation 

from surface contamination in an area or to secondary beta (i.e., produced by scatter of the hard 

gammas). A radium source was used to expose four sets of film: one pair was exposed with a 4800-

gauss magnet in place and the other pair with a set of phantom magnet yokes in place. Additionally, 

one of each pair had a 7/16” Lucite shield in place to eliminate direct betas from the radium source. In 

this way, the betas were drawn off by the magnet (including any secondary betas) or were taken out 

by the shield, allowing for a 98% reduction by the magnet in beta radiation arriving at the appropriate 

films and thus the assumption of a nearly pure gamma component being measured by those films. 

Mallinckrodt determined that at most a reduction of 19% in the open window could be attained by 

aggressive cleaning of surface contamination in an area, i.e., that most beta measured by an area film 

badge was not direct beta. MCW (1950x) observed, however, that film badges actually worn by 

workers showed greater window densities for equivalent shield densities than area film badges. This 

was attributed to workers’ badges’ more frequently actually “seeing” a primary beta source, e.g., 

during manhole access. Thus during some types of work, workers actually received beta radiation 

that they would not receive merely by standing in the area.

Mallinckrodt (MCW 1948k) stated that while badge readings were supposed to be accurate within 

10% (apparently an AEC requirement), Mallinckrodt’s standards were within 5% as checked by “New 

York” (presumably Rochester or NYOO). Mallinckrodt regarded the method of reporting by Rochester 

prior to 1 July 1947 as “inconclusive” (MCW 1950k), presumably because of the rounding off to the 

nearest percent of the tolerance level and possibly also because of uncertainty in the interpretation of 

what the tolerance was and of what zero dose might actually translate to.

Prior to readout at Mallinckrodt, badges were disassembled in a laboratory hood (MCW 1956f). A 

Weston photographic analyzer, Model 877, with two stabilizers, was used in reading the films in 1949; 
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in 1956, a Welch Densichron densitometer was being used (MCW 1956f). Due to the assumed near 

proportionality of film density and radiation exposure of any given type, Mallinckrodt assumed that the 

following relationship was valid: Net window density from beta exposure alone = Actual net window 

density - Net window density from gamma exposure alone (MCW 1956f).

5.4.3.3

Workers were not individually monitored for external dose prior to June 1945. Film badge records are 

available for the weeks ending 10 June, 24 June, and 1 July 1945 (by memorandum report, MED 

1945o) and for 29 July, 5 August, 12 August, and 26 August 1945; then there is a gap until the week 

ending 9 December 1945. Total (accumulated) doses for the period from 9 December 1945 to 25 

March 1946 were reported by individual in a memorandum from the University of Rochester School of 

Medicine and Dentistry (Rochester 1950). Besides the 1945 records, records appear to be available, 

with some gaps, for most weeks from 1946 through 1948; 1950 through 1951; 1952; and 1954 into 

1958. Furthermore, film badge records for the postoperations decontamination period (see Section 

8.0) may not be available either. It is not clear why records are missing after 1945, since clearly film 

badges were used continuously from 1945 on. Hence it is possible that these records may be found 

at some later time.

Film Badge Monitoring Periods and Wearing Practices

ORAU (1980b) stated that all Mallinckrodt employees who were cleared to have access to production 

areas wore film badges at all times, with the purpose being to provide maximum assurance that all 

exposures were registered. Contemporary references indicate that film badges were issued as a 

combination security-exposure badge to all employees, except for “office females” who presumably 

never entered process areas (MCW 1955d, MCW 1955c, MCW 1956i). MED (1944p) describes the 

film badge as being numbered and being combined with a photo ID. AEC (AEC 1948c) states that for 

Plant 6, the badge was a “photo [ID] film badge” for cleared employees; a film badge with a red insert 

and the employee clock number for employees awaiting Q clearance; plastic film badges for 

technicians, engineers, and management, with photo film badges in a separate rack at guard shack or 

changehouse for when they entered the plant from the clean buildings; plastic photo badges for office 

employees who did not go into plant areas; film badges with a red insert bearing a large letter V and a 

number, for subcontractors going into Limited and/or Contaminated areas (such badges to be issued 

by the guards); and a plastic badge (with no film) with bearing a large letter V and a number, for 

subcontractors going into Controlled and/or uncontaminated areas and for business visitors to the

offices (also to be issued by the guards). MCW (MCW 1955c) also states that all visitors and outside 

contractors entering process areas were issued badges. It is not clear how early the practice of 

issuing film badges to visitors and subcontractors entering process areas began.

AEC (AEC 1948c) states that film badges were to be used at Plant 4 for a trial period, suggesting that 

that Plant 4 employees were not badged prior to early 1948; it was recommended that employees be 

issued film badges but that subcontractors and business visitors be issued a plastic badge (no film) 

with a green insert and number.

In 1955, toward the end of operations, it was realized that since an “expansion” group had moved to 

new quarters in the “Rock Island Building” (presumably in St. Louis), film badges had not been worn 

by this group during normal activities, thus leaving a gap in exposure data for these individuals (MCW 

1955a). A directive was therefore issued for badges to be worn by this group, but of course the 

missed period was not covered in the records. Similarly, the Mallinckrodt Uranium Division 

employees who transferred to the Weldon Spring site startup group in 1956 were not being monitored 

any longer for radiation exposure with film badges; those who might still have business at the 

Mallinckrodt St. Louis site were to have their badges kept for them in St. Louis (MCW 1956b). Thus 

these individuals were likely to have intermittent badge readings associated with the St. Louis site.
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Since there were no potential sources of acute external exposure, the aim was to keep chronic 

exposures below tolerance levels (MCW 1950e; MCW 1955d). All exposures over 50% tolerance 

were reported to supervisors (MCW 1955d).

From 22 April 1946 through the end of MED/AEC work in 1957 or 1958, film badges were processed 

on a weekly basis as part of a routine dose monitoring program. Badges for Plant 4 began to be 

changed only every two weeks in early September 1954 (MCW 1954d); all badges began to be 

changed every two weeks as of 30 January 1955, because of a shortage of health personnel to read 

the badges and the comparatively low doses then being recorded on the badges (MCW 1955b, MCW 

1956i). Mallinckrodt (MCW 1955d; MCW 1956e) stated that (all) badges were changed every two 

weeks or more often when desired for information or because of expected higher dose. Film badge 

results were summarized quarterly and annually (MCW 1955d). 

The guards were directed that if there was no replacement film badge available for a given used 

badge when the badges were exchanged by the guards at midnight on Sunday, they were to remove 

the used one anyway (MCW 1948d). The guards were to notify the safety department of the situation 

on Monday morning and the individual was to wear his “current” badge for the first half of the shift, 

until a replacement badge was provided by the safety department. It is not clear from the reference if 

the used badge was for the week that ended on Sunday or for the week before that, although if it was 

the latter, that would explain why the individual still had a current badge that he could wear.

Workers were directed to wear their badges when they went to the Airport (SLAPS) or “the Range” 

(MCW 1949q); the Range appeared to be a firing range (near or at the storage area) where the 

guards would practice shooting and where some materials were stored (e.g., AEC 1949e reports that 

some UO2 was drawn from this storage area for use in production). Plant 6 workers were also told to 

wear their Plant 6 badges when they visited Plant 4, and vice versa for Plant 4 workers, and not to 

wear a visitors’ badge (MCW 1948k).

As shown in the comparison table in Section 5.4.1, data from AEC 1951b show that a weekly average 

of 358 badges were issued to Plants 4 and 6 workers in November 1948 through January 1949 and a 

weekly average of 403 badges were issued in January to June 1950. From the series of Mallinckrodt 

health group reports (an example of one is MCW 1951c), some 2000-3600 badges were read per 

month in the early 1950s, corresponding to about 460-830 badges issued weekly; these included 

visitors’ badges and probably also some area monitoring badges and some experimental, duplicate, 

and supplementary (double-badging) badges.

5.4.3.4

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. Records found to date show weekly badge processing cycles, 

with the following exceptions: records that show total dose by specified time period, records from the 

very later period of operations, and records from the decontamination and decommissioning period. 

Assignment of individual annual doses was based on deep-dose exposure (Dupree-Ellis et al. 2000). 

Film Badge Record Types, Arrangement, and Availability

For most of the period of operations, 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. 
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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 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. Many of the 1948-1958 records have annual 

totals written on them; ORAU (1980a) stated that a sampling of records showed that the summing for 

the 1950-1958 records was sufficiently accurate that the totals could be accepted (i.e., for 

epidemiological study purposes), but not the summing for 1948-1949.

Dose summaries (in memorandum form) generally 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. All employees not working at Plants 4, 6E, or 7 were included in the 

Plant 6 report (MCW 1956a). 

Up to 1950, Mallinckrodt recycled clock numbers, i.e., reused them as employees left and new 

employees were hired (MCW 1950a). However, this created difficulties with keeping dose records 

because as Mallinckrodt recognized, dose records needed to have identification numbers that would 

“remain unique in perpetuity” for tracking purposes. The film badge numbering system was therefore 

changed as of January 1950 and clock numbers were no longer used on the badges. However, in 

tabulating the doses (e.g., in memorandum form), the clock numbers were used. The potential for 

badge confusion due to non-unique numbers prior to 1950 must be recognized.

Some codes used in the records to identify the type of job or work area (e.g., M for maintenance) are 

found in MCW 1956g and similar memorandum reports or in Table A-5 of this technical basis 

document. Some codes used in the records to denote readout problems or other items of interest 

(e.g., GL for Gone Through the Laundry) are found in ORAU (1980a). These are also been shown in 

Table A-5.

Complete records and/or dose summaries may not have 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.

5.4.3.5

One record of extremity dosimeter results (as ring badge readings) has been found. 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 1949b). The results were 

reported for 25 April through 30 May 1949 in AEC 1949d. Figures were said to be “the Oak Ridge 

results” based on a radium gamma calibration, so presumably these films were read at one of AEC’s 

Oak Ridge sites. The beta exposure was taken to be a factor of 1.2 times the difference between the 

open window and the shielded reading. Such data from other sites had been found to be variable and 

inconclusive, probably, AEC thought, because of improper wearing of the rings and the regular 

badges worn at the same time; AEC had determined that the ring response depended very much on 

the direction and angle of the beta sources in the process areas. Individual records of the film ring 

Extremity Dosimeters
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readings do not appear to be available among the regular film badge records, but AEC (1949d) has a 

long table giving the results by worker name and identification number.

5.4.3.6

Film badges were placed in selected locations in the process areas to serve as integrating area 

monitors i.e., to register integrated dose in an area regardless of actual occupancy (MCW 1955d; 

MCW 1955h). Mallinckrodt (MCW 1955h) opined that such film badges were sometimes hung in outof-the-way places and not in the area of highest dose rate during occupancy; however, this did not 

appear to be demonstrated by the comparison between the dose rates measured by the film badges 

and those measured by a gamma survey meter (MCW 1955h). There is no information as to how 

often these area film badges were collected and read.

Film Badges Used as Area Dosimeters

5.4.4

MED (1944d) stated that the medical program recommended by MED was being followed by 

Mallinckrodt. This included a routine chest x-ray prior to employment and annually thereafter and a 

pelvic x-ray for those employees handling fluorides (MED 1944j). MED (1944d) thought that the 

pelvic x-ray could be dispensed with and apparently this was subsequently done. Later documents 

also indicate that Mallinckrodt uranium processing workers were given a pre-employment physical 

that included a chest x-ray (MCW 1955d); Mason 1958a); they were also given an annual physical 

that included a chest x-ray (MCW 1955d; Mason 1958a). That these x-rays were actually given is 

indicated in the series of Mallinckrodt Health Office and other reports (e.g., MCW 1951b; MCW 1954e; 

MCW 1955d). No information is available as to how much dose was received during these 

examinations or if all workers received annual x-rays. Therefore if no other information is available for 

a claim, it is assumed that all personnel who worked in the plants covered by this technical basis 

received an annual occupationally required medical chest x-ray. Personnel working with hydrogen 

fluoride or in the UO2-to-UF4 conversion process in 1942-1944 probably received an annual pelvic xray.

Information and Available Data Regarding Occupational X-Ray Examinations

5.4.5

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

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; a 19 August 1946 potential significant exposure to 

workers, probably in the removal of K-65 in the storage area, following which working space was 

rearranged and the pileup of residue material at the location was reduced (MED 1946c); 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).
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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 1952b). 

This may have been a longstanding problem, however, since the reference in early 1944 to the loss of 

material in “burnout” in the reduction step apparently was to this spontaneous ignition of derbies (MED 

1944o). Another such recurring incident was the plugging of floor drains, with the consequent 

formation of puddles of contaminated liquids on the floor at Plant 6 (AEC 1950m).

One incident of concern to the health department was the hand loading of raw pitchblende ore on 21 

February 1950, without the knowledge of the health department (MCW 1950d). Apparently there was 

a malfunction of part of the ore drum conveyor. Operational pressures motivated the production 

supervisor to have the star valve at the bottom of the ground ore hopper feeding into the secondary 

elevator disconnected. A three-man crew then picked up open drums off the conveyor using a chain 

hoist and tipped their contents onto the floor. One worker then shoveled the ore across the floor to 

another worker and he to the third worker, who shoveled it into the chute to the secondary elevator. 

One crew worked one half-hour session and a second crew worked another half-hour session, 

together handling 12 drums of ore. The crews wore respirators and the downdraft ventilation into the 

chute was working, but the workers were not otherwise protected from direct radiation or radon, and a 

crew of maintenance people working around the rod mill nearby were not wearing respirators. It was 

noted that the gamma exposure at the surface of such an ore drum was 80-100 mR/hr (MCW 1950d); 

in addition, Table A-24 shows some radon levels for open pitchblende ore drums. However, it was 

also noted by MCW (1950d) that health department clearance would have been granted for such an 

operation if the workers had worn oxygen or air line masks, if stay-times were limited to 1 hour per 

week per worker, and if workers were taken from low-dose areas such as the Ether House or the 

furnace room.

5.5 OTHER DATA OF DOSIMETRIC INTEREST

5.5.1

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 1944p). Regarding the total number of workers with 

potential for exposure, Fleishman-Hilliard (1967) and Mallinckrodt (1994) listed the total number of 

workers as 250, the former stating that this was in 1948; AEC (AEC 1948e) listed the total number as 

250 at Plant 6, but 400 if Plant 4 was included; AEC (1949b) listed the number of workers at Plant 6 

as 272 and the number at Plant 4 as 94; Mallinckrodt (MCW 1950w) gave the total number as 487 

and added that 275-300 individuals who worked on the project for more than three months had 

terminated as of 1 October 1949. Mason (1958a) stated 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 

(1951b) stated that in early 1951, Plant 4 employed 100 people.

Number of Workers

AEC dust study reports in the 1950’s gave the number of each classification of workers and the 

number on each shift (e.g., AEC 1954b); some of these AEC reports even listed the names of process 

and supervisory workers and their job classifications in an appendix. For Plant 6 alone, Mallinckrodt 

(MCW 1956a) gave the number of manufacturing personnel as 100 and the number of nonmanufacturing personnel as 550. As noted above in Section 5.4.3.3, over 2000 film badges a month 

were processed in the 1950’s. 
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5.5.2

From AEC dust study reports (e.g., AEC 1954b), the following information regarding time usage is 

provided:

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 workday 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.

AEC-NYOO took the number of work-hours per week to be 48 (or six 8-hour days) in calculating some 

of their early time-weighted average airborne concentrations (AEC 1949b). Also, AEC (1951b) stated 

Plants 6E and 7 operated on a six-day, three-shift schedule. However, this does not imply individual 

Mallinckrodt workers worked six days a week. MCW (1949g) states that the workweek is 8 hours per 

day, 5 days per week. Lippmann (1958) used 40 hours in reporting data regarding Harshaw workers

and Harshaw (1946) described its rotation schedule used for round-the-clock operation, which 

involved a 40-hour week for most individual workers and a 42-hour week for some workers. Besides 

these references, AEC, in its dust study reports, used the actual time worked, as given above, 

implying a five-day week. It can thus be assumed that Mallinckrodt workers typically worked for a full 

8 hours a day, 5 days a week, or 40 hours per week. When using daily weighted average dust 

sampling data (e.g., in Tables A-12 through A-16 and A-19 through A-22), 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), they were 

carried out 24 hours per day. It is not clear what other early processes ran 24 hours per day, but AEC 

(1951b) implies that Plants 4 and 6 did not. However, individual dust studies cite three shifts for many 

process workers positions, indicating that operations were in fact carried out 24 hours per day, 

although perhaps not on the weekends. Guardhouses were manned around the clock, with three 

shifts per day (MED 1944p).

5.5.3

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 listed such notes or keywords. Various AEC dust reports also list job titles and functional work 

types. Job titles and types discovered to date are given in Table A-18. 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 may have been part of 

preliminary decontamination prior to the decontamination and decommissioning of buildings and 

plants; however, it is known that there was a decontamination group that did ordinary or process 

decontamination and special decontaminations as necessary (e.g., of a lab area), so when 

“decontamination” appears in a record it is this type of work that should be assumed.

Job Types, Job Histories, Work Areas, and Work, Access, and Clothing Practices
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As previously stated, to aid in classifying workers whose job titles do not appear in Table A-18 and 

whose work descriptions do not make it clear which job title is appropriate for use, Table A-5, the 

keyword table, which includes information from these notes and from operational information in other 

references (particularly MED 1946a; AEC 1949b; and AEC 1967) is provided. Tables A-5 and A-18

should be used to help determine the principal occupational activity for an individual with missing or 

conflicting monitoring data.

According to ORAU (1983b), prior to 1 October 1949, there was no “concise” or in-one-place 

employment history for Mallinckrodt workers in the Uranium Division (i.e., doing MED/AEC work). 

Then because of the high doses, it was apparently thought that workers who had received more than 

90% of their (unspecified) “lifetime” exposure to dust should be removed from uranium work. 

Consequently, a work history was reconstructed by Mallinckrodt for the period of 1942 to 1 October 

1949. Mallinckrodt tried to be as accurate as possible because of the possibility of having to remove 

workers and apparently the workers examined the reconstructed records, because “the work history 

had to withstand the scrutiny of each worker” (ORAU 1983b). Four volumes of records were 

produced: one master list of workers, one volume with histories of those terminating prior to 1 October 

1949, one for current Plant 4 workers, and one for current Plant 6 workers.

In the Mallinckrodt Uranium Division, workers in maintenance (shop personnel, skilled craftsmen, and 

maintenance mechanics) and in plant services (guards, porters, nurses, laundry workers, power plant 

personnel, hygiene and safety personnel, etc.) were not usually assigned to a particular process, but 

served the entire division (Hickey and Dupree 1984). An exception may be the “area mechanic” 

mentioned in some AEC dust sampling reports, but even then it may be that the mechanic was 

assigned to cover multiple areas and his exposure may not have been characteristic of only one area. 

However, Hickey and Dupree (1984) note that the uranium monitoring records they examined showed 

a reasonably uniform exposure for both the field and the shop maintenance personnel and skilled 

craftsmen.

ORAU (1977) stated that the highly exposed workers were transferred out of uranium operations in 

1950; these were presumably some or all of the 52 workers found to have 100% or more of the 

lifetime tolerance dose as of October 1949 (AEC 1950a). Mallinckrodt (MCW 1951b) and AEC 

(1950e) also stated that some 34 workers were transferred in the summer of 1950: the former states 

that this was from Plant 6 to Plant 1 “for health reasons” and the latter states that this was from Plants 

4 and 6 on the basis of their estimated accumulated exposure to uranium dust. While Building 51 

(Plant 2) was said by ORAU to have been closed about 1 January 1947 and by Mason (1977) to have 

been “sealed” in about 1947, when Plant 6 started up, and information in ORNL (1981) suggests that 

Plants 1 and 2 might have been in use for some support work until decontamination began in 1948; 

also, MCW (1949e) indicated that special badges would be issued to workers at Plant 1, although this 

may have been for intermittent work in Plants 4 and 6. Thus the Plant 1 work, if any, would 

presumably have been non-uranium work. Whatever the case, Mallinckrodt stated that the transfer 

had been considered for over a year before it occurred and that it would send a report to AEC 

(1950d); this report does not appear to be available. Possibly partly as a result of the study reported 

by AEC (1950a), Mallinckrodt used an employee rotation program from about 1950 on (FleishmanHilliard 1967; AEC 1950b; MCW 1955d), 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.

A MED (1944p) security survey report gave 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. Non-Mallinckrodt truck 
Document No. ORAUT-TKBS-0005 Revision No. 02 Effective Date: 06/14/2007 Page 100 of 225

drivers were allowed to come in, but not truck drivers’ helpers unless needed to unload; all trucks 

were escorted while within the site. MED (1944l) stated that personnel in the “MED plant” were 

segregated as much as possible from the general (non-MED work) personnel, with the main point of 

contact being during the lunch hour.

From April 1949 on, the film/ID badges were color-coded to control access (MCW 1949h); for 

example, Plant 6 was divided into seven zones, each with its own color. “Controlled” areas were 

areas not expected to become contaminated, such as the offices, the power plant, the pumphouse, 

and the yard areas. “Limited” areas included the main processing building, the warehouse, the Scale 

House, the laboratories, and so forth. Mallinckrodt’s 1955 formal description of its health program 

(MCW 1955d) gave the following information, indicating that the area control system was continued 

throughout the processing years although the definitions might have changed somewhat. Mallinckrodt 

maintained three levels of controlled areas. These were the regulated areas, which were described 

as 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 the clear areas, 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 stated goal of 

control efforts. 

An internal Mallinckrodt memorandum circa April 1948 (AEC 1948c) stated that the Building 101 

facilities had been “revised” (implicitly for better control of contamination) and access status in the 

building and yard areas was to be broken into two classifications. The first was the “clear area,”

consisting of all areas in which the presence of “plant products” could be completely eliminated or

reduced to negligible amounts. The designated clear areas were the guard office, the locker room, 

the lunchroom, the Health Unit offices and laboratory, the AEC offices, the MCW offices, and the 

MCW laboratory offices. The second area was the “regulated area,” consisting of all areas that could 

not be kept clear of plant products. 

MED (1942) stated that employees were required to change work clothes daily, with the “uniforms” 

being provided by Mallinckrodt. AEC (1948c) states that in the clear areas, either street clothing or 

plant clothing with cover clothing was allowed, except in the lunchroom, where only street clothing 

was allowed; in the regulated areas, either plant clothing or street clothing with cover clothing was 

allowed. In 1955, the Mallinckrodt health program was said to include the following requirements 

regarding clothing (MCW 1955d). 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; and cover clothing was also 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) (MCW 1955d). Regulated area workers were issued two changes of clothing per day as of 

1955 (MCW 1955d) and probably much earlier due to contamination control requirements involving 

showering (see below).

Visitors were provided with cover clothing (smock, rubbers, etc.) (MCW 1950e). However, as of at 

least 1950, outside contractor personnel wore their own personal work clothing, on the grounds that 

visitors’ garb would be “too restrictive” and thus unsafe (MCW 1950e). Besides that, suitable change 

room facilities were not provided for outside contract personnel (MCW 1950e).

Process (manufacturing) workers were required to shower before changing into “clear” (clean) 

clothing; they typically took two showers a day in 1955 (MCW 1955d), but only one in the 1942-1944 

time frame (MED 1942, MED undated a) and apparently only one in April 1950 (MCW 1950e). The 
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procedure was as follows (MCW 1948c). After punching in, workers who would be changing into plant 

clothing would remove their street clothing in the locker room, don plant clothing in the change room 

(apparently the same clothing from the second half of the previous day’s shift), and enter the work 

area. At lunchtime, workers would remove the plant clothing in the change room, shower, don street 

clothing, and go to the lunchroom. After lunch, they would put on a clean set of plant clothing in the 

changes room and go back to work. At the end of the shift, workers would remove the plant clothing 

in the change room, shower, and don their street clothing in the locker room. An extra 15 minutes 

was allowed for the midday showering and changing and clogs were provided for traffic between the 

locker room, change area, and showers (MCW 1948c). The one-way door allowing traffic to pass only 

from the locker room to the change area was to be supervised at shift change and lunch hours (MCW 

1948c). No one was allowed to leave the plant in plant clothing (without cover clothing, presumably) 

(MCW 1948c). By February 1949, an AEC consultant, inspecting the change room for Plant 6 

process workers, termed the operation of the change room “creditable,” i.e., in line with AEC 

directions (AEC 1949k).

Plant 6 warehouse personnel were to follow the same general procedure in their change room, as 

were maintenance workers who used the process area locker room. The warehouse personnel 

leaving Plant 6 on trips to other regulated areas (such as Plant 4, SLAPS, or “the Range”), were 

allowed to make the trip in plant clothing, but in a “regulated” vehicle only; these vehicles could be 

jeeps, dumpsters, or large trucks) (MCW 1948c). Also, warehouse personnel leaving on trips to clear 

areas outside Plant 6 had to shower and change as if they were leaving work and they had to use a

“clear” vehicle only (MCW 1948c).

Laboratory personnel who normally wore plant clothing were to follow the general procedure for 

process workers, except they used their own locker room and change room; personnel going to the 

laboratory offices (a clear area) from another clear area could go directly in street clothing but had to 

use cover clothing to go to the laboratory; and personnel going from working in a laboratory to the lab 

offices had to put on cover clothing over their plant clothing (MCW 1948c).

The change requirements given above for process workers also applied to maintenance personnel 

who used the maintenance shop locker room: they were to use the same procedure but in their own 

locker room (MCW 1948c). Since the maintenance men in the shop might also work in the now nonAEC Plant 1 and presumably other clean areas at the general Mallinckrodt facilities, they had to 

change into special blue coveralls and either street shoes or clean safety shoes when visiting those 

areas; similarly, any Plant 1 maintenance men going to the AEC areas to work were to follow the 

change procedure as for the process workers (MCW 1948c). Instrument Shop personnel going to 

Plant 4; however, could go in plant clothing but in a regulated vehicle only (MCW 1948c).

Mallinckrodt (MCW 1948c) stated that no one was permitted to eat in plant clothing; that no one who 

had been working in plant clothing was allowed to eat without previously showering; and that there 

was to be no eating in regulated areas. Soft drink dispensers and washing facilities were provided in 

the change room, but hands were supposed to be washed before drinking (MCW 1948c). According 

to MED (1944p), 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.

5.5.4

Some quantities and dimensions of potential radiological interest (e.g., for special external dose 

calculations) follow. See also Table A-4 and Section 4 for other amounts.

Miscellaneous Product Information
Document No. ORAUT-TKBS-0005 Revision No. 02 Effective Date: 06/14/2007 Page 102 of 225

Bomb 10” OD × 40” long MED 1946a

Bomb liner (lime, etc.) Depth: 1” MED 1946a

Uranium billet 4.75” OD × 18” long MED 1949b

Uranium billet ~ 6” OD Mason 1977

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

5.5.5

The table below gives some Missouri and St. Louis area background measurements, except that the 

Applied Nuclear Safety measurements were taken at Building Z of the Mallinckrodt site, used as a 

control. Note that these measurements were all taken many years after the end of operations and 

were made for the purpose of characterization for possible future remedial action.

Missouri and St. Louis Area Background Levels

Measurement Unit

ORNL 1979 ORNL 1981

Applied 

Nuclear Safety 

1986

Bechtel 

1987

Range Average Range Average Average Average

Ac-227 pCi/g Below 

detectable

Ra-226 pCi/g 0.3 - 1.3 1.05 ± 

0.3

1.18

Th-232 pCi/g 0.3 - 1.3 1.15

U-238 pCi/g 0.3 - 1.7 1.25

Gamma background 

at points distant 

from site

uR/hr 6.0 ± 1.7 7 - 9 8 11

Beta-gamma (GM) mrad/hr 0.02 .01 - .04 0.02

Radon concentration pCi/L 1.0 0.33 0.3

Radon daughter 

concentration

WL 0.01 0.002

6.0

Doses related to Mallinckrodt operations (and initial decontamination) are only to be assigned during 

the period January 1, 1949, through December 31, 1958. The following summarizes work categories 

for the purpose of assigning internal doses to individuals not in the special exposure cohort and who 

might have had no bioassay. This section also applies to SLAPS employees (who are not included in 

the Mallinckrodt SEC, and so assignment of additional radionuclide intakes [see section 8.6 and 

subsections] might be needed to completely reconstruct SLAPS employees’ internal dose.) All 

workers are assigned uranium intakes. Only workers who fit the descriptions in the following table or 

whose job description and/or locations are unknown are assigned radium and thorium intakes. In 

Section 6.4, information for assigning radon exposure is provided.

RECONSTRUCTING INTERNAL DOSES DURING THE OPERATING YEARS
Document No. ORAUT-TKBS-0005 Revision No. 02 Effective Date: 06/14/2007 Page 103 of 225

Summary of internal dose reconstruction assignments by work category

Category Description How to identify Period of applicability

Dose reconstruction (if individual 

bioassay are not available)

General Probably had no 

contact with process 

area.

1/1/1949-12/31/1958 Coworker data: assume lognormally 

distributed uranium intakes with a 

minimum GSD of 3, if uranium bioassay 

data for an individual are not available.

General Had contact with 

process area.

1/1/1949-12/31/1958 Coworker data: assume constant 

distribution and use 95th percentile 

uranium intakes, if uranium bioassay 

data for an individual are not available.

6.1 URANIUM INTAKES

The uranium photofluorometry urinalysis detection threshold is assumed to be 0.01 mg/L, which is 

consistent with the value used for other sites during this period. Uranium is assumed to be either 

absorption type M or S.

6.1.1

For workers with no associated bioassay, the coworker uranium intakes based on the CER data 

provide a basis for uranium dose determination. The CER data is a set of worker and workplace 

monitoring data compiled by Oak Ridge Associated University for epidemiological studies (ORAU 

2005). The compiled worker information only includes data for white males. Reviews of the accuracy 

of the compiled urinalysis results performed by individuals from Sanford Cohen & Associates, Inc. and 

NIOSH revealed small differences between the original data and the CER data. A small number of 

differences in dates, results and type of data entered were found by the contributors to this document, 

but it is unlikely these few differences will grossly change summary analysis results of the CER 

uranium urinalysis data. (These differences were readily apparent in the claim file data, and so the 

appropriate data are readily available for use in individual dose reconstruction.) The geometric mean 

and 84th percentiles were determined by year from the CER uranium urinalysis results in mg/L for 

each of the years 1949 through 1957 using methods in ORAUT-OTIB-0019, were adjusted to units of 

pCi/d by multiply by 682.96 pCi/mg and by 1.4 liters of urine per day, and are shown below.

Assignment of Uranium Intakes for Unmonitored Workers or Unmonitored Periods

Uranium bioassay results in pCi/d

Year

Number of 

results

Geometric 

mean

84th 

percentile

1949 638 0.016 0.065

1950 1091 0.009 0.030

1951 1279 0.009 0.023

1952 1910 0.009 0.024

1953 2421 0.008 0.022

1954 1943 0.006 0.018

1955 1897 0.006 0.017

1956 802 0.005 0.013

1957 1038 0.004 0.014

The intakes shown below were calculated with IMBA ExpertTM ORAU-Edition, Version 4.0.9, 

assuming an absolute uniform error of 1 and normal error distributions for each annual bioassay 

result, which was assigned a date of July 1 for its year. The geometric standard deviations (GSDs) for 
Document No. ORAUT-TKBS-0005 Revision No. 02 Effective Date: 06/14/2007 Page 104 of 225

the intakes were calculated by dividing the intake from the 84th percentile regime by the intake from 

geometric mean intake regime.

If uranium urinalysis results are not available for a worker or for an extended period of employment, 

uranium intake rates estimated for coworkers based on compiled bioassay results may be used. The 

lognormal intake distribution information derived from the coworker data is used to estimate organ 

dose distributions for workers, who were unlikely to have had assignments associated with uranium 

and associated radioactive material processing operations, as determined from claim specific 

information. For workers, who were likely to be associated with process work or process work areas, 

the 95th percentile of the CER derived intake distribution using a minimum GSD of 3 is used to 

determine organ doses and is assumed to be a constant distribution. These uranium intakes are 

shown in the table below. 

The source term is natural uranium, but U-234 can be used for intake and dose calculations once 

results are converted to activity. Either type M or type S uranium is chosen, but not both. The dose 

distribution type is either lognormal for the 50th percentile intakes or constant for 95th percentile 

intakes.

Uranium lognormal distribution, and 95th percentile intakes based CER urinalysis data

Type Start End

50thpercentile intake 

rate (pCi/d) GSD 95th percentilea intake rate (pCi/d)

Workers outside process, 

storage and waste areas

Process and waste workers and 

workers in process and radioactive 

material storage buildings

M 1/1/1949 1/1/1950 270 4.15 2,810

1/1/1950 12/31/1958 95.8 2.79b 517

S 1/1/1949 1/1/1950 8440 3.87 78,000

1/1/1950 1/31/1958 762 1.87b 2,130

a. To calculate the 95th percentile intake, the 50th percentile intake was multiplied by the GSD (using a minimum of 3) 

raised to the 1.645 power

b. A minimum GSD of 3 should be used for estimating dose. 

6.2 RADIUM INTAKES

The radon breath analyses, when available for an individual, should be used to estimate Ra-226 

body activity using the method in ORAUT-OTIB-0025, i.e., by multiplying the radon breath result in 

pCi/L by 2.52E5 to get the activity in the body at the time of the bioassay. The lower limit (of 

detection) at HASL was 0.1 pCi/L (ORAU 1983b). The radon breath analysis recording level (the level 

at or above which a finite result was recorded) was 0.1 pCi/L for the vast majority of Mallinckrodt

radon breath analyses (AEC 1949-1956), and this is to be employed as the detection threshold. 

Although there are some factors that could cause radon breath results to be underestimated, the 

available information indicates that the collection of these samples would likely result in reasonable 

estimates of radon in breath for the purpose of this program, in other words, no information has been 

found to indicate that these results would have been underestimated. The records indicate that a 

number of Mallinckrodt samples were collected in areas of elevated radon (MED 1945e), which 

resulted in the breath radon emanations from radium within the body to being overestimated. In 

addition, breath radon samples should not have been collected within one to two hours after a meal. 

(Eating appears to increase the radon breath concentration and waiting would reduce the body 

emanation concentration by up to a factor of 2 and thus allow for appropriate application of the factor 

to convert from radon breath concentration to radium activity concentration within the body.)
Document No. ORAUT-TKBS-0005 Revision No. 02 Effective Date: 06/14/2007 Page 105 of 225

6.3 THORIUM INTAKES

About 70 thorium urinalysis results for the period March 4 through April 19, 1955 have been located 

for some workers, and should be used for individual dose reconstructions. Most results are reported 

as a value followed by a +/- and another value. The results are reported in dpm/L and the average of 

the +/- value is 1.1 dpm/L (AEC 1954-1955), and twice this value is assumed to be the detection 

threshold, because the result are near background and the lower limit of detection is about 2 times the 

one-sigma background counting error. It is suspected that most workers who had thorium bioassay 

results were ionium (Th-230) workers, engaged in recovery of thorium from AM-7, which occurred 

during the period January 1, 1955 through January 1, 1958, but this bioassay (chemical separation of 

thorium and gross counting for quantification) would also have detected intakes from earlier process 

work. Later some workers were involved in the dismantling and decommissioning of this area.

6.4 RADON EXPOSURE

Area radon data are used to assign radon exposure. CER radon data for Plant 6 and subsets of the 

Plant 6 data by areas were considered. A set of lower exposure data, for areas including Dispensary, 

Electric Shop, Furnace Room, Health Office, Instrument Shop, Laboratory, Locker Room, Machine 

Shop, Maintenance Shop, Manufacturing Area, Metal Dissolver Building, Millwrights Shop, Nitric Acid 

House, North Laboratory, Open Area, Open Storage Area, Orange Packing Area, Outside Area, 

Recast Area, Receiving, Refrigeration Room, Research Lab, Rest Room, Shipping & Receiving, 

Shotgun Laboratory, Slag Building, Slag Pilot Plant, Smoking Room, South Laboratory, Welding 

Shop, and X-Ray Lab, where the radium source term was not predominant, are also considered to 

assign radon exposure to workers who did not enter areas associated with radium and associated 

processing or storage areas. Data for SLAPS were also analyzed, as were the data for Ore Storage 

and the Scalehouse. 

The CER data were analyzed by assuming the data are lognormally distributed. Data were ranked 

from low to high. The z-scores of the ranks and the transformed data were fit to a line for 

determination of the 50th percentile and geometric standard deviation of each distribution in 

accordance with ORAUT-OTIB-0019 methods. Results that were “less than” results were ranked, but 

not fit. No attempt was made to sort data prior to analysis to remove those data that were not 

representative of occupied areas, for instance, measurements taken in a drum head space or in the 

thawing furnace would likely have been included. For some areas, the large geometric standard 

deviations provide some indication that two different concentration distributions are being measured. 

A summary of the results is shown below.

1949-1957 range of radon exposure distribution parameters for different percentile categories based 

on data in CER

Location 50th (pCi/L) GSD 84th (pCi/L) 95th (pCi/L)

95th

Occupancy

0.5

95th

Occupancy

0.05

Low radon areas 1.6 4.7 7.36 20.04 - -

Plant 6 3.4-19 3.3-6.6 20.4-72.2 58.9-244 - -

C-3 4.9-35 2.7-9.9 32.7-231.6 78.9-1012 - -

FEinc 3.6-28 2.2-5.3 13.9-66.5 33.3-181 - -

K-65 centrifuge 3.1-13.3 2.2-8.3 14.6-59.5 23.9-192 - -

Ore storage 1.2-26 3.7-22 13-150 41-590 22-300 2.2-30

Scalehouse 1.1-59 3.2-8.3 4.2-260 9.6-680 5-340 0.5-34

SLAPS (1948-1949) 53-85 4.3-7.3 231-622 595-2249 300-1125 30-113

Summary All < 100 - Most less 

than 100

Most less than 

200

- -
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To determine exposure due to a specified radon concentration (assumed to be Rn-222), the 

concentration C in pCi/L was converted to potential alpha energy exposure (PAEE) per year units of 

working level months (WLM)/y using:

PAEE = C * F* M / (100 pCi/L per WL)

where the equilibrium factor F was taken as 0.4, a value recommended by the ICRP (1981) and the 

United Nations Scientific Committee on the Effects of Atomic Radiation (UNSCEAR 1993) and M is 

the number of months of exposure, 12, in a year, where a month is assumed to be 170 work-hours.

To assign radon exposures, workers are categorized into three exposure categories as shown in the 

next table. By 1949 workers were being restricted in the number of hours they could work at SLAPS 

or the Scalehouse. Workers were rotated through some of the higher radon exposure jobs. 

Therefore it would be reasonable to apply an occupancy factor for the Ore, Scalehouse and SLAPS. 

A bounding value would be 0.5, but it is more likely on the order of 0.05. The selection of the radon 

exposure values for the table below included consideration of the occupancy associated with Ore 

Storage, the Scalehouse and SLAPS.

Radon exposure assignments

Category Description Period and exposure Basis

General plant Did not work in or routinely enter 

process or associated radioactive 

material storage/waste areas.

1949-1958

0.962 WLM/y

Constant distribution.

95th percentile of the low 

radon areas plus outdoor 

storage area

Process areas Worked in on near process areas, 

but did not encounter ore 

products, residues or associated 

storage areas directly.

Year WLM GSD Lognormal distribution of 

the Plant 6 radon 

concentrations includes 

the possibility of lower 

and higher exposures

1949 0.52 6.62

1950 0.91 3.57

1951 0.82 3.39

1952 0.54 4.30

1953 0.45 4.11

1954 0.40 5.65

1955 0.47 4.68

1956 0.24 4.48

1957

-

1958 0.16 6.04

Process workers Includes operators, ore samplers, 

residue samplers, laborers, 

maintenance workers, etc.

All SLAPS workers would be 

included in this category, 

including guards 

1949-1958

14.4 WLM/y

Constant distribution.

Based on radon 

concentration of 300 

pCi/l, but records indicate 

these levels trended 

downward over time.

6.5 UNMONITORED RAFFINATE EXPOSURES (1949 - 1958)

NIOSH has determined, and the Secretary, Health and Human Services has concurred, that it is not 

feasible to reconstruct radiation doses (both internal and external) for individuals who worked at 

Mallinckrodt prior to 1949 (HHS 2005a). For the period 1949 through1958 NIOSH has determined, 

and the Secretary, Health and Human Services has concurred, that it is not feasible to reconstruct 

internal radiation doses from internal exposures to non-uranium radionuclides for Mallinckrodt 

raffinate workers, nor is it possible to identify which workers were raffinate workers (HHS 2005b, HHS 

2009). 
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Consequently, only internal radiation doses associated with specific bioassay types (uranium, radon in 

breath, and thorium) and radon area measurements will be estimated for individuals who worked for 

Mallinckrodt during the period 1949 through 1958. 

7.0

This section applies to determination of external gamma, beta, and neutron doses during the period 

1949 through 1958 only. For external doses during the remediation period see Section 8.0. Film 

badge coworker data have not been completely analyzed, but if needed for more accurate 

compensability determinations, they could be more rigorously evaluated. 

RECONSTRUCTING EXTERNAL DOSES DURING THE OPERATING YEARS

7.1 ASSUMPTIONS AND GENERAL INSTRUCTIONS

The following assumptions should be made in the determination of external dose.

1. Most external dosimetry monitoring records appear to be available, but where such records 

are missing doses must be estimated. As alternatives for dose estimation information, area 

monitoring data for external radiation is sparse and although much is known about the 

characteristics of the process material (source term), the lack of knowledge of such relevant 

factors as geometry make dose estimation using this data subject to a great deal of 

uncertainty. Thus the approach incorporating the least uncertainty is likely to be using existing 

dose monitoring records to create an applicable surrogate(s) set of data.

2. Mallinckrodt dose records demonstrate that a substantial proportion of employees had film 

results recorded as 0-50 mrep or mr. From this it can be assumed that the dose-monitoring 

program was sufficiently conservative that even many individuals who did not receive 

significant occupational exposure were also monitored. It can be further assumed that the 

converse is true, that individuals who were not monitored were unlikely to receive significant 

occupational exposure. This assumption is supported by the observation of Hickey and 

Dupree (1984) that Mallinckrodt health physicists and industrial hygienists presumably 

monitored those jobs and workers where they thought the problems were and that it can thus 

be inferred that infrequently monitored workers such as the general administrative staff were 

not considered to be at risk of significant exposures.

3. Photon doses for all workers should be assumed to be in the 30 – 250 keV energy range. This 

is consistent with NIOSH (2007). It is also consistent with the fact that all but a fraction of the 

gammas emitted by U-235, U-238, and their daughters down to radon are in this range, 

including the principal gamma of Ra-226 (0.186 MeV) and the principal gammas of Th-230 

and Th-232. 

4. 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). Beta dose should thus be assigned as 

shallow dose from electrons of energy greater than 15 keV.

5. The descriptions of the film badge in Section 5.4.3.2 above do not suggest calibration using a 

phantom, so it is likely that open-air calibrations were performed. Thus the recommendation is 

that Mallinckrodt recorded doses be converted using dose conversion factors for Roentgen-toHT dose for photons from Appendix B of NIOSH (2007). While film badges are known to 

overestimate doses from low-energy photons, the low-energy component does not seem to be 

a significant characteristic of the Mallinckrodt spectrum. Thus no modification is proposed to 
Document No. ORAUT-TKBS-0005 Revision No. 02 PC-1 Effective Date: 05/25/2009 Page 108 of 225

recorded deep doses, once converted to organ doses using the Roentgen-to-HT dose conversion 

factor.

7. Examination of entries for “gamma” and “beta” in the Mallinckrodt dose records shows that the 

beta doses were obtained by subtracting the optical density for the shielded portion of the film 

from that of the unshielded part. This is also shown by a series of memoranda between the 

AEC and Mallinckrodt (AEC 1950f, AEC 1950h, MCW 1950o) in which the method of 

subtracting the two quantities was discussed. For Mallinckrodt, it is assumed that the beta 

readings were obtained by subtracting the density under the shield from the density under the 

window and assuming the difference in density was due to beta radiation (MCW 1950o).

8. For the purposes of dose reconstruction, the minimum level of photon detection should be 

assumed to be 50 mrem, based on 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.” The minimum level of electron detection should be assumed 

to be 80 mrem.

9. OCAS-TIB-0013 (NIOSH 2006) geometry factors should be used to adjust external doses in 

individual records and in this document for operators and material handlers and for trades and 

crafts personnel. 

7.2 INFORMATION REGARDING THE 1950 AEC CUMULATIVE DOSE 

RECONSTRUCTION STUDY

The lack of early external monitoring data for Mallinckrodt 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 at the end of the war and 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 Estimate of Cumulative Multiple Exposures to Radioactive Materials, 

Mallinckrodt Chemical Works Plants 4 and 6, July 1942 to October 1949” (AEC 1950a). AEC’s results 

are not quoted here for two reasons. First, in dose reconstruction different assumptions are made 

than AEC made. Second, the study develops cumulative dose data, but unfortunately does not apply 

it to individual workers or distinct work groups, while doses applied to individual workers for dose 

reconstruction should rely on recorded doses for actual worker or worker groups. However, dose 

reconstructors should be aware of this report.

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.

7.3 GAMMA AND BETA DOSE

NIOSH (2007), the external dose reconstruction guide, states that the hierarchy of dose determination 

sources is personal dosimeter (the film badge worn by the individual); pocket ionization chamber; and 

group or co-worker dosimeter. Pocket chamber data does not appear to be available for Mallinckrodt 

and in fact, use of the pocket chamber (pocket meter) was discontinued in 1949, as explained in 

Section 5.4.1. However, such data may come to light at a later date and in that case, some 

calibration and other information can be found in Section 5.4.1 and the references it cites. NIOSH 

(2007) does not list the area monitoring dosimeter, such as Mallinckrodt used in many areas as a 
Document No. ORAUT-TKBS-0005 Revision No. 02 Effective Date: 06/14/2007 Page 109 of 225

secondary means of monitoring. However, such data, if available, could be used in a comparative 

way if none of the other data sources can be found.

Generally, for a given claim, dates of employment should be compared to the available dose 

monitoring information. If dose monitoring records are likely to be available, the dose reconstructor 

should request project personnel to search the available records. 

7.3.1

Surrogate (analogous co-worker) data, in Mallinckrodt cases, is a matter of identifying an appropriate 

set of workers with a similar work history. That is, in general, to form a surrogate set the dose 

reconstructor should identify at least several other workers with appropriate monitoring data who did 

the same work at the same time as the subject worker; the monitoring results for these surrogate 

workers should then be pooled to arrive at a statistical representation of the likely dose to the worker.

Selecting a Surrogate Cohort and Formulating a Surrogate External Data Set

For the monitored years, many, perhaps most Mallinckrodt film badge records and urinalysis records 

had some indication of job title, task, or work area. Thus the formation of a surrogate set of data 

should be possible for all monitored years by inspection of the weekly dose reports, supplemented as 

necessary by inspection of the quarterly urinalysis reports. (Note that the urinalysis reports are a 

useful and available additional source of co-worker and job assignment information for external dose 

reconstruction, even though the urinalysis results themselves are not relevant to external dosimetry.) 

Dose reconstructors must compare information available from the DOE record and the computeraided telephone interview (CATI) to the records in order to identify workers with a similar work history.

One complication of selecting a surrogate set for the monitored years is that from some point on in the 

postwar years (maybe beginning in 1949), a rotation system was used in work assignments in order to 

reduce individual process worker external exposure (MCW 1950t; AEC 1951b). The work done by an 

individual would therefore vary over the course of a year as he was rotated in and out of highexposure areas. Thus for best results, where data for an individual known to be working at 

Mallinckrodt in AEC operations are lacking or are unclear, some care must be taken to identify coworkers whose assignments are most analogous to the unmonitored workers. The rotation system 

was done principally on a crew basis, i.e., the same small subgroup of individuals were assigned to

the same work at any given time, so the most appropriate individuals are those whose names appear 

on the known records for the individual as doing the same work at the same time and not, say, 

individuals doing the same work in a later period. However, if the number of same-time co-workers is 

deemed by the dose reconstructor to be too small to provide adequate statistics in view of the 

variation in the doses, the similarly assigned co-workers doing the same work in the period 

immediately before or after the period of interest should be included. The rotation system apparently 

ended when the doses went down sufficiently, probably some time in the early 1950’s. Hence from 

about 1952 on, the records will likely show more sustained assignment of the same work to an 

individual and identification of same-time coworkers would be less important.

7.3.1.1

Monitored workers may have received up to 150 mrep in a week without the dose being recorded in a 

dose summary. 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 favorable to claimant assumption is that any monitored worker 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 should be 

Notes Regarding Dose Summaries
Document No. ORAUT-TKBS-0005 Revision No. 02 Effective Date: 06/14/2007 Page 110 of 225

based on an average ratio derived from the weeks with specific dose monitoring entries for that 

individual. 

7.3.2

This section applies to workers for whom doses must be estimated for the operating years, i.e., 1949-

1958. Plants 1 and 2 were not used for AEC work after about 1946 and underwent decontamination 

and decommissioning (D&D) in 1948-1950, after which they were released to Mallinckrodt in 1951 for 

non-AEC work. Similarly, the other plants stopped being used at varying points in 1957-1958 and 

underwent D&D thereafter. This section does not apply to any worker or work area after the end of 

operations; see Section 8.0 for dose estimation for the D&D years and the later post-operations years.

Determining Dose - Monitored Period 1949 - 1958

After the operations and D&D years, Mallinckrodt received a license from the AEC for the use of 

certain radioactive materials and so some workers may have been continued to be badged. Also, 

there may be continuing badge records for some individuals for years after operations ended at the 

St. Louis (Destrehan) site because these individuals transferred to or made visits to the Weldon 

Spring site, also run by Mallinckrodt, and had badges issued for work there. . 

7.3.2.1

When there are missing data in the monitoring records, they should be filled in by bridging small gaps 

by interpolation and bridging larger gaps using statistical surrogate data and/or standard missed dose 

estimates.

Monitored Workers

7.3.2.2

As stated above, most workers who received significant occupational exposure were likely to have 

been monitored and thus if a worker was not monitored, especially for a lengthy period of 

employment, this indicates that he was deemed to have little or no exposure potential. To account for 

the possibility of an unmonitored individual receiving incidental exposure to photons, dose should be 

assigned to unmonitored Mallinckrodt employees as follows.

Unmonitored Workers - Plants Still Operating under MED/AEC Contracts

1. An attempt should be made to find a surrogate category. 

2. If no surrogate category can be identified – i.e., if there are no monitored workers with the 

same job title or the job title listed in the worker’s records cannot be associated with any 

surrogate job title – then either the Mallinckrodt subject matter expert for the dose 

reconstruction project should be consulted for guidance (and possibly further research) or the 

worker should be assigned a nominal category. Normally, the appropriate nominal category 

will be a low-exposure job title such as office worker.

7.3.3

Unbadged non-decontamination and decontamination workers, if any, should have their 1949-1950 

exposures determined on the basis of decontamination worker surrogate data (see Section 8.0).

Unmonitored Workers - Plants No Longer Operating under MED/AEC Contracts

For the period after decontamination and decommissioning, i.e., from 1951 on, the exposures should 

be determined as described in Section 8.0. 
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7.4 X-RAY DOSE

7.4.1

Employees of Mallinckrodt (cf. Section 5.4.4) received an annual occupationally required medical xray (MCW 1955d; Mason 1958a). Up to some time in 1944, a pelvic x-ray was also performed for 

employees working with fluorides (see Section 5.4.4). The annual chest x-rays and presumably the 

pelvic x-rays were taken at the Barnes Hospital (the Washington University School of Medicine) (AEC 

1950m; MCW 1955d) and the x-ray records remained the property of the hospital (AEC 1950m). 

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 favorable to claimant assumption would be that chest x-rays were 

performed annually from 1949-1962. Since the radiographs were made at a hospital, and hospital 

procedures routinely used both PA and lateral chest projections much more commonly than nonhospital facilities, the annual dose from chest x-rays for Mallinckrodt workers should consist of the

dose from both projections. No evidence so far indicates that photofluorographic chest films were 

performed. A physical checkup of all prospective decontamination and decommissioning workers was 

provided. This included chest x-rays (PA and lateral, using full-size 14 x 17 films) and it should also 

be assumed that subsequent x-ray examinations were performed annually. 

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.

7.4.2

Information to be used in dose reconstruction for the early years for which no specific x-ray technique 

information is available is provided in ORAUT-OTIB-0006, the dose reconstruction project technical 

information bulletin covering occupationally required medical x-ray procedures (ORAUT 2003c).

Chest X-Rays

7.5 NEUTRON DOSE

As indicated in Section 5.4.2, there was no neutron monitoring done at Mallinckrodt. Therefore it is 

not expected that there will be any neutron monitoring data or neutron dose rate measurement data 

found for Mallinckrodt. Section 5.4.2 describes the Mallinckrodt sources of neutron exposure. Table 

A-34 shows the results of the alpha-neutron dose rate calculations based on mixing of uranium and 

thorium with low atomic number materials such as fluorine and oxygen and also some calculated 

doses, together with the occupancy assumptions on which they were based. Assumptions not given 

in the table may be found in Section 5.4.2.

The annual neutron dose from the uranium is accounted for with a neutron:photon dose equivalent 

ratio of 0.2 [0.1 x 2 based respectively on ORAUT 2004, Table 6-10 and ORAUT 2004, Section 

6.6.2], which is assigned as the geometric mean of a lognormal distribution with a geometric standard 

deviation of 1.71, for a neutron energy range of 0.1 to 2.0 MeV as mentioned in Section 5.4.2 and as 

derived for Fernald (ORAUT 2004). These doses are assigned to workers who spent to time in the 

process areas and to residue workers (see table in Section 6.0).

The thorium neutron doses in Table A-34 should be considered as appropriate for thorium worker 

dose reconstructions. This dose (constant distribution) is applicable only for those years or parts of 

years in which the thorium was processed, i.e., from after July 1955 through approximately March 

1957.
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7.6 OTHER DOSE CONTRIBUTIONS

7.6.1

Given the relatively small number of claims and the unknown proportion requiring calculation of 

extremity dose, this subject is not treated in this document. Extremity dose estimates, when 

necessary, will need to be formulated on a case-by-case basis. See Section 5.4.3.5 for information 

about individual ring dose data and interpretation. 

Extremity Dose

7.6.2

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

7.6.3

As stated in Section 7.1, Item 6, Mallinckrodt dose records contain “beta” values obtained by 

subtracting the optical density of the film behind the cadmium shield from that behind the open 

window. These recorded values are assumed to be equivalent to Hp(0.07).

Shallow Dose
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8.0

As part of the process of dose reconstruction, the doses from postoperational decontamination and 

decommissioning activities and from postoperational work activities must be determined, insofar as 

they are due to residual contamination remaining after MED-AEC operations ceased. For this 

purpose, operations at the Mallinckrodt St. Louis main site are considered to have ended in 1957-

1958 and decontamination and decommissioning (D&D) of the various buildings used for MED/AEC

work to have begun in 1959 and continued through 1961; and the postoperational period to have 

extended through 1962. Doses from remediation activities in 1995 are to be included for workers who 

were exposed to radiation/radioactivity from this activity. Regular operations at SLAPS are 

considered to have ceased in 1958 and the Mallinckrodt caretaker/D&D operations period to have 

extended from 1959-1966. 

DETERMINATION OF EXPOSURES DUE TO RESIDUAL CONTAMINATION 

REMAINING FROM MED/AEC OPERATIONS

The dates of applicability for the tables in section 6.0 overlap the periods described in this section. 

For most workers, individual dosimetry and work descriptions will be available to make the 

determination of which intake values should be used. If such information is not available, the dose 

reconstructor should default to the exposure that is most favorable to the claimant.

8.1 DESCRIPTION OF THE DECONTAMINATION AND DEMOLITION WORK AT THE ST. 

LOUIS MAIN SITE, INCLUDING HEALTH AND SAFETY PRECAUTIONS

As noted in Section 3.0, for buildings in Plants 1 and 2, decontamination had been done some years 

earlier. Mason 1977 states that this started in 1949; Mallinckrodt (1994) states that there were 

programs to do so in 1950 and 1954 and even in 1970; and ORNL (1981) mentions only the 

decontamination of 1948-1950. Little is known of the Plants 1 and 2 decontamination campaign 

because although Mason (1977) states that detailed reports of the work, including the final surveys, 

were prepared, these records are not available. 

What is presently known is as follows. In 1946, Mallinckrodt took core samples of the subsoil under 

Buildings 51 and 52 in Plant 2 (MED 1946f), boring through the concrete floor, presumably at AEC’s 

behest. This indicates that even the earliest characterization effort likely took place after the start of 

film badging and urinalysis; decontamination workers may have been given badges and urinalysis due 

to the dust and dose rate potential of the work. The decontamination that began in 1949 was 

performed by “Main Plant” (Plants 1 and 2) crews under directions from Mallinckrodt Uranium Division 

“monitors,” as the health and safety specialists were called (Mason 1977). The contaminated waste 

material was collected and delivered to AEC, presumably in drums or boxes, while contaminated 

equipment was either recycled to Plant 6 or was transferred to AEC for disposal (Mason 1977). Some 

of the material apparently ended up buried at SLAPS. When the work was done to the satisfaction of 

the AEC St. Louis Area office, monitoring personnel from AEC-NYOO did a final survey of Plants 1 

and 2 (Mason 1977). When AEC-NYOO was satisfied, AEC released the plants to Mallinckrodt for 

unrestricted use in 1951 (FUSRAP 2002; DOE 1981). 

With regard to Plants 4 and 6, some preliminary work, such as some initial cleanup, was done in 1955 

and in 1958-1959 (AEC 1955b; MCW 1958; MCW 1959; MCW 1961b). For example, the process 

equipment was thoroughly cleaned by Mallinckrodt at the time the plants went into standby, before the 

formal decontamination and decommissioning work began (MCW 1961b). Some initial surveys to 

evaluate Plants 4, 6, 6E, and 7 were then performed (AEC 1955b; MCW 1958; MCW 1959; MCW 

1961b). It was determined that many buildings at the site were so contaminated that it would have 

been too expensive to clean them to usable levels, and they were torn down. All of the Plant 4 
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buildings and most of the Plant 6 buildings were torn down, but much of Plant 6E and 7 survived 

(Mallinckrodt 1994), amounting to about 20% of the MED/AEC plants at the site (MCW 1961b).

MCW (1961b) describes the health and safety (H&S) aspects of the decontamination and demolition 

(D&D) work at Plants 4, 6, 6E, and 7; this is the source of the information that follows in the rest of this 

section (8.1), except where another source is explicitly indicated. This work was performed by several 

construction and demolition companies apparently contracting directly with AEC. Mason (1977) 

states that one of them, or the principal one, was Arch Wrecking Company. H&S coverage was 

provided by Mallinckrodt, again under contract with AEC, because AEC considered that while the 

D&D companies were qualified to deal with normal hazards of the work, they were not experienced in 

the sort of chemical and radiological hazards they would encounter at the Mallinckrodt site (MCW 

1961b; Mason 1977). The Mallinckrodt H&S coverage was provided by experienced full-time 

personnel from the Destrehan (St. Louis) site and part-time personnel from the Weldon Spring site, 

which was also run by Mallinckrodt (MCW 1961b; Mason 1977). Oversight was provided by AEC

(MCW 1961b; Mason 1977); Mason (1977) states that only AEC made the decisions regarding items 

that could be released for restricted or unrestricted use. The Mallinckrodt Uranium Division performed 

a final decontamination survey of the four remaining plants for AEC in 1960-1961 and gave a report to 

AEC (Mason 1997). Only part of the report of this closeout survey is available.

A physical checkup of all prospective workers on the project was provided. This included chest x-rays

(PA and lateral, using full-size 14” x 17” films) and blood tests. Approximately 125 contract workers 

were examined. Work clothing was provided and post-work showers were required. 

Film badges were issued to all persons working in or visiting the Destrehan site during all phases of 

equipment removal, decontamination, and demolition. These were of the standard badge type always 

used at Mallinckrodt: an A. M. Samples stainless steel film badge holder with open-window and 

cadmium filters, using DuPont Type 552 dosimeter film; the gamma standard was a platinumencapsulated radium needle and the beta standard was an aged natural uranium block. Film badges 

were processed monthly, with 80 to 124 contractor badges processed each month. Film badges worn 

by Mallinckrodt and AEC personnel were those provided them through the Weldon Spring facility for 

their work at the Destrehan site in St. Louis.

Workers were also required to submit urine samples before the start of the project, at approximately 

six-week intervals during the project, and at the end of the project. The regular samples were taken 

on Friday after work and on Monday before work. Special samples were taken from those working in 

high dust exposure areas or in special operations such as sandblasting contaminated tanks. 

Urinalysis was done at Weldon Spring as a uranium-in-urine fluorometric analysis using a Jarrell-Ash 

fluorimeter, sodium fluoride as the fluxing agent, and a platinum dish as the sample holder. Each 

urine sample was analyzed in triplicate, with a 0.1 ml aliquot used each time.

AEC, the Mallinckrodt H&S people, and the D&D companies held pre-work planning discussions to 

determine how to do the work on the basis of the surveys. As work began in each area, additional

meetings and work area inspections were held to re-evaluate the approach for cleaning the particular 

area, based on how much contamination was revealed by equipment removal, how tenacious the 

contamination was, etc. Thus any of various decontamination and survey methods were used, 

depending on the characteristics of the area. Methods used for decontamination included mechanical 

“scratching” (to remove embedded particles on tar and gravel roofs); dry sandblasting; water blasting; 

shotblasting; steam cleaning with caustic-detergent additives; paint stripping; acid-detergent wash; 

immersion in solvents; burning; and sweeping, shoveling, and vacuum-cleaning.

Some particular operations performed and precautions taken were as follows.
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1. In scrap removal and decontamination, areas were first inspected to verify that there was no 

gross radioactive material content and to plan the means of containment, such as removal to 

containers or sealing. Approximately 80% of the equipment was handled during the D&D 

work.

To steam-clean scrap, surfaces were wetted using high-pressure fog or spray, to which a 

caustic-detergent agent had usually been added; the scrap was removed with hoists; and 

washing with water hoses was done, with runoff retained in the localized area. Grossly 

contaminated items were steam-cleaned on a grating placed over a vat, with the overflow 

going into the sewer through a weir. Half-face respirators were used for steam cleaning.

Paint stripping was also done where necessary prior to steam cleaning, with the stripper being 

applied by brushing or dipping, often on the grating over the vat. Nickel-bearing steel pieces 

were cleaned in the open air by lowering the pieces into a vat containing acid-detergents and 

allowing them to soak for several hours; then a wire brush was used to loosen the scale and 

steam-cleaning was used to wash it away. Solvents (chlorinate hydrocarbons) were used on 

some items of low contamination level; this cleaning was done in the open air using one vat for 

cleaning and one for rinsing, with the solvent recirculating through a 5-micron Cuno filter and 

with the workers wearing face shields and gloves. After draining, the items were dried with 

forced air.

For sandblasting of relatively portable items, a large walk-in hood in Building 116 of Plant 6E 

was reactivated. Tarpaulins were used to extend the hood area and to enclose the hood face 

partially. The hood exhaust was through a high-efficiency bag filter. For any item that could 

not be taken to this hood, a portable canvas hood was set around the item, with ventilation 

provided. About 60 vessels were cleaned by dry sandblasting; the remaining several vessels 

(5%) that could not be cleaned to required levels had been used in the pitchblende digestion 

and radium extraction operations.

Insulation was burned off cable with added fuel, followed by a water rinse.

2. In building and structure decontamination, the most effective and intensive surface cleaning 

was done by dry or wet sandblasting. Dry sandblasting was the preferred method for removal 

in depth of masonry materials, paint, and encrusted surface contamination. A method called 

Vacu-blast, using steel shot, was also used; it provided vacuum pickup of the resulting 

particulates, but recovery of the shot was time consuming. Full-face air-line supply masks and 

hoods were required, so care was taken to locate the air pump intakes upwind of the 

sandblasting operation or in a dust-free area. Generated dust was recovered or washed to 

sumps. Air changes were minimized by controlling outside openings; it was recognized that 

this increased interior concentrations in the short term, but it decreased exterior concentrations 

and, by decreasing air velocities in the area, accelerated the settling of the dust.

Water blasting was used on many roof structural parts and beams and was also used on other 

parts to remove loose particles before sandblasting. Scrubbing with a detergent and brush 

was also used. Sweeping, shoveling, and vacuum cleaning were generally done with 

respiratory protection, due to the dust.

3. In decontamination of concrete pads in the yard, dry sandblasting was done in a canvascovered booth on wheels, which was moved by two workers to sit over the spot to be blasted. 

Ventilation in the booth was provided by a fan discharging through a large furnace cleaning 

bag; the materials collected in the bag were removed every time the bag was moved. The 
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operator in the booth wore supplied-air respiratory protection and his helper outside wore a 

half-face mask. It was noted that the radiation levels of the particles collected by the fan and 

caught in the filter bag were higher than the levels of the coarser particles that remained on 

the surface being cleaned, so that capture of radioactivity appeared to be efficient.

Dust from process areas was considered to be contaminated even before sampling and thus masks at 

a minimum were required for all dust-producing and fume-producing operations, including torch 

cutting. Respirators used for radiological hazards included high-efficiency mechanical dust 

respirators; bottled-air-supplied masks (for short-duration entries into high-concentration atmospheres 

and emergency use); and airline-supplied masks and hoods. Airline-supplied masks and hoods were 

required for sandblasting operations. The half-face dust masks were typically the MSA Dustfoe 66.

Regarding interior air sampling, samples were taken on a routine basis during the work shift, with 

filters being changed at the end of each daily sampling period; also, additional samples were drawn, 

either in the general work area or downwind, during the particular operations listed above and during 

other scrap cleaning. Air sampling was done with high-volume air samplers by Gelman Hurricane, 

Staplex Type TF1A, or the equivalent. The collection medium was Whatman #41 or HV-70 filter 

paper or Gelman AM-4 membrane filters. Sampling rates averaged 5 liters of air per square 

centimeter of filter media. Analysis for uranium was done by acid-leaching the filter, then counting an 

aliquot by photofluorometric analysis. 

It was estimated that about 1% of the work during equipment removal, decontamination, and 

demolition was in areas where the airborne uranium concentration was on the order of 10-100 MPCa; 

for this category, full-flow air-supply masks were used. 10% of the effort was estimated to have been 

spent in areas where the concentration was 1 to 10 MPCa; for this category, personal half-face, dusttype masks were used. The rest of the work was done in areas of less than 1 MPCa. General area 

(“zone”) concentrations during sandblasting were found to be higher around the local tarpaulin 

enclosures than around the fixed hood, but still below the (unspecified) AEC guide levels. 

Regarding exterior air sampling, the generation of dust during demolition was at or near ground level; 

there were no discharge points to the upper atmosphere since the stacks were not being used. Dust 

sampling stations were established around the property line so as to have at least one always 

downwind of the work. The average of the uranium-in-air concentrations measured outside the 

buildings during demolition was 4 x 10-13 uCi/cc, said to be about one-tenth of the MPCa. It was 

observed that the gross alpha concentrations in the outside air were not much different from the 

uranium measurements, leading to the conclusion that radionuclides other than uranium were not 

present in significant quantities. The gross beta activities measured were judged to reflect the 

background seen at the time due to nuclear weapons testing.

The cleanup goals were surface alpha activity, 1000 dpm/100 cm2 average and 5000 dpm per 100 

cm2 maximum spot; beta-gamma activity, 0.1 mrep/hr average and 1.0 mrep/hr maximum spot. This 

was achieved for the site and buildings, although the average beta-gamma level in some areas was 

somewhat above 0.1 mrep/hr. For alpha radiation surveys, the survey meters were calibrated in dpm

per 100 cm2 against normal uranium; they could be of the air ionization, proportional, or scintillation 

type and could detect 500-50,000 dpm per 100 cm2

, depending on type. For beta-gamma activity 

surveys, the survey meters were calibrated in mrad/hr or mrep/hr against normal uranium; they could

be of the Geiger tube, air ionization, or scintillation type, could measure 0.3 to >1.0 mrad/hr, 

depending on type, and had to have a minimum unshielded probe face area of 2 square inches. 

For surveys of removable contamination, for both alpha and beta-gamma, two wipe passes with half 

of a 4” disk of Whatman paper over 100 cm2 were performed. All material items exceeding 50 lbs 
Document No. ORAUT-TKBS-0005 Revision No. 02 Effective Date: 06/14/2007 Page 117 of 225

were monitored separately. Material items weighing less than 50 lbs, except for pipes, were 

monitored as a lot using a 20% random sample, providing the lot was from the same plant sources; if 

destined for uses other than smelting, they were monitored individually. Items were monitored both 

inside and out, for both fixed and removable contamination. If an item could not be monitored 

internally, it was considered contaminated. If it could be, then for beta-gamma, an item was 

considered to be contaminated if any surface scan reading was greater than 1.0 mrep/hr or if any 

smear exceeded 0.3 mrep/hr; no wipe test was done if all scan readings were less than 0.3 mrep/hr. 

For alpha, an item was considered to be contaminated if any surface scan reading was greater than 

25000 dpm per 100 cm2

, if the average of the surface scan readings was greater than 5000 dpm per 

100 cm2 or if any smear exceeded 2000 dpm per 100 cm2

; no wipe test was done if no scan reading 

exceeded 2000 dpm per 100 cm2

. 

Because alpha probes were too large to be put inside pipes but Geiger tubes were not, the betagamma measurement was used as a surrogate for the alpha measurement; this was said to be 

because Th-234 and Pa-234, both beta emitters, reached equilibrium with uranium in approximately 

250 days. However, it was recognized that nonequilibrium conditions or the presence of other beta 

and gamma-emitting isotopes could significantly affect the accuracy of the estimate of inferred 

uranium content. Pipes were not sampled as a lot. They were cleaned and unbent or unkinked, when 

possible, before monitoring. The presence of visible uranium deposits automatically caused a pipe to 

be classified as contaminated, as did the presence of bends, elbows, valves, fittings, or crushed 

sections. A Thyac meter on a ten-foot pole (“ten-foot probe”) was used for long sections.

Final contamination measurements are given in detail in MCW (1961a) and in summary in MCW 

(1961b). During the final survey, the following instruments were used: for alpha, the Eberline PAC3G gas proportional counter and the Victoreen Model 356 ionization chamber survey meter; for betagamma, the Victoreen Model 389C Thyac (G-M) survey meter, background about 0.1 mrep/hr. 

Calibrations were as given above. A spot sampling method was used for wall, ceilings, and beams, 

while a path scanning method was used for floors, yards, pads, and roofs. The alpha measurements 

for the roofs, however, were spot readings. During demolition and the final survey, it could be seen 

that some amount of residual contamination was buried at the site. This contamination was generally 

located adjacent to foundations, between joints and concrete floors, under storage pads, and between 

joints and structural steel members. This was judged to be fixed and not to present a problem.

Table A-37 shows the experience reported on a group-wise basis in MCW (1961b), including film 

badge, urinalysis, and air sampling results.

8.2 POST-DECONTAMINATION CONDITIONS AT THE ST. LOUIS MAIN SITE

After the decontamination of Plants 1 and 2, the plants were released to Mallinckrodt in 1951 

(FUSRAP 2002; DOE 1981). No information about post-decontamination conditions for Plants 1 and 

2 appears to have been produced until the ORNL survey of 1977 (ORNL 1981).

Following the decontamination or demolition of the buildings composing Plants 4, 6, 6E, and 7; postdecontamination use of the site with residual MED/AEC contamination began. This is considered to 

have started in 1962 and continued through 1995, the year when the Formerly Utilized Sites Remedial 

Action Project (FUSRAP) took over the site for final decontamination to modern standards. But in 

fact, release of the site to Mallinckrodt may have occurred plant by plant over 1961-1962, as 

suggested by Mallinckrodt (1994). A survey was done at the end of the decontamination period to 

demonstrate that the site was ready for release by AEC. This survey is documented in MCW (1961a).
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No further survey of any of the plants to gauge the residual contamination appears to have been done 

in the post-D&D period until 1977, when a survey team from Oak Ridge National Laboratory did a 

detailed contamination, radiation, and radon survey of the site (ORNL 1981). A summary of the 

ORNL contamination findings is shown in Table A-38. Also shown in Table A-38 are radon 

concentrations reported in 1981 by ORNL (1981) and in 1986 and 1990 by Applied Nuclear Safety 

(1986; 1991). These show that the post-decontamination radon levels were low compared to the 

levels of the operational years (Table A-25).

8.3 GENERAL CONSIDERATIONS AND APPROACH FOR ST. LOUIS MAIN SITE 

DECONTAMINATION AND POST-DECONTAMINATION DOSE RECONSTRUCTION 

As noted in Section 8.0, exposures during the D&D work appear to have been completely monitored, 

with film badge, urinalysis, and supplementary air sampling data being available. For individual D&D 

workers for which personal monitoring data are unavailable, surrogate exposure data could possibly 

be assembled. 

An applicable note in MCW 1961b should be mentioned here with respect to surrogate data for the 

D&D period. MCW (1961b) stated that Mallinckrodt experience at the Destrehan plant during the 

operating years was that for individuals whose daily integrated particulate air exposure was 50 μg of 

uranium per cubic meter of air, the after-work urine samples collected on Friday typically were in the 

range of 0.050-0.060 mg per liter and the before-work samples collected on Monday were in the 

range of 0.025-0.030 mg per liter. The D&D urine samples typically were found to be about half this 

level. (Note that the reference gave 50 “mg” per cubic meter of air as the daily integrated exposure, 

but it is likely that 50 micrograms per cubic meter was meant, since that was the standard tolerance 

level and would have been the logical quantity to cite.)

It should be noted regarding the post-D&D years that although some employees (remaining) at the St. 

Louis main site may be found to have urinalysis records extending into the late 1960’s, this monitoring 

was apparently done for Mallinckrodt’s post-AEC work with radioactive materials (e.g., its columbiumtantalum processing operations). These urinalysis records may be helpful in establishing the 

elimination curve of radioactivity from the body for anyone who (1) worked in the uranium processing 

operations up to 1958 and then worked in Mallinckrodt’s non-AEC operations or (2) was involved in 

the D&D work and then continued to work at the site in Mallinckrodt’s non-AEC work. Although, these 

records may be misleading because they may indicate new intakes from non-AEC sources, they can 

be used to bound dose. Also, people who worked in the decontaminated and released buildings from 

the release date on would likely not have been covered by film badge and urinalysis programs; in any 

case, persons monitored after 1961 or 1962 were not monitored under the aegis of AEC. Thus 

bioassay or external dose information for them, if any, may not be available in dose reconstruction 

project files. 

8.4 GENERAL CONSIDERATIONS AND APPROACH FOR SLAPS POST-OPERATIONS 

DOSE RECONSTRUCTION

Because SLAPS did not receive wastes on a regular basis after the end of Mallinckrodt St. Louis main 

site operations in 1958, the “postoperations” period for SLAPS may be considered to have started at 

that point, i.e., in 1959, although it appears that some contaminated materials from Mallinckrodt D&D 

were brought to SLAPS after 1958. The end of this period is set on December 31, 1966, although it is 

likely that there were few if any Mallinckrodt workers associated with the site after 1962, when a 

private entity was granted an AEC license for site caretaking and operations. 
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There was no decontamination of the site per se during the period 1959-1962. However, some 

material was removed and some new material was added, as follows. In April 1959, railroad siding 

and loading facilities were installed, apparently in preparation for moving out the residues and other 

material (AEC 1959); the tailings from the magnesium fluoride slag processing were then sent to the 

Fernald site (AEC 1959). This appears to have been the only removal of residues from the site during 

this period (the rest were removed after the transfer of control to the private entity). Additionally, 

some of the scrap and rubble from the demolition and decontamination of the main St. Louis site 

facilities was buried at the west end of the site (AEC 1972; ORNL 1979) in about 1959-1960; this was 

the same area where contaminated metal and debris had been buried previously (presumably in 

1948-1950, following the decontamination of Plants 1 and 2), and where in 1952 a large area was 

filled in deeply with earth (AEC 1964).

It is not clear who did the work, but it seems likely – and is assumed here – that the burial, at least, 

was done by the same contractor(s) as was (were) doing the demolition and removal work at the St. 

Louis main site. Thus it appears likely – and is assumed here – that this work was at least overseen 

by AEC and Mallinckrodt health and safety people. Because the work was done in the same period, it 

is similarly assumed that the railroad construction and the tailings removal were performed with AEC 

and Mallinckrodt health and safety oversight and support. These assumptions are strengthened by 

the fact that at some point in about 1959, a building consisting of a changeroom, a shower, and office 

space was moved to the site (AEC 1959; AEC 1964).

No information is available as to the conditions under which the railroad facility installation and the 

tailings removal were done, but presumably these were typical construction and earthmoving 

processes such as had taken place previously at SLAPS. No information is available as to the 

conditions under which the 1959-1960 burial was done, but presumably it was a typical process of 

digging a hole(s) in the ground, depositing the debris, and pushing the removed soil back into the 

hole. As noted above, there was a building installed that included a changeroom, shower, and office. 

Thus it appears that normal contamination control procedures of the time were followed.

The workers might not have been badged, however, because the potential for exposure may have 

been viewed as too low; for example, during 1949, the crane and bulldozer operators at SLAPS (who 

were apparently working only with the residue piles and not with any aspect of the K-65 storage or 

removal operation) were not being badged because they averaged less than 8 hours per week on the 

raffinate heap work and the radiation level from the raffinate/residue heaps was believed to be low 

(MCW 1949g). If the workers were badged, however, this was undoubtedly done through the Weldon 

Spring site, as was done for the decontamination and decommissioning workers at the main St. Louis 

site. Thus it is not clear if the workers were badged but in any case the Weldon Spring film badge 

records at the time this paragraph was written were unavailable.

Before and after the burial and tailings removal work, it can be assumed that only caretaker functions 

were performed, that is, that periodically a guard would visit the site to check the locks and verify that 

no entry into or disturbance of the site had been made. In addition, there were likely some inspection 

visits by AEC in the company of prospective purchasers of the residues. The AEC people may have 

been wearing their own badges, but there is no indication that the visitors did.

This document covers Mallinckrodt employees only, but an effort has been made to provide 

information about other workers on the Mallinckrodt St. Louis main site and SLAPS, for information 

and for efficiency of later evaluation of the other workers’ exposure, if required. Thus the categories 

of exposed workers are assumed to include all those below, but they should not be assumed to be 

Mallinckrodt employees. Visitors are not included.
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Note that in 1976 and 1978, Oak Ridge National Laboratory surveyed SLAPS for AEC years after the 

piles had been removed (ORNL 1979). They found that although much of the soil inside the SLAPS 

fence had terrestrial levels of U-238 and Ra-226, there were several places where the levels were 

elevated. The highest U-238 level was 260 pCi/g (normal range 0.3 – 1.7) and the range of elevated 

Ra-226 values was 1.4 – 78 pCi/g (normal range 0.3 – 1.3). Ac-227 was also found in the range of 

0.5 – 77 pCi/g; the source was most likely the Sperry cake because of its concentration of Pa-231, a 

precursor of Ac-227, but the highest reading was found where the AM-7 was stored and some 

elevated readings were also said to be associated with a location near where the barium sulfate cake 

was stored.

8.5 CALCULATION OF DECONTAMINATION AND POST-DECONTAMINATION INTERNAL

AND EXTERNAL EXPOSURES FOR THE ST. LOUIS MAIN SITE WHEN INDIVIDUAL 

AND SURROGATE DATA ARE LACKING

To estimate doses to workers from MED/AEC contamination alone when individual data and 

surrogate data from records are lacking, use was made of the results of the initial and release 

Mallinckrodt contamination and dose rate surveys, i.e., those performed before and after the 1959-

1961 decontamination (reported in MCW 1958 and MCW 1959 for before decontamination; in MCW 

1961a for after). The data taken from these references and used in the calculation of exposure are

shown in summary form in Table A-39; it must be stressed that these represents only a small and 

select subset of data from larger sets of hundreds of data points.

The RESRAD-BUILD computer code (ANL 2003) was used to calculate annual exposures from 

inhalation of airborne particulates and radon (and its daughters). Separate hand calculations were 

performed to estimate annual ingestion and external exposures.

Source terms for both types of calculations were derived from the measured data shown in Table A39. Conservative maximum averages of surface and bulk (volume) contamination were used to 

produce the inhalation and radon source terms for RESRAD-BUILD and the inhalation RESRADBUILD results were used to produce the source term for the ingestion calculations, while the 

maximum (or average maximum) and typical measured dose rates from gamma and beta radiation 

were used in the hand calculations to estimate annual external exposures.

For RESRAD-BUILD parameters other than the source term, conservative but Mallinckrodt-suitable 

values were used, when they could be determined; when no specific or suitable values could be 

determined, conservative default values given in the RESRAD-BUILD manual (ANL 2003) or other 

guidance documents were used. These values are considered favorable to claimants. 

8.5.1

The principal assumptions made for the RESRAD-BUILD calculations are given below.

Assumptions Made in RESRAD-BUILD and Related Internal Exposure Calculations

1. The inhalation and radon source terms were derived on the basis of the highest average 

surface and bulk contamination levels, respectively, in each plant, regardless of building or 

room location. Without regard to the actual correspondence of gamma dose rate, surface 

contamination, and bulk contamination levels in particular rooms, the room model (work area) 

was assumed to contain the highest measured average surface contamination concentration, 

the highest bulk contamination concentration, and the highest spot or average gamma and 

beta dose rates found anywhere in the respective plant (given in Table A-39). This 

assumption is favorable to the claimant, since there was no location where all these indicators 

were simultaneously at the maximum.
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2. The room model was assumed to have the measured surface concentration over all wall and 

floor surfaces and to have the measured bulk concentration extend 6 inches into the walls and 

floors, which were taken to be concrete. This is reasonable because in MCW 1958 

contamination in concrete was measured typically from 4 to 6 inches into the concrete (but 

much less in steel) and MCW (1958; 1959; 1961a) generally found the overheads to be far 

less contaminated than the walls and floor and so the overheads would contribute negligibly to 

the total.

3. Contamination was measured as gross alpha and as either total beta and total gamma 

separately (MCW 1958; MCW 1959) or total beta-gamma together (MCW 1961a). 

4. The worker was assumed to spend his entire work time (8 hours per work day) in the room, 

i.e., in the most contaminated area of the given plant. 

5. The worker was assumed to spend 2000 hours per year in the one location. The takedown of 

a building may have been on the order of weeks and decontamination on the order of months; 

however, continuous decontamination and demolition work over the course of a 50-week year 

is assumed for such workers. 

6. The room size was taken to be about 10’ x 20’ x 10‘ (3 m x 6 m x 3 m). There were many 

process areas that were larger, but they were often partitioned and they were undoubtedly 

decontaminated in sections. Thus assuming a smaller room would be conservative in terms of 

concentrating or confining the contamination in the ventilated space.

7. One air change per hour was assumed. While only limited information is available regarding 

the ventilation systems at Mallinckrodt, it was clear that the process areas had forced 

ventilation. These were apparently not always used during significantly dusty work, as 

suggested by MCW (1961b), but in those cases respirators were worn by the workers and 

enclosures were typically used as well. The enclosures had forced ventilation (MCW 1961b). 

Thus it is reasonable to assume that either the normal forced ventilation was used in the 

general area, in which case one air change per hour is a conservative rate, or vented 

enclosures were used, in which case the air change rate would have been far higher, the 

worker would have been wearing a respirator, and the calculated exposure would represent a 

marked overestimate of the likely actual exposure.

8. For the decontamination period and the post-decontamination period, the resuspension factors 

for the transferable contamination was assumed to be 1 x 10-4; and 1 x 10-6, respectively. The 

latter value is based on NRC (2002) and the former is taken to be a favorable to claimant

value for non-respirator work, as is consistent with the discussion and tables in the RESRADBUILD manual (ANL 2003). 

9. The deposition (settling) velocity was taken to be 0.00075 m/sec, a reasonable value for 

particles of 5 μm, as shown in Figure J.3 of ANL 2003.

10. The removable fraction for the decontamination period was assumed to be 30%, based on the 

fact that some early decontamination was done at the end of operations (e.g., rinsing out the 

process vessels and vacuuming the floors). The removable fraction for the inhalation dose 

calculation in the post-decontamination period was assumed to be 10%. This should be 

reasonably favorable to the claimant, since the post-decontamination period followed an 

extensive decontamination. Default erosion, radon emanation, and associated values were 

used because they are conservative and thus favorable to the claimant. 
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11. Three types of exposure levels were used to represent the different exposure potentials of 

different types of workers. “High” exposure potential represents those working in the most 

contaminated areas, i.e., the former process areas; such as employees performing 

decontamination or later working substantial periods of time in the former process areas. 

“Moderate” exposure potential represents those accessing the less contaminated areas or 

infrequently accessing the former process areas; such as employees supervising the 

decontamination on an intermittent basis or those having only occasional need to enter the 

former process areas. It would also include employees working in the former laboratories. 

“Low” exposure potential represents employees accessing the slightly contaminated or 

uncontaminated areas, such as the former office areas.

12. Since the inhalation and radon calculations did not depend on the position of the receptor in 

the room model and since the radionuclide proportions were taken to be the same at the 

beginning of the calculation, one wall of the maximally contaminated plant was modeled with 

the D&D sources for the surface case and similarly for the volume case; the same was done 

with the post-D&D sources for the surface and volume cases. The results were then ratioed to 

produce results for the entire wall and floor area for the various plant and exposure potential 

cases. 

13. Because 1959-1962 data for Plants 1 and 2 were not available and because the post-D&D 

surface and volume data of Plants 1 and 2 were bounded by the Plants 4, 6, 6E, and/or 7 data, 

as shown by comparisons of the data for all the plants (ORNL 1981), it was assumed that 

workers in Plants 1 and 2 could be assigned the maximum post-D&D inhalation exposures for 

the “High” category from calculations for the other plants. Thus it was assumed that it was not 

necessary to create separate source terms and models for inhalation exposures for Plants 1 

and 2.

14. Because 1959-1962 data for Plants 1 and 2 were not available, it was assumed to be 

appropriate to use the post-D&D calculated radon exposure data for the other plants. While 

the typical radon measurement data of Plants 1 and 2 (ORNL 1981) were in the same range 

as the radon measurement data for Plants 4, 6, 6E, and/or 7 (ORNL 1981), there was one high 

area reading in one building in Plant 1 and one in Plant 2 that far exceeded the other readings. 

Although the high reading may not be representative of the true exposure, it was assumed that 

the maximum calculated exposure for the year 1975 (the evaluated year closest to 1977, the 

year the ORNL survey was done) could be ratioed by the factor or difference seen in the 

ORNL 1981 data for the respective plants and the calculation. Also, due to the lack of 

variation in the calculated doses over time, it was assumed that the calculated 1962 post-D&D 

results, after ratioing, could be extrapolated back to 1959 for Plants 1 and 2 (since they had 

already been cleaned up and were not in D&D). Thus it was assumed that it was not 

necessary to create separate source terms and models for radon exposures for Plants 1 and 

2.

8.5.2

For use in dose reconstruction, the inhalation and radon doses that were the results of computations 

in RESRAD-BUILD had to be converted back to activity units, in this case to pCi and WLM of intake, 

respectively. The RESRAD family of codes uses the dose conversion factors for inhalation given in 

Eckerman et al. (1988), as also listed in the RESRAD-BUILD manual (ANL 2003). The radon 

conversion is also from the RESRAD-BUILD manual (ANL 2003). Since the conversion factors are 

Results of RESRAD-BUILD Calculations and Conversion to Dosimetrically Useful 

Quantities
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applied at the end of the RESRAD-BUILD calculation, it is appropriate to reverse the conversion using 

the factors. The converted results, assumed to be a constant distribution, are given in Table A-40. 

Because of RESRAD-BUILD’s limitation on how many yearly printouts can be made, inhalation and 

radon exposures were calculated for each year for the D&D period but only for the first few years and 

every five years thereafter for the post-D&D period. This is clearly appropriate since as the output 

data show, the values change little from year to year. So although multiple years may be indicated in 

the column headings in Table A-40, the figures below them are for each year and are not the sum for 

the indicated years. The results are assumed to be a constant distribution.

8.5.3

It was assumed that it was appropriate to use the annual inhalation intakes derived from the 

RESRAD-BUILD inhalation dose results in order to produce source terms for the calculation of 

bounding ingestion doses according to the methodology of NIOSH (2004).

Assumptions for and Results of Ingestion Calculations

1. The appropriate inhalation exposures in pCi that were calculated as in Section 8.3.2 from the 

RESRAD-BUILD results were converted from pCi per year to an effective air concentration in 

pCi/m3

, assuming a breathing rate of 1.2 m3/hr and a 2000-hour work year.

2. The air concentration in pCi/m3 was multiplied by 0.2 to obtain the pCi ingested per day.

3. The pCi ingested per day was multiplied by 250 work days per year to obtain the pCi ingested 

per year.

The results, assumed to be a constant distribution, are shown in Table A-40. 

8.5.4

For the external exposure calculations, separate calculations were performed for gamma and beta 

radiation. For Plants 4, 6, 6E, and 7, the high maximum average and typical average measured dose 

rate values in accessible areas were used for the “High” exposure potential case; moderate maximum 

average and typical average values for the “Moderate” case; and low maximum average and typical 

average values for the “Low” exposure case. These are defined as in Item 11 of Section 8.5.1 above.

Assumptions for and Results of External Exposure Calculations

1. Because exposure (dose) rates were used, the source terms did not have to be translated into 

activity units. However, while the pre-decontamination survey measured values were given 

separately as gamma and beta dose rates (MCW 1958; MCW 1959), the final release survey 

gave values as combined beta-gamma dose rates (MCW 1961a). These were ratioed using 

the pre-decontamination data in order to produce separate gamma and beta dose rates. The 

resulting assumed source terms are shown in Table A-39.

2. In the surveys, the measurement point for betas or mixed beta-gamma radiation was usually a 

contact or near-contact dose rate for both walls and floors, while for gammas, it was most 

often a contact or near-distance dose rate for walls and a three-foot measurement for floors. 

However, the gamma-only measurements with a microrem-reading detector was usually taken 

in the middle of each survey grid block. It was thus assumed that the measured beta dose 

rate and the measured mixed beta-gamma dose rate represented all-beta radiation emanating 

from a wall surface to a receptor point at 1 cm from a wall surface; similarly, the measured 

gamma dose rate was assumed to represent gamma radiation emanating from a wall surface 

to a receptor point at 1 meter from a wall surface.
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3. A series of calculations were done to see what size of source (e.g., point, small-radius, and 

large-radius) was most appropriate for the measured data for each type of radiation. It was 

found for both beta and gamma that a large-radius source was most appropriate. For the 

gamma case, it was assumed that the source was of infinite radius (because it was not a very 

large increase from, e.g., a 4-m radius while the room could be assumed to be on the order of 

the room used for the RESRAD-BUILD calculation, or about 3.3 x 6.7 meters for the wall 

lengths). For the beta case, it was assumed that the source was of essentially infinite radius, 

i.e., 8.5 m, the range of the most energetic beta emitted from the uranium-daughter source 

mix.

4. The dose rates as per Item 2 above were used to determine the areal source strength for beta 

and gamma separately and then the source strengths were used to calculate the dose rates at 

1 foot and 1 meter for beta and at 1 cm and 1 foot for gamma. As per NIOSH direction, the 

respective dose rates at 1 cm were then considered to be the maximum dose rates, the dose 

rates at 1 foot the most likely dose rates, and the dose rates at 1 meter the minimum dose 

rates.

5. The receptor was assumed to stay in the maximum average location for his exposure potential 

for two hours per workday and in the typical average location for his exposure potential for six 

hours per workday, for a total time of 2000 hours per year. This ignored break time, which 

was usually spent in areas of very low or no contamination, such as a lunchroom.

6. For Plants 1 and 2, there were not sufficient data to create the three dose potential categories 

and there were data for only the one time (i.e., the 1977 ORNL survey data point (ORNL 

1981)). Thus there is only a single category of dose potential and the same measured 

numbers are used for all the post-D&D years. Because the figures were generally mixed betagamma and not separate gamma and beta exposure rates, except for one specific gamma 

measurement, the maximum value was assumed to represent gamma and beta in turn, except 

for that one gamma measurement. It should be recognized that the exposures calculated in 

this way for Plants 1 and 2 are likely to be extremely conservative.

Once the annual gamma doses had been calculated for each case from ambient measurement data, 

the doses had to be converted to photon doses to individual organs. Conversion factors from 

Appendix B of NIOSH (2007) were used to do this. It was assumed that for gammas, the dose was 

from photons of energy between 30 and 250 keV. A geometry of 50% AP and 50% ROT was 

assumed, as seems reasonable from Section 5.4.3.1 above. The results are shown in Table A-41

and represent a triangular dose distribution. 

The beta dose rate data given in Table A-39 above was used to calculate electron doses to skin, 

breast, and testes. To account for attenuation by clothing (see Section 7.6.3), a factor of 0.8 was 

applied in the calculation of the breast and testes doses. The results are shown in Table A-42 and

represent a triangular distribution. 

8.6 CALCULATION OF DECONTAMINATION AND POST-DECONTAMINATION INTERNAL

AND EXTERNAL EXPOSURES FOR SLAPS WHEN INDIVIDUAL AND SURROGATE 

DATA ARE LACKING

When there are no individual monitoring data to estimate exposures through 1962 and an individual 

worked at both Mallinckrodt and SLAPS, the internal exposures specified in section 6 should be 

assigned; a similar approach should be taken to estimate external exposures prior to 1962. There 

does not appear to be any work time or work practice information available for SLAPS for the period 
Document No. ORAUT-TKBS-0005 Revision No. 02 Effective Date: 06/14/2007 Page 125 of 225

1959-1966. There also does not appear to be any external dose rate, radon, or dust concentration 

data available for SLAPS for this period. Thus work conditions must be inferred that are judged to be 

favorable to the claimant based on indications in the reference; source terms or dose rates must be 

inferred from the data from other periods, chosen because the known sources of those other periods 

are comparable to or higher than those for 1959-1966. 

The exposure modes during the removal work at SLAPS are considered to be external gamma and 

beta radiation, radioactive dust; and radon. The radioactive sources would have been any residues 

dumped on the ground in or near the work area. It is not clear which residues were dumped where on 

the site, thus it must be assumed that the digging for burial and removal was near the worst source. 

(It is assumed that burial would not have been done under the piles, as that would have required 

moving more material and would also have involved more health hazards.)

The worst source appears to be the barium sulfate (AJ-4) wastes. (Note that the K-65 had been 

removed years earlier.) First, although as Table A-4 indicates there was more total uranium in the 

AM-7 and AM-10 raffinates -- 113 and 48 tons respectively versus the 29 tons total in the leached and 

unleached barium sulfate residues – the barium sulfate contained the residual radium that had not 

been captured in the K-65 precipitation. As Table A-4 shows, this was about 1 mg per ton. The 

radium would contribute to both the external and radon exposures. Second, even after the removal of 

the wastes in the late 1960’s, a beta-gamma survey showed that the area that had contained the AJ-4 

residues read higher than any other area (AEC 1972), also suggesting a high radium content 

compared to the other two residues. Third, as Table A-33 shows, the gamma dose rates were highest 

from the “aged” barium cake material, which would correspond to the piles found after 1958 at 

SLAPS. Thus the assumption is made that the calculated doses can be based on barium sulfate as 

the maximal residue.

8.6.1

It is not clear how long the burial and removal work took, but it does not appear to have been a 

continuous process since the generation of material during the St. Louis main site D&D would not 

have been continuous. Thus it is assumed that the work took place over a period of six months, 

evenly spaced out over 1959-1960. The excavation/removal workers were assumed to be present for 

8 hours per day, while the Health and Safety workers and AEC representatives were assumed to be 

present for 4 hours per day. The guard is assumed to have been present for two hours a day (to open 

up the site and lock up, if necessary) during the work and for one hour per week during the remainder 

of the time through1966. Health and Safety and AEC representatives are assumed to have spent 

two hours per month (i.e., 12 two-hour visits) at the site during 1961-1966. The fractions of the workyear (2000 hours) spent at the site are thus as shown below.

Assumptions for Work Time and Worker Types

Fraction of the work year spent at SLAPS.

1959 1960 1961-1966

Bulldozer/truck driver, crane operator 0.25 0.25 ---

Other material workers (e.g., riggers) 0.25 0.25 ---

Health & Safety worker 0.125 0.125 ---

AEC 0.125 0.125 0.012

Guards 0.081 0.081 0.025

8.6.2

AEC (1949m) states that the barium sulfate residue contained 0.1% uranium (by weight) and 4 x 10-9

grams of Ra-226 per gram of residue. Assuming that the other daughter isotopes of U-238 were in 

Assumptions for Radioactive Source Terms
Document No. ORAUT-TKBS-0005 Revision No. 02 Effective Date: 06/14/2007 Page 126 of 225

equilibrium with their parents and assuming a natural uranium mix, this gives the following 

radionuclides of internal dose interest: 

Concentration of radionuclides in barium sulfate residue. 

Isotope Quantity, pCi/g Activity fraction

U 6.83E+02 5.24E-02

Ra-226 3.96E+03 3.04E-01

Th-230 3.35E+02 2.57E-02

Pb-210 3.96E+03 3.04E-01

Po-210 3.96E+03 3.04E-01

Pa-231 1.56E+01 1.21E-03

Ac-227 1.56E+01 1.21E-03

For each pCi of inhalation or ingestion intake, assume the activity fractions in the table above. 

8.6.3

It is conservative to assume that the residue – which was deposited loose on the ground in piles – can 

be treated in calculations as though it were soil. ORNL (1979) performed a calculation of the 

concentration in air due to wind alone and due to mechanical resuspension (i.e., excavation or other 

soil disturbances), based on measurements of the radioactive content of SLAPS soil and on 

measured wind speeds and directions at SLAPS. Although these results are not directly applicable 

since the residues had been removed by the time the measurements and the calculation were done, it 

appears that the assumptions made in the calculation were such that it can be scaled to the actual 

situation, that is, by treating the barium sulfate residue as though it resuspended in the same way as 

the soil when disturbed by wind or mechanical action.

Assumptions for Inhalation Exposure Calculations

Inferred values of airborne dust concentrations for barium sulfate residue piles.

Radionuclide

Reference point (ORNL 1979) Ratios Barium sulfate

Wind 

resuspension, 

pCi/m3

Mechanical 

resuspension, 

pCi/m3

Mean 

soil 

conc, 

pCi/g

Ratio, wind 

resupension/

mean soil 

conc

Ratio, mech

resupens/

mean soil conc

Mean 

residue 

conc, pCi/g

Wind 

resuspension, 

pCi/m3

Mech 

resuspension, 

pCi/m3

Ra-226 1.00E-08 0.03 12.6 7.94E-10 2.38E-03 3.96E+03 3.14E-06 9.42E+00

Th-230 1.00E-08 0.03 8.3 1.21E-09 3.63E-03 3.35E+02 4.05E-07 1.22E+00

Ac-227 6.00E-09 0.01 8.8 6.80E-10 1.13E-03 1.56E+01 1.06E-08 1.76E-02

Pb-210 1.00E-08 0.03 - 7.94E-10 2.38E-03 3.96E+03 3.14E-06 9.42E+00

Po-210 - - - 7.94E-10 2.38E-03 3.96E+03 3.14E-06 9.42E+00

Pa-231 - - - 6.80E-10 1.13E-03 1.56E+01 1.06E-08 1.76E-02 

Ua 8.16E-08 0.20 16.9 4.83E-09 1.21E-02 6.83E+02 3.30E-06 8.26E+00

a. Activities in the uranium row were adjusted upward by a factor of 683/335 from those in revision 01 of this document to 

account for the complete uranium source term rather than only the activity of U-238. 

The wind suspension values in the next-to-last column are to be applied when no work is going on 

(e.g., when the guard or AEC is making an inspection visit) and the mechanical resuspension values 

in the last column are to be applied when work is going on.

As for the St. Louis main site case, the assumption is made that the worker breathing rate is 1.2 m3

/hr 

and that the standard work year is 2000 hours. 

The results, assumed to be a constant distribution, are shown in Table A-43. 
Document No. ORAUT-TKBS-0005 Revision No. 02 Effective Date: 06/14/2007 Page 127 of 225

8.6.4

Although the method of NIOSH (2004) appears to be inappropriate for the SLAPS work situation 

because of the outdoor location, it is clear that the method will give a very conservative result and it is 

an approved method. Thus it will be applied as follows in this technical basis document until further 

guidance is provided.

Assumptions for Ingestion Exposure Calculations

1. The air concentration in pCi/m3 from Section 8.6.3 was multiplied by 0.2 to obtain the pCi 

ingested per day.

2. The pCi ingested per day was multiplied by 250 work days per year and by the fraction of the 

year from Section 8.6.1 to obtain the pCi ingested per year.

The results, assumed to be a constant distribution, are shown in Table A-43. 

8.6.5

No radon measurements have been found for SLAPS after 1949, i.e., after the high-radium K-65 

residue was removed. The content of the barium sulfate residue, 4 x 10-9 g per gram of residue, is 

equivalent to about 4000 pCi/g of residue, as given in Section 8.6.2. This can be compared to the 

1.0-1.2 pCi/g of soil that ORNL (1979; 1981) found to be the typical Missouri soil content. Clearly the 

concentration of Ra-226 was elevated over background levels and thus radon emanation would be 

greater than in typical Missouri soils.

Assumptions for and Results of Radon Exposure Calculations

ORNL (1979) provides some measured data by which the radon emanation rate can be related to 

surface soil content of radium. The maximum onsite radon concentration measured was 1.3 x 10-1

pCi/L of air. The average surface soil concentration of Ra-226 for the three highest radon-emanating 

points was found from data in Tables 3 and 13 of ORNL (1979) to be 1.3 pCi/g. This gives a ratio of 

0.1 pCi/L of radon per pCi/g of Ra-226. Thus for the assumed 3960 pCi/g in the residue (see Section 

8.6.2 above), there would be a radon concentration in the air of 396 pCi/L. Although this 

concentration is higher than the concentration used to estimate the radon exposure for the operational 

period, this document uses the larger value to estimate the exposure in later years. Radon exposures 

were not likely larger during this period, but the duration of exposure was more uncertain. 

Although in the outdoors environment the equilibrium factor of the radon daughters is likely to be very 

low, the conservative assumption will be made that the ratio is 0.4. The occupancy factors can be 

taken to be as given above. Thus we can calculate the SLAPS intakes as shown below.

Annual intake, WLM = occupancy factor x equilibrium factor x 0.12 WLM/yr per pCi/l

x radon concentration, pCi/l

The results are 19 WLM/y times the occupancy factor from Section 8.6.1 and are shown in Table A-43

and are assumed to be a constant distribution. 

8.6.6

The dose rates reported in MCW 1949g for a bulldozer cab and at waist or chest height over the aged 

piles were used to calculate the exposures from the piles of barium sulfate residue. Also, the 

bulldozer cab appeared to be appropriate because the work appeared to consist entirely of processes 

such as excavation, dumping from trucks, and inspection from trucks. The maximum reported hourly 

dose rates from fresh raffinate at chest height and in the bulldozer cab were respectively, 33 and 2.7 

mrep beta and 1.3 and 1.0 mrep gamma (MCW 1949g). This indicates that the bulldozer provided a 

Assumptions for and Results of External Exposure Calculations
Document No. ORAUT-TKBS-0005 Revision No. 02 PC-1 Effective Date: 05/25/2009 Page 128 of 225

beta reduction factor of about 10 and a gamma reduction factor of 1.3. The hourly dose rates 

reported for the aged barium cake heap were 10.0 mrep beta and 3 to 10.0 mrep gamma. The beta 

dose rate for dose reconstruction for this period was calculated by dividing 10.0 mrep/hr by the 

bulldozer reduction factor of 10, which resulted in a beta dose rate of 1 mrad/hr. The gamma dose 

rate was calculated by averaging the high and low value for the heap and dividing the result by the 

reduction factor of 1.3, which resulted in a gamma dose rate of 5 mR/h.

Annual beta and gamma dose rates were calculated and both are shown in Table A-44. Once the 

annual gamma doses had been calculated for each worker case from the gamma dose rate, 

considered as ambient measurement data, the doses had to be converted to photon doses to 

individual organs. Conversion factors from Appendix B of NIOSH (2007) were used to do this. It was 

assumed that the dose was from photons of energy between 30 and 250 keV. A geometry of 100% 

ROT was assumed because of the large-pile geometry.

The results are shown in Table A-44, and are assumed to be a constant distribution. 

8.7 USE OF DECONTAMINATION AND POST-DECONTAMINATION EXPOSURE DATA 

FOR ST. LOUIS MAIN SITE DOSE RECONSTRUCTION

As stated above, where urinalysis and film badge data are available for an individual, they should be 

used to determine the internal and external exposure to the individual; the methodology of Sections 

6.0 and 7.0 respectively should be used as guidance for this, as appropriate. Also, for the D&D years 

it should be assumed that a worker had an annual x-ray (see Sections 8.1 and 7.4 above for the 

rationale and for guidance, respectively). It should be remembered that as noted in Section 8.1

above, some AEC and Mallinckrodt personnel film badge records for this period may be found in the 

Weldon Springs site set of records (and not in the Destrehan-St. Louis site set of records) because 

their badges were issued from Weldon Springs.

Where there is missing dose data in records, the dose reconstructor should identify surrogate worker 

data from records available in the dose reconstruction project, based on similarity of job title and 

location. Surrogate data are not provided in this TBD for this time period because of the low number 

of likely claimants requiring it and because records have not been located. 

Where appropriate surrogate data cannot be found for the decontamination period, the data of Table 

A-37 may be used to provide surrogate data on a group basis. A favorable to claimant value should 

be assumed. In particular, the default assumption should be “High” exposure, where that choice is 

available. Where gaps still exist, reference can then be made to the data given in Tables A-40 

through A42. 

For the post-decontamination period, where individual surrogate data cannot be found, as is likely, the 

data of Tables A-40 through A-42 should be used. The default assumption should be “High” 

exposure, where that choice is available.

In both cases, in dose reconstruction using Tables A-40 through A-42, the applicable years of 

employment for each indicated period should be determined and then the number of years should be 

multiplied by the annual value.

With regard to determining the exposure potential category in Tables A-40 through A-42, it should 

most often be possible to determine the degree of exposure of a worker by his job title, e.g., if a 

worker was a secretary, then the worker should be assigned to the “low” exposure potential category. 

In cases where it is not possible to determine the category, then the “high” category should be used.
Document No. ORAUT-TKBS-0005 Revision No. 02 Effective Date: 06/14/2007 Page 129 of 225

8.8 USE OF DECONTAMINATION AND POST-DECONTAMINATION EXPOSURE DATA 

FOR SLAPS DOSE RECONSTRUCTION

As stated earlier, where urinalysis and film badge data are available for an individual, the data should 

be used to determine the internal and external exposure to the individual; the methodology of Section 

and 7.0 should be used as guidance for external dose reconstruction, as appropriate. If a worker was 

badged or given urinalysis during the years 1959-1961, it should be assumed that a worker had an 

annual x-ray (since if the SLAPS workers were involved in the decontamination and demolition 

operations, whether as Mallinckrodt workers or as contractor workers) this would have been the case. 

It should be remembered that as noted in Section 8.1 above, some AEC and Mallinckrodt personnel 

film badge records for this period may be found in the Weldon Springs site set of records (and not in 

the Destrehan-St. Louis site set of records) because their badges were issued from Weldon Springs; 

this may be the case as well for the D&D contractor workers.

Where there is missing dose data in records, the dose reconstructor should identify surrogate worker 

data from records available in the dose reconstruction project, based on similarity of job title and 

location. Surrogate data are not provided for this period because of the low number of likely claimants 

requiring it and because records have not yet been located.

Where appropriate surrogate data cannot be found for the years 1959-1966, as is likely, the data of 

Table A-44 should be used to determine conservative annual external doses. The applicable years of 

employment for each indicated period should be determined and then the number of years should be 

multiplied by the annual value. The default assumption should be “High” exposure, where that choice 

is available.

9.0

All information requiring identification was addressed via references integrated into the reference 

section of this document.

ATTRIBUTIONS AND ANNOTATIONS
Document No. ORAUT-TKBS-0005 Revision No. 02 Effective Date: 06/14/2007 Page 130 of 225

REFERENCES

ACJ 2002

ACJ & Associates (ACJ) and the UK National Radiological Protection Board. Integrated Modules for 

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Applied Nuclear Safety 1991

Applied Nuclear Safety. St Louis Plant Rn-222 Gas Monitoring -- 3rd Quarter 1990. Memorandum to 

T. Byrd from D. Soldan. 28 January 1991. [SRDB Ref ID: 3995]

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Atomic Energy Commission (AEC). An Estimate of Cumulative Multiple Exposures to Radioactive 

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Airborne Contamination. [HASL-]MCW-15; January 1954. [SRDB Ref ID: 20652]

AEC 1954c

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AEC 1956b

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Airborne Contamination. HASL-MCW-27; 12 October 1956. [SRDB Ref ID: 20646] 

AEC 1953

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Occupational Exposure to Airborne Contamination. 22 May 1953. [SRDB Ref ID: 9367] 

AEC 1954d

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Airborne Contamination. [HASL-]MCW-18; February 1954. [SRDB Ref ID: 20657] 

AEC 1954e

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Airborne Contaminants. HASL-MCW-20; November 1954. [SRDB Ref ID: 20639] 

AEC 1956c

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Airborne Contamination. HASL-MCW-25; 21 September 1956. [SRDB Ref ID: 15618] 

AEC 1956d

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Occupational Exposure to Airborne Contamination. HASL-MCW-28; 25 October 1956. [SRDB Ref ID: 

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AEC 1954f

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Airborne Contamination. [HASL-]MCW-16; February 1954. [SRDB Ref ID: 20654]

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Airborne Contamination. HASL-MCW-21; October 1954. [SRDB Ref ID: 15597] 

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AEC 1948c

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AEC 1952b

Atomic Energy Commission (AEC). Mallinckrodt Chemical Works -- Routine Inspection. Memorandum 

from W. B. Harris to files; 26 March 1952. [SRDB Ref ID: 9236, p. 120]

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to F. M. Belmore from B. S. Wolf; 16 March 1948. [SRDB Ref ID: 3926] 

AEC 1947c

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[Dr. Ernest Wollan?]" from "C. B. B..” 5 August 1947. [SRDB Ref ID: 10342] 

AEC 1950d

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Safety Division. 19 October 1950. [SRDB Ref ID: 13632] 

AEC 1949d

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Report by W. E. Kelley. 10 May 1949. [SRDB Ref ID: 6291] 

AEC 1948f

Atomic Energy Commission (AEC). Monthly Status and Progress Report for February 1948, NYOO. 

Report by W. E. Kelley. 5 March 1948. [SRDB Ref ID: 6287] 

AEC 1949e

Atomic Energy Commission (AEC). Monthly Status and Progress Report for February 1949, NYOO. 

Report by W. E. Kelley. 7 April 1949. [SRDB Ref ID: 6291] 

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from S. A. Anonsen of Mallinckrodt; 7 December 1953. [SRDB Ref ID: 3835, p. 12]

MCW 1949e

Mallinckrodt Chemical Works (MCW). Hand Gamma Exposure at K-65 Drum Opening Operation. 

Memorandum to H. E. Thayer from K. J. Caplan. 31 January 1949. [SRDB Ref ID: 11173, p. 17]

MCW 1950d

Mallinckrodt Chemical Works (MCW). Hand Loading of Raw Ore into Secondary Elevator. 

Memorandum to K. J. Caplan from M. G. Mason. 23 February 1950. [SRDB Ref ID: 12368] 

MCW 1956c

Mallinckrodt Chemical Works (MCW). Hand Scooping of QM-2 from the Pots in the Pot Room at Plant 

Six. Report by W. F. Bushman. 29 December 1956. [SRDB Ref ID: 11641] 

MCW 1953d

Mallinckrodt Chemical Works (MCW). Health Conditions in the MgX Area. Memorandum to E. I . Miller 

from M. G. Mason. 23 August 1953. [SRDB Ref ID: 11638, p. 17]

MCW 1948e

Mallinckrodt Chemical Works (MCW). Health Evaluation of the Feinc Platform Area. Letter to J. J. 

Koenig of the Atomic Energy Commission from K. J. Caplan of MCW. 8 October 1948. [SRDB Ref ID: 

12397, p. 27]

MCW 1951b

Mallinckrodt Chemical Works (MCW). Health Group Monthly Report for December 1950. 

Memorandum and report to H. E. Thayer from K. J. Kaplan, 29 January 1951. [SRDB Ref ID: 9362] 

MCW 1955d

Mallinckrodt Chemical Works (MCW). Health Inspection of Plant Six on October 17, 1955. 

Memorandum to E. L. Ball from M. G. Mason. 18 October 1955. [SRDB Ref ID: 12370] 

MCW 1950e

Mallinckrodt Chemical Works (MCW). Health Physics Program for Plant Six and Plant Four. Letter to 

J. J. Koenig of the Atomic Energy Commission from K. J. Caplan of MCW. 14 April 1950. [SRDB Ref 

ID: 10327, p. 2]

MCW 1950f

Mallinckrodt Chemical Works (MCW). Health Physics Survey of the Airport K-65 Storage Area. No 

author listed. 16 January 1950. 

MCW 1949g

Mallinckrodt Chemical Works (MCW). Health Problems at the Airport Properties. Memorandum to H. 

E. Thayer from K. J. Caplan. 11 August 1949.[SRDB Ref ID: 11193]

MCW 1950g

Mallinckrodt Chemical Works (MCW). Health Problems of Special Ore Inventory, February 1, 1950. 

Memorandum to Dr. C. M. Walden from K. J. Caplan. 10 February 1950. [SRDB Ref ID: 12397, p. 

13]
Document No. ORAUT-TKBS-0005 Revision No. 02 Effective Date: 06/14/2007 Page 145 of 225

MCW 1947

Mallinckrodt Chemical Works (MCW). Health Protection for Ore Refinery Employees. Letter to D. 

Duffey of AEC from H. E. Thayer of MCW. 20 March 1947. [SRDB Ref ID: 6513, p. 150]

MCW 1946a

Mallinckrodt Chemical Works (MCW). Health Report for Plant #6. No author. Radiation summaries for 

7/46 and 8/46 (direct radiation). 1946. [SRDB Ref ID: 15575, pp. 19-26]

MCW 1955e

Mallinckrodt Chemical Works (MCW). Interim Report on Film Badge Exposures in the Shotgun 

Laboratory. Memorandum to M. G. Mason from E. M. Wilcox. 13 October 1955. [SRDB Ref ID: 

12361] 

MCW 1950h

Mallinckrodt Chemical Works (MCW). Investigation of Uranium Analysis Results from Barnes 

Hospital. Memorandum to K. J. Caplan from M. G. Mason. 2 February 1950. [SRDB Ref ID: 3967, p. 

12]

MCW 1954a

Mallinckrodt Chemical Works (MCW). K-65 Sampling Lab. Report by (apparently) E. Roddy. Circa 5 

January 1954. [SRDB Ref ID: 11650] 

MCW undated

Mallinckrodt Chemical Works (MCW). Mallinckrodt Radiation Summary APR 46 TO MAR 48. Author 

unknown, date unknown. From electronic file "Mallinckrodt_1946.pdf" in scanned records. [SRDB Ref 

ID: 6586] 

MCW 1955f

Mallinckrodt Chemical Works (MCW). Mallinckrodt Chemical Works Health Program: Policies, Status, 

and Summary. Report by M. G. Mason of Mallinckrodt. 21 February 1955. [SRDB Ref ID: 13180] 

MCW 1955g

Mallinckrodt Chemical Works (MCW). Maximum Emergency Non-Survey Time [for] Cells, Dust 

Collectors, and Tanks. 1 March 1955. [SRDB Ref ID: 12369, p. 7]

MCW 1955h

Mallinckrodt Chemical Works (MCW). Meter Readings (Gamma) at Plant Monitoring Film Badge 

Locations. Memorandum to M. G. Mason from E. M. Wilcox. 15 December 1955. [SRDB Ref ID: 

6513, p. 46]

MCW 1947-1957

Mallinckrodt Chemical Works (MCW). MZ Reports. Periodic reports of radon measurements. 1947-

1957. [SRDB Ref ID: 9357, p. 36]

MCW 1948f

Mallinckrodt Chemical Works (MCW). Mz Study of Centrifuge Bays, Full Drums in Bays. 

Memorandum to K. J. Caplan from R. E. Youtzy. 18 January 1950. [SRDB Ref ID: 12397, p. 5]

MCW 1949h

Mallinckrodt Chemical Works (MCW). New Type Plant Badge -- Plant 4 and 6. Memorandum to J. A. 

Gerk from K. J. Caplan. 8 April 1949. [SRDB Ref ID: 6513, p. 100]
Document No. ORAUT-TKBS-0005 Revision No. 02 Effective Date: 06/14/2007 Page 146 of 225

MCW 1949i

Mallinckrodt Chemical Works (MCW). No subject. Letter to B. S. Wolf of AEC from K. J. Caplan of 

Mallinckrodt Chemical Works. 23 March 1949. [SRDB Ref ID: 3904, p. 23]

MCW 1955i

Mallinckrodt Chemical Works (MCW). No subject. Letter to C. H. Bingham of National Lead Company 

of Ohio from H. A. Schlueter of MCW. 11 February 1955. [SRDB Ref ID: 6513, p. 57]

MCW 1956d

Mallinckrodt Chemical Works (MCW). No subject. Letter to C. H. Walden of National Lead Company 

of Ohio from H. J. Schaeffer of MCW. 16 April 1956. [SRDB Ref ID: 8649] 

MCW 1951d

Mallinckrodt Chemical Works (MCW). No subject. Letter to C. L. Karl of the Atomic Energy 

Commission from H. S. Thayer of MCW. 23 February 1951. [SRDB Ref ID: 15606, p. 6]

MCW 1950i

Mallinckrodt Chemical Works (MCW). No subject. Letter to H. I. Blatz of AEC from K. J. Caplan of 

MCW. 18 January 1950. [SRDB Ref ID: 3976, p. 18]

MCW 1950k

Mallinckrodt Chemical Works (MCW). No subject. Letter to H. I. Blatz of AEC from M. G. Mason of 

MCW. 6 April 1950. [SRDB Ref ID: 9433, p. 63]

MCW 1950l

Mallinckrodt Chemical Works (MCW). No subject. Letter to H. I. Blatz of AEC from M. G. Mason of 

Mallinckrodt. 20 April 1950. [SRDB Ref ID: 3976, p. 9]

MCW 1949j

Mallinckrodt Chemical Works (MCW). No subject. Letter to J. H. Harley of AEC from K. J. Caplan of 

Mallinckrodt. 26 October 1949. [SRDB Ref ID: 3976, p. 28]

MCW 1950m

Mallinckrodt Chemical Works (MCW). No subject. Letter to J. H. Harley of AEC from K. J. Caplan of 

Mallinckrodt. 15 May 1950. [SRDB Ref ID: 3976, p. 3]

MCW 1950n

Mallinckrodt Chemical Works (MCW). No subject. Letter to J. J. Koenig of the Atomic Energy 

Commission from M. G. Mason of MCW. 16 November 1950. [SRDB Ref ID: 11661] 

MCW 1956e

Mallinckrodt Chemical Works (MCW). No subject. Letter to L. Johnson of the US Atomic Energy 

Commission from F. M. Belmore of Mallinckrodt. 18 June 1956. [SRDB Ref ID: 3987, p. 23]

MCW 1949k

Mallinckrodt Chemical Works (MCW). No subject. Letter to M. Eisenbud of AEC from K. J. Caplan of 

Mallinckrodt Chemical Works. 6 May 1949. [SRDB Ref ID: 3904, p. 17]

MCW 1950o

Mallinckrodt Chemical Works (MCW). No subject. Letter to M. Eisenbud of AEC from K. J. Caplan of 

MCW. 13 July 1950. [SRDB Ref ID: 4020, p. 7]
Document No. ORAUT-TKBS-0005 Revision No. 02 Effective Date: 06/14/2007 Page 147 of 225

MCW 1950p

Mallinckrodt Chemical Works (MCW). No subject. Letter to M. Eisenbud of AEC from K. J. Caplan of 

Mallinckrodt. 18 August 1950. [SRDB Ref ID: 4020, p. 12]

MCW 1950q

Mallinckrodt Chemical Works (MCW). No subject. Letter to M. Eisenbud of AEC from K. J. Caplan of 

Mallinckrodt. 4 October 1950. [SRDB Ref ID: 3967, p. 10]

MCW 1948g

Mallinckrodt Chemical Works (MCW). No subject. Letter to R. Hayes of the University of Rochester 

from K. J. Caplan of Mallinckrodt. 5 June 1948. [SRDB Ref ID: 6513, p. 137]

MCW 1946b

Mallinckrodt Chemical Works (MCW). No subject. Letter to the St. Louis Area Engineer, MED, from 

W. F. Bale of MCW. 3 January 1946. [SRDB Ref ID: 7711, p. 3]

MCW 1946c

Mallinckrodt Chemical Works (MCW). No subject. Letter to the St. Louis Area Engineer, MED, from 

W. F. Bale of MCW. 16 January 1946. [SRDB Ref ID: 7711, p. 14]

MCW 1946d 

Mallinckrodt Chemical Works (MCW). No subject. Letter to the St. Louis Area Engineer, MED, from 

W. F. Bale of MCW. 24 January 1946. [SRDB Ref ID: 7711, p. 8]

MCW 1946e

Mallinckrodt Chemical Works (MCW). No subject. Letter to the St. Louis Area Engineer, MED, from 

W. F. Bale of MCW. 28 January 1946. [SRDB Ref ID: 7711, p. 18]

MCW 1946f

Mallinckrodt Chemical Works (MCW). No subject. Letter to the St. Louis Area Engineer, MED, from 

W. F. Bale of MCW. 1 February 1946. [SRDB Ref ID: 7711, p. 21]

MCW 1951e 

Mallinckrodt Chemical Works (MCW). No subject. Letter to W. B. Harris of the Atomic Energy 

Commission from K. J. Caplan of MCW. 19 January 1951.

MCW 1946g

Mallinckrodt Chemical Works (MCW). No subject. Letter to W. D. Fleming of the Manhattan Engineer 

District from S. H. Anonsen of Mallinckrodt. 9 November 1946. [SRDB Ref ID: 9123] 

MCW 1951f

Mallinckrodt Chemical Works (MCW). Badge Reading Reports. Memorandum to K. J. Caplan from E. 

Miller. 5 April 1951. [SRDB Ref ID: 6513, p. 83]

MCW 1950r

Mallinckrodt Chemical Works (MCW). No subject. Memorandum to K. J. Caplan from M. G. Mason. 23 

February 1950. [SRDB Ref ID: 15578, p. 20]

MCW 1956f

Mallinckrodt Chemical Works (MCW). Pickling of Hi-Carbon Scrap Metal. Memorandum to M. G. 

Mason from J. Horner. 23 October 1956. [SRDB Ref ID: 12365, p. 13]
Document No. ORAUT-TKBS-0005 Revision No. 02 Effective Date: 06/14/2007 Page 148 of 225

MCW 1953e

Mallinckrodt Chemical Works (MCW). Plant 4 Recast Operation. Report by J. R. Fields. Circa 

November 1953. [SRDB Ref ID: 12390] 

MCW 1955j

Mallinckrodt Chemical Works (MCW). Plant 6 Cloth Storage Room Radon Concentration. 

Memorandum to M. G. Mason from D. J. Huelster. 14 January 1955. [SRDB Ref ID: 12397, p. 3]

MCW 1954b

Mallinckrodt Chemical Works (MCW). Plant 6 Laboratory Sample Room Radioactive Dust 

Concentrations. Report by E. Roddy. 5 January 1954. [SRDB Ref ID: 11647] 

MCW 1954c

Mallinckrodt Chemical Works (MCW). Plant 6. Report by (apparently) E. Roddy. 11 February 1954. 

[SRDB Ref ID: 11647, p. 8] 

MCW 1954d

Mallinckrodt Chemical Works (MCW). Plant Four Film Badges. Memorandum to M. G. Mason et al. 

from J. R. Messerli. 10 September 1954. [SRDB Ref ID: 6513, p. 73]

MCW 1950s

Mallinckrodt Chemical Works (MCW). Plant Six Dust Study, August 1950. Report; 21 August 1950. 

[SRDB Ref ID: 9347] 

MCW 1955k

Mallinckrodt Chemical Works (MCW)(?). Procedure for Sampling and Determination of Radon. No 

author, no date. Procedure No. C-02-04-01. Circa 1955? [SRDB Ref ID: 3857] 

MCW 1949l

Mallinckrodt Chemical Works (MCW). Production of Uranium Metal for Commercial Uses. Letter to W. 

E. Kelley of the Atomic Energy Commission from J. Fistere of Mallinckrodt; 25 April 1949. [SRDB Ref 

ID: 9205] 

MCW 1956g

Mallinckrodt Chemical Works (MCW). Quarterly Film Badge Summary -- Second Quarter, 1954. 

Memorandum to C. M. Harrington from R. J. Marschner. 17 August 1956. [SRDB Ref ID: 9432, p. 6]

MCW 1951g

Mallinckrodt Chemical Works (MCW). Radiation Exposure of Feinc Cloth Operators. Memorandum to 

H. E. Thayer from K. J. Caplan. 5 March 1951. [SRDB Ref ID: 12391, p. 3]

MCW 1950t

Mallinckrodt Chemical Works (MCW). Radiation Exposure of MCW Plant Six Employees. Letter to C. 

L. Karl of the Atomic Energy Commission from H. E. Thayer of MCW. 5 May 1950. [SRDB Ref ID: 

10327, p. 24]

MCW 1956h

Mallinckrodt Chemical Works (MCW). Radioactive Airborne Dust Study of Sampling of Feed Materials 

for Oil Analysis. Report by D. Darr of Mallinckrodt. 4 December 1956. [SRDB Ref ID: 12364] 
Document No. ORAUT-TKBS-0005 Revision No. 02 Effective Date: 06/14/2007 Page 149 of 225

MCW 1958

Mallinckrodt Chemical Works (MCW). Radioactive Contamination Survey, Mallinckrodt Chemical 

Works -- Uranium Division, Plant 4 Site. Report of a survey by the MCW Health and Safety 

Department; 26 November 1958. [SRDB Ref ID: 4023, p. 115]

MCW 1959

Mallinckrodt Chemical Works (MCW). Radioactive Contamination Survey, Mallinckrodt Chemical 

Works -- Uranium Division, Destrehan Plant. Report of a survey by the MCW Health and Safety 

Department; 24 August 1959. [SRDB Ref ID: 4023] 

MCW 1955l

Mallinckrodt Chemical Works (MCW). Radioactive Dust Concentrations, U-Con Grinding in the 

Shotgun Laboratory. Report by J. M. Horner of Mallinckrodt. 9 November 1955. [SRDB Ref ID: 

12365, p. 3]

MCW 1957

Mallinckrodt Chemical Works (MCW). Radon and Gamma Levels in the Plant 6 Cloth Storage Room. 

Report by D. G. Darr. 18 January 1957. [SRDB Ref ID: 12397, p. 35]

MCW 1948h

Mallinckrodt Chemical Works (MCW). Radon and Radiation Problem of the "New" Ore Cars. 

Memorandum to S. H. Aronsen from K. J. Caplan. 14 October 1948. [SRDB Ref ID: 12397, p. 18]

MCW 1948i

Mallinckrodt Chemical Works (MCW). Radon Concentrations in K-65 Centrifuge Platform. 

Memorandum to H. E. Thayer from K. J. Caplan. 31 December 1948. [SRDB Ref ID: 12397, p. 6]

MCW 1949m

Mallinckrodt Chemical Works (MCW). Radon Concentrations in Ore Room. Memorandum to E. Miller 

from K. J. Caplan. 1 July 1949. [SRDB Ref ID: 12369, p. 3]

MCW 1949n

Mallinckrodt Chemical Works (MCW). Radon in General Yard Area of Plant Six. Memorandum to H. 

E. Thayer from K. J. Caplan. 11 August 1949. [SRDB Ref ID: 10848] 

MCW 1949o

Mallinckrodt Chemical Works (MCW). Radon in Warehouse Ore Drying Oven. Memorandum to H. E. 

Thayer from K. J. Caplan. 4 May 1949. [SRDB Ref ID: 12397, p. 10]

MCW 1951h

Mallinckrodt Chemical Works (MCW). Radon Survey Downwind from East End of Manufacturing 

Building Early Monday after Week-end Shutdown.” Memorandum to M. G. Mason from R. E. Youtzy. 

15 January 1951. [SRDB Ref ID: 3962] 

MCW 1950u

Mallinckrodt Chemical Works (MCW). Recheck of Dust Survey in the Ore Room and Ore Room 

Addition. Memorandum to M. G. Mason from R. E Youtzy. 29 September 1950. [SRDB Ref ID: 

10324] 
Document No. ORAUT-TKBS-0005 Revision No. 02 Effective Date: 06/14/2007 Page 150 of 225

MCW 1954e

Mallinckrodt Chemical Works (MCW). Report on Medical and Industrial Hygiene Talks Between 

Representatives of MCW and NYOO AEC Health Departments. Memo to C. D. Harrington from M. G. 

Mason, 24 February 1954. [SRDB Ref ID: 9362, p. 24]

MCW 1950v

Mallinckrodt Chemical Works (MCW). Revised Warehouse Work-Time Schedule. Memorandum to J. 

Lang from K. J. Caplan. 21 June 1950. [SRDB Ref ID: 11640, p. 5]

MCW 1949p

Mallinckrodt Chemical Works (MCW). Revised Work Time Schedule for Warehouse Operations. 

Memorandum to J. Lang from K. J. Caplan. 13 July 1949. [SRDB Ref ID: 11640, p. 3]

MCW 1955m

Mallinckrodt Chemical Works (MCW). Sampling of Feed Materials. Memorandum to M. G. Mason et 

al. from W. J. Shelley. 13 October 1955.

MCW 1948j

Mallinckrodt Chemical Works (MCW). Scale House and Scale House Sample Room. Memorandum to 

H. E. Thayer from K. J. Caplan. 12 June 1948. [SRDB Ref ID: 12369, p. 8]

MCW 1955n

Mallinckrodt Chemical Works (MCW). Scalehouse Sample Room Study [of] Occupational Exposure to 

Radioactive Dust. Report by E. L. Roddy. 23 June 1955. [SRDB Ref ID: 12371] 

MCW 1956i

Mallinckrodt Chemical Works (MCW). Summary of the Mallinckrodt Chemical Works Film Badge 

Program. Memorandum to J. W. Miller from K E. Brandner. May 1956. [SRDB Ref ID: 6513, p. 35]

MCW 1951i

Mallinckrodt Chemical Works (MCW). Techniques of Breath Radon Sampling. Letter to W. B. Harris of 

the Atomic Energy Commission from K. J. Caplan of MCW. 8 January 1951.

MCW 1950w

Mallinckrodt Chemical Works (MCW). Termination Physical Examinations for MCW Employees on 

AEC Project. Letter to C. L. Karl of the Atomic Energy Commission from H. E. Thayer of MCW. 17 

May 1950. [SRDB Ref ID: 10327, p. 14]

MCW 1950x

Mallinckrodt Chemical Works (MCW). The Role of Primary Beta Radiation in Area Radiation Dose 

Rates, Plant Six Digest Area. Report by R. F. Nagel of (presumably) Mallinckrodt. 17 November 1950. 

[SRDB Ref ID: 9362, p. 195]

MCW 1949q

Mallinckrodt Chemical Works (MCW). Use of Film Badge during Off-Plant Work. Memorandum to Mr. 

Anonsen et al. from K. J. Caplan. 21 April 1949. [SRDB Ref ID: 6513, p. 99]

MCW 1949r

Mallinckrodt Chemical Works (MCW). Use of Rubber Gaskets on K-65 Drums to Prevent Escape of 

Radon. Memorandum to H. E. Thayer from D. J. Huelster. 8 March 1949. [SRDB Ref ID: 12397, p. 

25]
Document No. ORAUT-TKBS-0005 Revision No. 02 Effective Date: 06/14/2007 Page 151 of 225

MCW 1948k

Mallinckrodt Chemical Works (MCW). Weekly Badge Summary. Memorandum to H. E. Thayer et al. 

from K. J. Caplan. 29 June 1948. [SRDB Ref ID: 6513, p. 130]

MCW 1956j

Mallinckrodt Chemical Works (MCW). Weekly Report -- Radiation Exposures. Memorandum to J. W.

MIller from K. E. Brandner. 17 August 1956. [SRDB Ref ID: 15571, p. 11]

MCW 1949s

Mallinckrodt Chemical Works (MCW). Work Time Permits for Warehouse Operations. Memorandum 

to J. Lang from K. J. Caplan. 28 February 1949. [SRDB Ref ID: 11640, p. 2]

MCW 1955o

Mallinckrodt Chemical Works (MCW). Investigation of Dust Problem in Cage Area of Plant 6E. 

Memorandum to M. G. Mason from J. W. Miller. 7 July 1955. [SRDB Ref ID: 15616, p. 4]

MCW 1955p

Mallinckrodt Chemical Works (MCW). Plant 6E Derby Brushing Station. Report by W. Linenweber. 5 

December 1955. [SRDB Ref ID: 15616, p. 27]

MCW 1955q

Mallinckrodt Chemical Works (MCW). Plant 6E -- Grinding Station Radioactive Dust Concentrations. 

Report by J. M. Horner of MCW. 16 September 1955. [SRDB Ref ID: 15616, p. 84]

MCW 1955r

Mallinckrodt Chemical Works (MCW). Study of Plant 6 Pot Room and Orange Packing Station. Report 

by E. Roddy of MCW. 20 August 1955. [SRDB Ref ID: 15616, p. 46]

MCW 1948l

Mallinckrodt Chemical Works (MCW). Health Physics Survey of the Airport K-65 Storage Shed and 

Area. No author listed. 25 February 1948.

MCW 1950y

Mallinckrodt Chemical Works (MCW). Mallinckrodt Chemical Works Plant Four Dust Study. 

Memorandum report to K. H. Caplan from D. J. Huelster. 1 March 1950. [SRDB Ref ID: 9340] 

MCW 1956k

Mallinckrodt Chemical Works (MCW). Radiation Exposures. Memorandum to M. G. Mason from K. E. 

Brandner. 21 December 1956.

MCW 1955s

Mallinckrodt Chemical Works (MCW). Radiation in the Upper YM-5 Furnace Area, Plant Six-E. Report 

by E. L. Roddy. 11 November 1955. [SRDB Ref ID: 17244] 

MCW 1956l

Mallinckrodt Chemical Works (MCW). Radiation Study of the Shotgun Laboratory Operations, April 

27, 1956. Report by E. L. Roddy. 16 May 1956. [SRDB Ref ID: 15581] 

MCW 1956t

Mallinckrodt Chemical Works (MCW). Visual Inspection of Plant 6E Operations and Facilities. 

Memorandum to J. W. Miller from W. Linenweber. 29 September 1955. [SRDB Ref ID: 15616, p. 

201]
Document No. ORAUT-TKBS-0005 Revision No. 02 Effective Date: 06/14/2007 Page 152 of 225

MCW 1956m

Mallinckrodt Chemical Works (MCW). Weekly Report -- Radiation Exposures. Memorandum to J. W. 

MIller from K. E. Brandner. 14 September 1956. [SRDB Ref ID: 15571, p. 8]

MCW 1956n

Mallinckrodt Chemical Works (MCW). Weekly Report -- Radiation Exposures. Memorandum to J. W. 

MIller from K. E. Brandner. 24 August 1956.

Mallinckrodt 1994

Mallinckrodt, Inc. (Mallinckrodt). Columbium-Tantalum Plant Characterization Plan. Plan for the St. 

Louis, Missouri plant; 19 January 1994. [SRDB Ref ID: 3840] 

MED 1946a

Manhattan (Engineer) District (MED), (Army) Corps of Engineers, Office of the District Engineer. 

Description of Uranium Producing Processes. Memorandum from Major W. E. Gates to the Area 

Engineer (Madison Square Area); 17 October 1946. [SRDB Ref ID: 9442, p. 16]

MED 1949a

Manhattan Engineer District (MED), Division of Technical Advisors. Uranium Metal Production. 

Memorandum from F. G. Stroke to Manhattan Engineer District Files; 4 March 1949. [SRDB Ref ID: 

9442. p. 11]

MED 1949b

Manhattan Engineer District (MED), Division of Technical Advisors. Uranium Technology. 

Memorandum from F. G. Stroke to Manhattan Engineer District Files; 3 March 1949. [SRDB Ref ID: 

9442, p. 12]

MED 1948

Manhattan Engineer District (MED). Explosion at Mallinckrodt Chemical Works. Memorandum to K. D. 

Nichols from G. W. Beeler; 13 June 1948. [SRDB Ref ID: 4024, p. 15]

MED 1943a

Manhattan Engineer District (?) (MED). Fire in Oxide Plant (Buildings 51 and 52). Memorandum from 

J. H. McKinley; 14 July 1943. [SRDB Ref ID: 4024, p. 16]

MED 1943b

Manhattan Engineer District (MED). Fire in Rubbish Truck -- Plant 4, July 3, 1943. Memorandum to 

the St. Louis Area Engineer from J. R. Ruhoff of the St. Louis(?) Area Engineer. 27 July 1943. [SRDB 

Ref ID: 4024, p. 33]

MED 1945a

Manhattan Engineer District (MED). Installation of Dust Control System in the Reduction Step, 

Mallinckrodt Chemical Works. Memorandum to the New York Area Engineer from E. M. Velten of the 

St. Louis Area Engineer. 14 May 1945. [SRDB Ref ID: 3996, p. 19]

MED 1944a

Manhattan Engineer District (MED). No subject. Letter to E. V. Thayer of Mallinckrodt Chemical Works 

from R. A. Tybout of the Corps of Engineers. 15 July 1944. [SRDB Ref ID: 9136, p. 5]

MED 1944b

Manhattan Engineer District (MED). No subject. Letter to H. E. Thayer of Mallinckrodt Chemical 

Works from J. L. Ferry of the Corps of Engineers. 4 March 1944. [SRDB Ref ID: 5823, p. 10]
Document No. ORAUT-TKBS-0005 Revision No. 02 Effective Date: 06/14/2007 Page 153 of 225

MED 1944c

Manhattan Engineer District (MED). No subject. Letter to J. A Kyger of Mallinckrodt Chemical Works 

from J. L. Ferry of the Corps of Engineers. 31 March 1944. [SRDB Ref ID: 5888] 

MED 1944d

Manhattan Engineer District (MED). No subject. Letter to H. E. Thayer of Mallinckrodt Chemical 

Works from J. L. Ferry of the Corps of Engineers. 11 May 1944. [SRDB Ref ID: 9134] 

MED 1944e

Manhattan Engineer District (MED). No subject. Letter to H. E. Thayer of Mallinckrodt Chemical 

Works from J. L. Ferry of the Corps of Engineers. 9 June 1944. [SRDB Ref ID: 9144, p. 3]

MED 1944f

Manhattan Engineer District (MED). No subject. Letter to H. E. Thayer of Mallinckrodt Chemical 

Works from R. A. Tybout of MED. 26 June 1944. [SRDB Ref ID: 9144, p. 6]

MED 1944g

Manhattan Engineer District (MED). No subject. Letter to H. E. Thayer of Mallinckrodt Chemical 

Works from R. A. Tybout of the Corps of Engineers. 6 July 1944. [SRDB Ref ID: 9130] 

MED 1944h

Manhattan Engineer District (MED). No subject. Letter to H. E. Thayer of Mallinckrodt Chemical 

Works from R. A. Tybout of MED. 14 July 1944. [SRDB Ref ID: 9144, p. 13]

MED 1944i

Manhattan Engineer District (MED). No subject. Letter to H. E. Thayer of Mallinckrodt Chemical 

Works from R. A. Tybout of MED. 21 October 1944. [SRDB Ref ID: 9136, p. 11]

MED 1945b

Manhattan Engineer District (MED). No subject. Letter to H. E. Thayer of Mallinckrodt Chemical 

Works from R. A. Tybout of MED. 12 January 1945. [SRDB Ref ID: 9144, p. 10]

MED 1945c 

Manhattan Engineer District (MED). No Subject. Letter to H. E. Thayer of Mallinckrodt Chemical 

Works from R. A. Tybout. 19 January 1945. [SRDB Ref ID: 9136, p. 22]

MED 1945d

Manhattan Engineer District (MED). No subject. Letter to H. E. Thayer of Mallinckrodt Chemical 

Works from R. A. Tybout of MED. 26 February 1945. [SRDB Ref ID: 9127] 

MED 1946b

Manhattan Engineer District (MED). No subject. Letter to H. E. Thayer of Mallinckrodt Chemical 

Works from W. F. Bale of MED. 6 February 1946. [SRDB Ref ID: 9124] 

MED 1946c

Manhattan Engineer District (MED). No subject. Letter to H. E. Thayer of Mallinckrodt Chemical 

Works from W. F. Bale of MED. 18 November 1946. [SRDB Ref ID: 7709] 

MED 1943c

Manhattan Engineer District (MED). No subject. Letter to J. A. Kyger of Mallinckrodt Chemical Works 

from H. L. Friedell of the Corps of Engineers. 16 June 1943. [SRDB Ref ID: 5885, p. 5]
Document No. ORAUT-TKBS-0005 Revision No. 02 Effective Date: 06/14/2007 Page 154 of 225

MED 1945e

Manhattan Engineer District (MED). No subject. Letter to S. H. Anonsen of Mallinckrodt Chemical 

Works from B. J. Mears of MED. 9 July 1945. [SRDB Ref ID: 10837] 

MED 1945f

Manhattan Engineer District (MED). No subject. Letter to S. H. Anonsen of Mallinckrodt Chemical 

Works from B. J. Mears of MED. 1 August 1945. [SRDB Ref ID: 9144, p. 15]

MED 1946d

Manhattan Engineer District (MED). No subject. Letter to the St. Louis Area Engineer from J. W. 

Howland of the Medical Corps. 9 September 1946. [SRDB Ref ID: 7711, p. 25]

MED 1946e

Manhattan Engineer District (MED). No subject. Letter to the St. Louis Area Engineer from J. W. 

Howland of the Medical Corps. 14 November 1946. [SRDB Ref ID: 7711, p. 29]

MED 1945g

Manhattan Engineer District (MED). No subject. Memorandum to the St. Louis Area Engineer from 

W. F. Bale. 15 November 1945. [SRDB Ref ID: 9144, p. 43]

MED 1945h

Manhattan Engineer District (MED). No subject. Memorandum to the St. Louis Area Engineer from 

W. F. Bale. 20 November 1945. [SRDB Ref ID: 9144, p. 50]

MED 1945i

Manhattan Engineer District (MED). No subject. Memorandum to the St. Louis Area Engineer from 

W. F. Bale. 26 November 1945. [SRDB Ref ID: 9144, p. 59

MED 1945j

Manhattan Engineer District (MED). No subject. Memorandum to the St. Louis Area Engineer from 

W. F. Bale. 6 December 1945. [SRDB Ref ID: 9144, p. 71]

MED 1945k

Manhattan Engineer District (MED). No subject. Memorandum to the St. Louis Area Engineer from 

W. F. Bale. 11 December 1945. [SRDB Ref ID: 9144, p. 65]

MED 1945L

Manhattan Engineer District (MED). No subject. Memorandum to the St. Louis Area Engineer from 

W. F. Bale. 21 December 1945. [SRDB Ref ID: 9146, p. 122]

MED 1945m

Manhattan Engineer District (MED). No subject. Memorandum to the St. Louis Area Engineer from 

W. F. Bale. 26 December 1945. [SRDB Ref ID: 9144, p. 79]

MED 1943d

Manhattan Engineer District (MED). No title. Air sampling data sheets series for measurements taken 

8 April - 16 June 1943. 1943. [SRDB Ref ID: 10838] 

MED 1944j

Manhattan Engineer District (MED). Progress Report of the Special Materials Division of the Medical 

Section. Memorandum to S. L. Warren from J. L. Ferry. 3 January 1944. [SRDB Ref ID 14490]
Document No. ORAUT-TKBS-0005 Revision No. 02 Effective Date: 06/14/2007 Page 155 of 225

MED 1944k

Manhattan Engineer District (MED). Report of Visit to Mallinckrodt Chemical Company, 17 and 18 July 

1944. Memorandum to the Madison Square Area Engineer from R. A. Tybout. 22 July 1944. [SRDB 

Ref ID: 5885] 

MED 1944l

Manhattan Engineer District (MED). Security Survey Report, Mallinckrodt Chemical Works, St. Louis, 

Missouri. Memorandum to The District Engineer, Oak Ridge Office, from E. M. Velten of the St. Louis 

Area Office; 21 December 1944. [SRDB Ref ID: 4024] 

MED 1942

Manhattan Engineer District(?) (MED). Supplemental Report To Complete Preliminary Survey of 

Plants Engaged in Production of Materials. Report by H. L. Friedell of the Corps of Engineers (MED). 

23 December 1942. [SRDB Ref ID: 3996] 

MED 1944m

Manhattan Engineer District (MED). Visit to Mallinckrodt Chemical Company -- 20 September 1944. 

Memorandum to The Madison Square Area Engineer from R. A. Tybout. 22 September 1944. [SRDB 

Ref ID: 9144, p 27]

MED 1944n

Manhattan Engineer District (MED). Visit to Mallinckrodt Chemical Works -- 6 September 1944. 

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5885, p. 9]

MED 1943e

Manhattan Engineer District (MED). Radiation Hazards -- Mallinckrodt Plant. Memorandum to H. L. 

Friedell of the Chicago Area Engineer from T. T. Crenshaw, 22 January 1943. [SRDB Ref ID: 3996, 

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MED 1944o

Manhattan Engineer District (MED). Safeguarding of Metal. Memorandum to the New York Area 

Engineer, Attn C. Hadlock, from F. Huke, 27 January 1944. [SRDB Ref ID: 3996, p. 9]

MED 1944p

Manhattan Engineer District (MED). Security Survey Report, Mallinckrodt Chemical Works. Form 

submitted by the Chicago Branch Intelligence Office; 22 November 1944. [SRDB Ref ID: 4024, p. 19]

MED 1945n

Manhattan Engineer District (MED). Shipment Security Survey at Mallinckrodt Chemical Works. 

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the Intelligence and Security Division; 15 February 1945. [SRDB Ref ID: 9161] 

MED undated a

Manhattan Engineer District (?) (MED). Supplemental Report To Complete Preliminary Survey of 

Plants Engaged in Production of Materials. Undated (1942-1944 vintage). [SRDB Ref ID: 3996] 

MED undated b

Manhattan Engineer District (?) (MED).Supplementary Report on the Explosion in the Ether Building 

on 4 May 1946 at the New Process Plant, St. Louis Area. Memorandum from S. H. Brown of MED(?). 

Undated. [SRDB Ref ID: 4024, p. 34]
Document No. ORAUT-TKBS-0005 Revision No. 02 Effective Date: 06/14/2007 Page 156 of 225

MED 1946f

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MED 1945o

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ATTACHMENT A TABLES

LIST OF TABLES

TABLE TITLE

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

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

PAGE

165

A-2 Summary chronology of operations at the Mallinckrodt St. Louis Downtown 

Site and at the St. Louis Airport Storage Site...................................................................... 168

A-3 Principal changes made in sites, processes, and equipment .............................................. 169

A-4 Types and quantities of material produced in association with Mallinckrodt 

uranium refining and related operations.............................................................................. 170

A-5 Functional and process keywords and codes ..................................................................... 175

A-6 Thorium and daughter content of the AM-7 residue ............................................................ 180

A-7 Airborne uranium particle size in process areas.................................................................. 180

A-8 Uranium dust concentrations, Plants 4 and 6, in alpha dpm/m3 .......................................... 181

A-9 Uranium dust concentrations, Plants 6E and 7, in alpha dpm/m3 ........................................ 182

A-10 Early air sample data, short-term sample readings in alpha dpm/m3................................... 182

A-11 Airborne uranium dust concentrations in Plant 4 Areas, 1948: AEC versus 

Mallinckrodt measured data................................................................................................ 183

A-12 Plant 4 measured daily weighted average exposure concentrations ................................... 184

A-13 Plant 6 measured daily weighted average exposure concentrations ................................... 185
Document No. ORAUT-TKBS-0005 Revision No. 02 Effective Date: 06/14/2007 Page 163 of 225

A-14 Plant 6E measured daily weighted average exposure concentrations................................. 187

A-15 Plant 7 measured daily weighted average exposure concentrations ................................... 188

A-16 Plant 7E measured daily weighted average exposure concentrations................................. 188

A-17 Average and highest airborne dust concentrations in the laundry ....................................... 189

A-18 Job titles and classifications, with geometry factors ............................................................ 190

A-19 Uranium dust daily weighted average exposure levels, Plant 4........................................... 194

A-20 Uranium dust daily weighted average exposure levels, Plant 6........................................... 195

A-21 Uranium dust daily weighted average exposure levels, Plant 6E ........................................ 196

A-22 Radioactive dust daily weighted average exposure levels, Plant 7 (including 

the Slag Separation Plant) and Plant 7E............................................................................. 197

A-23 Early radon data, 1945 ....................................................................................................... 198

A-24 Measured radon concentrations, 1947-1949, in units of 1 × 10-10 Ci/L ................................ 199

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

of 1 × 10-10 Ci/L................................................................................................................... 201

A-26 RESERVED........................................................................................................................ 202

A-27 Contamination levels and associated dose rates from work clothing................................... 202

A-28 RESERVED........................................................................................................................ 202

A-29A RESERVED........................................................................................................................ 202

A-29B RESERVED........................................................................................................................ 202

A-29C RESERVED........................................................................................................................ 202

A-29D RESERVED........................................................................................................................ 202

A-29E RESERVED........................................................................................................................ 202

A-29F RESERVED........................................................................................................................ 202

A-30 RESERVED........................................................................................................................ 202

A-31A RESERVED........................................................................................................................ 202

A-31B RESERVED........................................................................................................................ 202

A-31C RESERVED........................................................................................................................ 202

A-31D RESERVED........................................................................................................................ 202

A-31E RESERVED........................................................................................................................ 202

A-31F RESERVED........................................................................................................................ 202

A-32 Exposure rates from K-65 residue and Q-11 ore, in mR/hr ................................................. 203

A-33 Measured dose rates in various work areas........................................................................ 204

A-34 Neutron dose rates and doses from the alpha-neutron reaction sources and 

from a RaBe source............................................................................................................ 210

A-35 External gamma dose rates from processing AM-7 residue to concentrate 

thorium ............................................................................................................................... 211

A-36 Reserved............................................................................................................................ 211

A-37 Monitoring experience from the decontamination/demolition period, 1960-

1961 ................................................................................................................................... 212

A-38 Summary of post-decontamination exposure data as measured by two 

survey groups..................................................................................................................... 213

A-39 Measured data used to produce source terms for decontamination and postdecontamination exposure calculations .............................................................................. 214

A-40 Annual internal exposures during the decontamination and postdecontamination periods..................................................................................................... 215

A-41 Annual external photon organ doses, decontamination and postdecontamination periods..................................................................................................... 216

A-42 Annual external electron organ doses during the decontamination and postdecontamination periods..................................................................................................... 218

A-43 Annual dust inhalation, ingestion, and radon daughter intakes during the 

SLAPS postoperations period............................................................................................. 218
Document No. ORAUT-TKBS-0005 Revision No. 02 Effective Date: 06/14/2007 Page 164 of 225

A-44 Annual external beta doses, gamma doses and gamma organ doses during 

the SLAPS postoperations period....................................................................................... 219
Document No. ORAUT-TKBS-0005 Revision No. 02 Effective Date: 06/14/2007 Page 165 of 225

Table A-1. Plants and buildings used at the St. Louis Downtown Site for MED/AEC 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 For MED/AEC work, used only from 1944-1946

1 A General plant maintenance (for all of Mallinckrodt)

1 P Engineering Department (for all of Mallinckrodt)

1 X Locker room

1 Z Company headquarters; administrative offices Administrative headquarters for MED work until Plant 6 began operating in 1946

2 38B Personnel change house FUSRAP (2003a) says 38B, but ORAU (Mason 1977) says 38A

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, mechanical repairs, 

tank storage, other

Ore concentrates, UO3 product; tanks of stored process liquids; UF4 experiments performed in 

the “sulfur burner room”

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 This was a special restricted area because of the 100-mg RaBe source used in the neutron 

absorption testing 

4 400 Production of UF4, U metal; slag handling, production control laboratory, 

changehouse; 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 Maintenance, metal storage, 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, NH3, 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 (air conditioning), 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 Pitchblende

6 105 Main refinery: the “ Ether House” Extraction

6 106 Nitric acid recovery

6 106A Nitric acid recovery

6 107 Nitric acid recovery, tank farm pump house
Document No. ORAUT-TKBS-0005 Revision No. 02 Effective Date: 06/14/2007 Page 166 of 225

Plant Building Process, function, or area Notes

6 108 Shotgun sample preparation lab Superseded by lab in Building 102? [“Old shotgun lab”]

6 109 Acid unloading station Includes 109, 109A, 109B

6 110 Main warehouse for bagged goods, storage of pitchblende ore and ore 

concentrates, 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 Automotive repair

6 111 Sample preparation and LeDoux Lab, north end; 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 (MCW and AEC offices), Health Laboratory, dispensary; 

maintenance shop, instrument shop, maintenance stores

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

6 123 Ammonia & 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).

6E 116B Electrical substation

6E 116C Slag (recycle) building (Magnesium fluoride) Slag handling, grinding, and packaging. 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.
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Plant Building Process, function, or area Notes

7 700 Warehouse, safety office, electrical and carpenter shops, temporary 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 and packaging 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.

The information in this table is drawn from MED 1944p; FUSRAP 2003a ; MCW 1958; MCW 1959; Mason (1977); and ORNL 1981. 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.

Illegible areas in the references are indicated by dots (.....).

Buildings that remained after the general demolishment of 1959-1961 are shown in Table A-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.
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Table A-2. Summary chronology of operations at the Mallinckrodt St. Louis Downtown Site (SLDS) 

and at the St. Louis Airport Storage Site (SLAPS). 

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 or early 1946, 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.

1946-1947 AEC acquires SLAPS in 1946 and Mallinckrodt residues begin to be sent there for storage in about 1947.

1947-1951 Decontamination of Plants 1 and 2, with unrestricted release to Mallinckrodt in 1951.

1948-1949 K-65 residue is brought back from SLAPS for reprocessing. After about 1949, no more K-65 residue is 

sent to SLAPS, only low-radium residues.

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. Processing of high-grade pitchblende ore ceased and 

concentrates became the principal feed material.

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 1958, all regular site operations ceased, with some Plant 7 activities continuing until July 1958. 

Postprocessing and shutdown-related activities continued into 1958.

1959-1962 The St. Louis site is decontaminated in 1959-1961, with unrestricted release of the site to Mallinckrodt in 

about 1962.

1962 AEC issues a license to a private entity, which then took over control of the SLAPS site.

1976 and 1978 Oak Ridge National Laboratory surveys the St. Louis Airport Site (SLAPS) for DOE.

1977 Oak Ridge National Laboratory surveys the St. Louis site (SLDS) for DOE.

1986, 1990 Applied Nuclear Safety does some surveys of the St. Louis site (SLDS) for DOE.

1995 FUSRAP takes over the St. Louis site (SLDS) for remediation.
Document No. ORAUT-TKBS-0005 Revision No. 02 Effective Date: 06/14/2007 Page 169 of 225

Table A-3. Principal changes made in sites, processes, and equipment. 

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

1946-7 SLAPS Began operation of SLAPS Storage of pitchblende residues

1947 SLAPS Construction of K-65 storage shed Control external dose, radon exposures

1948 4, 6 Start of formal health program Track and reduce exposures

1948 6 Began to use ventilated ore railcars Reduce radon exposures

1948 6 Began reprocessing the K-65 residue Recover residual uranium; concerns regarding 

corroding drums

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 (e.g., 

adding hoods to the tray loading, tray dumping, 

milling, blending, and packing operations)

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

1949 6, 

SLAPS

Instituted limits on time spent handling ore and K65 in railcars and at SLAPSControl external and radon exposure to individuals

1950 4 Installed plantwide vacuum cleaning system Enhance dust control and reduce airbornes

1950 6 Installed equipment decontamination room, 

respirator decontamination facility

Control contamination

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

1952-3 6 New Ledoux sampling labs with better ventilation Lower airborne dust levels

1952-4 6 Use of pitchblende ore began to be phased out

and soluble feeds were used increasingly

Use other available feeds

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? 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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Table A-4. Types and quantities of material produced in association with Mallinckrodt uranium refining and related operations. 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 ore 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, AQ-4) and 

leach products 

(tailings 

concentrates)

Most of the pitchblende processed by 

Mallinckrodt was obtained as a concentrate 

from the Belgian Congo in 1944 (AEC 1967), 

processed at Middlesex (AEC 1951b), and 

shipped to St. Louis in 55-gal drums (AEC 

1967).

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 1949b); 

0.185 ppm equilibrium Ra for Q-11 (60%) ores 

(AEC 1949b); 0.1 Ci/ton (total?), up to 70% U, 

average U concentration >25%, about 100 mg/ton 

ore for 25% U (Eisenbud 1975). Tailings: 30-50% 

U3O8 (AEC 1951b)

3400 tons U produced during wartime (through 

1944?) (Eisenbud 1975); AEC (1951b) implies that 

Mallinckrodt was using only Congo ore, Congo ore 

tailings concentrates, and soda salt around 1951 

Canadian 

pitchblende ores 

(Great Bear Lake, 

Port Hope)

 Ores at perhaps 10% U (Eisenbud 1975); at 25-

30% in 1951 (AEC 1951b) 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 1967)

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).

Ore concentrates 

(MGX, etc.)

Special types of ore concentrates, e.g., the 

MGX, a magnesium uranate prepared in Africa 

from low-grade ore tailings (AEC 1955a)

Sodium diuranate

(soda salt)

Na2U2O7. Packed in fiber containers (MED 

1945a)

Vitro converted U ores to sodium diuranate (AEC 

1951b; DOE 1997); some apparently also came 

from Anaconda, Durango, and Fernald (AEC 

1956b). Fiber containers weighed about 75 lbs 

each when full (MED 1945a).
Document No. ORAUT-TKBS-0005 Revision No. 02 Effective Date: 06/14/2007 Page 171 of 225

Material Process or Operation Content and Form Notes Amount

REFINING PRODUCTS

UNH (uranium 

nitrate hexahydrate)

An intermediate product in the digestionextraction process: UO2(NO3)·6 H2OSmall quantities produced for research purposes 

as needed (AEC 1951b)

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 with 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). Plant 6 produced a 

monthly average of 21 tons in 1950 for R&D work 

on the continuous UO3-to-UF4 production process 

(apart from what went into the normal UO3-UO2

production process) and 146 tons in 1Q 1951 as 

the principal product for the new UO3-to-UF4

process (AEC 1951b).

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 (MED 1946a). 2.5-gal 

fiber containers weighed about 75 lbs each when 

full (MED 1945a). There was 349 lbs on a 4-tray 

charge (MED 1944o).

By DOE 1942 Mallinckrodt was producing a ton a 

day (DOE 1996). Mallinckrodt used 32,000 lbs 

weekly (MED 1944o). 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 (MED 1949a). 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 NovDec 1944, and 13,000 lbs/week after that (MED 

1946a). The design capacity of Plant 6 was 200 

tons/month, using ore @ 65% U3O8 (AEC 1951b). 

The monthly averages from Plant 6 were 94 tons in 

1947, 246 tons in 1948, 204 tons in 1949, 207 tons 

in 1950 and 133 tons in 1Q 1951; large-scale 

production of UO2 (as a separate product) ended in 

March 1951 (AEC 1951b). Plant 7’s design 

capacity was 154 tons/month (i.e., could be 

produced if desired as a tapoff from the continuous 

UO3-to-UF4 process) (AEC 1951b).

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 1945a) or 5-gal containers for 

transfer to Plant 4 or 6E or another site (AEC 

1949b).

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 1944o).

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 1944o). 

In 1945, 8000 lbs/week was being sent to Harshaw 

and 20000 lbs/week to Iowa State (MED 1945a). 

Average monthly production in Plant 4 was 76 tons

through Dec 1946, 131 tons in 1947, 147 tons in 

1948, 151 tons in 1949, 169 tons in 1950, and 189 

tons in 1Q 1951 (AEC 1951b). Plant 7’s design 

capacity was 225 tons/month originally, 375 after 

expansion (AEC 1951b) (unclear if expansion took 

place).
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Material Process or Operation Content and Form Notes Amount

U metal in derby 

form

Reduction with magnesium in furnace to U 

metal (slag + derby); slag chipped off to leave 

derby

In 1944-1949, there was 135-140 lbs of UF4 per 

bomb, along with 55 lbs of liner (MED 1944o, AEC 

1949g). 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 1944o).

In 1944, 7500 lbs of biscuit was produced weekly 

(MED 1944o). 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 shipped out every two weeks to Hanford in a 

carload lot of about 30,000 lbs, from a weekly 

production of 13,000-15,000 lbs (MED 1945a).

Average monthly production in Plant 4 was 33 tons 

through December 1946, 84 tons in 1947, 102 tons 

in 1948, 100 tons in 1949, 169 tons in 1950, 189 in 

1Q 1951; average monthly production in Plant 6E 

was 83 tons in 4Q 1950 and 162 tons in 3Q 1951; 

the latter was said to be AEC’s entire requirement 

for metal at the time (AEC 1951b). Plant 6E’s 

design capacity was 150 tons/month originally, 265 

tons/month after the expansion of 3Q 1951 (AEC 

1951b).

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 1956a) 

In late 1955 or early 1956, this replaced the derbybillet operation, except for occasional experimental 

derby production in Plant 4, per AEC (1956a). But 

AEC (1956c) reported that in mid-1956 (all?) billet 

recasting was being done in 6E, using new 

graphite molds. A dingot weighed about 3300 lbs 

(AEC 1956a).

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 1944o) 

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), presumably as the metal. 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.
Document No. ORAUT-TKBS-0005 Revision No. 02 Effective Date: 06/14/2007 Page 173 of 225

Material Process or Operation Content and Form Notes Amount

U slag (recast) Recast slag was recovered as residue from the 

recast furnace.

As the uranium melted, the slag floated to the top 

of the crucible and beta-emitting Th-234 and Pa234 (UX1 and UX2) sublimed and condensed on 

the underside of the furnace lid; recast furnace 

slag was thus highly concentrated in UX1-UX2

(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 1949h)

 

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.

 

RESIDUES AND OTHER WASTES

Pitchblende raffinate 

(AM-7)

“Airport cake”; produced as part of the 

pitchblende digestion process (extraction 

raffinate); see also Sperry cake

0.2% U (AEC 1949b); 0.15% U (AEC 1959); 29 

ppm Th-232, 3.8 ppm Th-230 (11.6 isot %) (Figgins 

and Kirby 1962). Pitchblende residues:Th-232/Th230 ~ 8. In AEC (1960), sampling in June 1953 

showed highest sample at .0038% Th and 0.14% U

33000 lbs/day (AEC 1949b); 74K tons total, 113 

tons U stored through at least 1965 (AEC 1960; 

AEC 1964; ORNL 1979) 

De-thoriated pitchblende raffinate 

(AM-9)

Residue after processing AM-7 for ionium (Th230)0.12%U (AEC 1959). Carnotite residues: Th232/Th-230 ~ 15-20 (AEC 1949b) 

Domestic ore 

raffinate (AM-10)

“Airport cake”; produced as part of the nonpitchblende digestion processCarnotite residues: Th-232/Th-230 ~ 15-20 32.5K tons total, 48 tons U stored through at least

1965 (AEC 1959, AEC 1960; ORNL 1979)

Pb-Ra precipitate

(K-65, gangue lead 

cake) 

“Lead cake”; a Ra-bearing residue produced as 

part of the pitchblende digestion process

(resulting from addition of sulfuric acid); sent to 

Lake Ontario or the airport for storage

750-900 mg Ra/ton (AEC 1949b); 750 mg/ton Ra 

and 0.2% U (AEC 1949b); up to 300 Ra mg/ton 

(Eisenbud 1975)

8000-12000 lbs/day (AEC 1949b)

Ba sulfate cake

(AJ-4)

Produced as part of the digestion process

(resulting from addition of barium carbonate)

0.28% U (AEC 1959); 1 mg Ra/ton, 0.1% U (AEC 

1949b); .1% U, 1E-9 g Ra per g U (AEC 1949m) 

6800 lbs/day total AJ-4 (AEC 1949b); stored: 1.5K 

tons total unleached (wit 22 tons U), 8.7K tons total 

leached (with 7 tons U) (FUSRAP undated a, AEC 

1960), i.e., 10.2K tons total (with 29 tons U (AEC 

1959; ORNL 1979) 
Document No. ORAUT-TKBS-0005 Revision No. 02 Effective Date: 06/14/2007 Page 174 of 225

Material Process or Operation Content and Form Notes Amount

Sperry cake Produced in Sperry press from aqueous tails 

from the ether extraction step; some later sent 

to Mound for processing to extract Pa; 

apparently a subset of AM-7

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 0.33% U (AEC 1959). C-6 and V-10 were stored in 

a total of 2400 drums (FUSRAP undated)

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 1949b)

U-containing sands, 

precipitates (V-10)

Captured from the Japanese C-6 and V-10 were stored in a total of 2400 drums 

(FUSRAP undated)

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

<2% U (AEC 1959); 1.6% U (AEC 1959). 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).

7800 tons, 122 tons U (AEC 1959); 4000 tons, 49 

tons U I 1964 (ORNL 1979) 

Interim Residue 

Plant Tailings

(C-701, D-701) 

Resulted from scalping the U content from the 

Mg fluoride slag from 1955 on; 701 apparently 

refers to Bldg 701 (AEC 1956d) 

2.1% U (AEC 1959) 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 ether extraction step; filtered and treated 

with lime; precipitate become airport/Sperry 

cake, filtrate disposed of as liquid effluent

See the text for process details. See the keywords table (Table A-5) for other code numbers and terms.
Document No. ORAUT-TKBS-0005 Revision No. 02 Effective Date: 06/14/2007 Page 175 of 225

Table A-5. Functional and process keywords and codes. Plant Keyword or code Note

2 MED hazard index number for soluble uranium material

4 MED hazard index number for insoluble uranium material

6 MED hazard index number for nitrous fumes

14 MED code for Building 51

16 MED hazard index number for hydrogen fluoride

32 MED hazard index number for handling of metal

38 MED code for the laboratory area in Plant 2

51 See “Building 51” below

52 MED code for Plant 4

128 MED hazard index number for hydrogen fluoride gas

162, 172, 182 Black oxide (U3O8); could appear as “Chemical 162,” etc.

256 MED hazard index number for radium

264, 272 Orange oxide (UO3); could appear as “Chemical 264,” etc.

 306 Brown oxide, i.e., UO2; could appear as “Chemical 306” 

512 MED hazard index number for radon

1024 MED hazard index number for solvents

2048 MED hazard index number for radiation

4-bagger Type of dust collector

A In film badge records, denotes “absent”; elsewhere, denotes non-specification grade black 

oxide (U3O8)

6 Acid Acid addition: in ore digestion (nitric) or Pb-Ra precipitation (sulfuric)

7 Adams A type of polishing filter used in Plant 7

6 AEC Atomic Energy Commission

Airport SLAPS, the former airport site later used for waste storage by the AEC

6 AJ-4 Barium sulfate cake produced from pitchblende ore (ore-to-UO3)

AJ-7 Another barium cake

6 AM-7 Raffinate cake produced by treating pitchblende ore

6 AM8 Unclear what this indicates

4,6,6E,7 Area M Area mechanic

6 Assist LO Assisting the lead operator

6 AQ-4 Pitchblende ore

4 Ballard Vertical turret lathe (manufactured by Ballard) used to scalp the dingot

6 Barium Barium salt addition

Beets(?) See Egg

4, 6E Billet Final form of uranium metal, produced by recasting from derbies

6 Bird centrifuge Solid-bowl centrifuge (manufactured by Bird) used to separate liquids and solids

Black oxide U3O8

4, 6 Blender Apparently ore in Plant 6; UF4 + Mg in Plants 4, 6E; slag in Plant 6E

4, 6E BM Bottom man

6 Boildown A step between digestion and ether extraction

4 Bomb Container for Mg-UF4 in the metal reduction process

4, 6E Bomb step The UF4-to-derby process

4, 6E Bottom Lower (furnace or F machine)

4, 6E Bottom man Bottom man in the YM-5 production process -- would physically enter the bottom part of the 

furnace in Plant 4

4 Box Crucible holder (external assembly)

6E Breakout Removal of the derbies from the bombs

Brown Brown oxide, i.e., UO2

Brown monster Apparently the Rockwell furnace(s) (UO3-to-UO2)

Building 51 The 50 series of buildings at Plant 2 (i.e., 50, 51, 51A, 52, 52A, and 55)

Building K Apparently Building K1E in Plant 1

6E Burnout, burn Crucible burnout (heating)

C On film badge records, denotes no badge issued (“clean”?)

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., centrifuging, packing) related to use of the carbonate leaching 

process to recover uranium from Feinc filter cake (K-65); also dust collection related to this

C-27 Derby chippings (residue)

C-55 As “Chemical K-35,” etc., denotes black oxide (U3O8) (may be scrap)

4, 6E Cage Scrap holding area

4, 6E Cage man Handled billet and other metal scrap, degreased billets

6E Cap Valved top put on the Mg-UF4 drum for jolting

Cats High-grade ore
Document No. ORAUT-TKBS-0005 Revision No. 02 Effective Date: 06/14/2007 Page 176 of 225

Plant Keyword or code Note

CB On film badge records, denotes contaminated badge and film

6 CEN, Cent, Cntr Centrifuge. Also abbreviated CNF?

Chemical Start of some code names; was followed by letters or numbers

4 Chipping Removal of the slag from the derbies using a manual or power hammer

Classifier Equipment used in the Ore Room apparently to sort ore pieces by size

--- Cleanup Generic for area or item cleanup: see associated keyword (e.g., TA-7)

C-liner Dolomite slag left after the bomb is opened and the derby removed

C-Special Liner slag left on bomb and possible also derby (had to be chipped off)

6 Cloth Cloth (actual or metal) used to filter solids from liquid streams

Cl Up Cleanup, i.e., cleanup of process area

6 CM Cloth man?

Cocoa Brown oxide, i.e., UO2

4, 6 Continuous furnace, Cont 

furn

Furnace for processing UO3 to UF4 in a continuous run (in Plant 4, experimental process in 

Pilot Plant)

Croppings Chips or pieces taken off derbies, billets, and dingots

6E Crucible, Cruc Crucible or crucible assembly (6 should probably be 6E)

Crusher Probably the UO2 crusher (mill)

CS Refinery feed material that was an impure calcium uranate

 C-Special Magnesium fluoride slag formed on top of the derby

CX Calcium uranate leach product from the processing of ore tailings at African mills; sent to 

Mallinckrodt for laboratory research and pilot-scale recovery of the uranium

D Radium (the element)

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)

D-701 Residues collected in the slag grinding area. shipped to slag recovery

6 DA Dissociated (cracked) ammonia or Digest area

6,6E, 7 Decontamination Generic

 Deheading Same as Delidding

Delidding Removing the lid of a drum, usually of ore

4, 6E Derby Crude uranium metal form produced from UF4

6 Digest, DIG, Dig Digestion process (ore dissolution by acid), digestion area, digestors

4 Dingot “Direct ingot,” an extra-large ingot produced instead of derbies and billets

DR ???

DX Probably decontamination work

 E Designates the Mallinckrodt plants in St. Louis

Egg 1.9-kg egg-shaped piece of U metal produced for external assays

6 Ether, Ether House, EH Ether storage, handling for U solvent extraction process

4, 6 Extra, Extra Man Extra (floating) chemical operator for a process or area (4, 6); possibly also “extraction” (6)

6 Extraction Extraction of uranium in the ether extraction process

F Generic feed material designation for types without a code name

4, 6E Fce Furnace (metal-making areas)

FD On film badge records, denotes factory-damaged film

6 FE, Fe Feinc (FE Inc) filter (equipment or area)

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

FF On film badge records, denotes film fogged, torn, or open (in badge?)

6 FH Feed hopper?

6 Filter press Press type of filter for separating solids from liquids and leaving a dewatered cake

Fin Slag or magnesium burr on billet, bomb shell lid, etc. that had to be chipped off

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

 Gangue, gangue lead cake The Pb-Ra cake also known as K-65

GL In film badge records, denotes “gone through the laundry” in context of film condition; denotes 

work with gangue lead cake in context of worker area

6 GLC Gangue lead cake, i.e., the Pb-Ra cake also known as K-65

Green Green salt (UF4)

6 Gulping Vacuum-sucking UO3 out of the pot it was produced in into drums

GWT-5 Uranium metal eggs (test samples)

GY-3 U3O8 (black oxide), having a minimum assay of 97% U3O8

 H-20 ??? (H2O, i.e., water?)
Document No. ORAUT-TKBS-0005 Revision No. 02 Effective Date: 06/14/2007 Page 177 of 225

Plant Keyword or code Note

7 Hapmann, Hapman 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 dedicated additive handling and storage building or area

4, 6E Ingot Same as billet

6 Instrument, Instru, Instr Instrument Shop

Ionium Thorium-230

Iron maiden C-Special, while still attached to derby

4, 6E Jolter, jolting The jolter-filling machine in UF4-to-U metal production, also the operator; the operation

Juice Orange oxide, i.e., UO3

K-35, K-82 As “Chemical K-35,” etc., denotes black oxide (U3O8)

6 K-65 K-65 (Ra-Pb) residue (gangue lead cake) from the (pitchblende) ore-to-UO3 process

6E KB-2 Derby form of uranium metal

4 Label Apparently labeling product containers

6 Laboratory, Lab, LAB Any of several laboratories (Analytical, Shotgun, Ledoux, etc.)

6 Laundry For both contaminated and uncontaminated clothing; Mallinckrodt had own laundry from ~ 

1948 on

LB In film badge records, denotes film that was opened (in badge?)

 Ld Loading (e.g., TA-7)

LF-9 Brown oxide (UO2)

LFD In film badge records, denotes film lost in the darkroom (torn open)

LFP In film badge records, denotes film lost in plant

LG Lead gangue? See GLC

6 Ledoux (LeDoux) 

Laboratory

The raffinate and uranium assay laboratory in Plant 6

6 Liquor Extracted liquid concentrate, usually after removal of undesirable materials as solids

LL Ledoux Laboratory

LO Lead operator

4, 6 Loading, Load, Ld Generic: see associated process keyword, e.g., TA-7

Location E The Mallinckrodt St. Louis site (SLDS)

 M Powdered magnesium metal, work in the metal-making (derby and billet) areas, or 

maintenance (depending on context)

6 M(+number), M-(+number) (1) Designates tank(s) used in processing. (2) Designates a type of development work, e.g., 

M-1 for slag liner work, M-2 for mold improvement, and M-7 for crucible testing

M-20 Step or tank during which barium sulfate was added

ME Refers to processing of AM-7 raffinate to concentrated liquid Th form (Minor Elements)

MFG, Mfg Manufacturing: generic for Plant 6 and Plant 4

6 MGX, MgX, MGX Process Refinery feed material that was an impure magnesium uranate from the Belgian Congo

4 Mag Room Magnesium [Storage] Room

1, 2 Main Plant Plants 1 and 2

6, 6E Maintenance, M, MNT Maintenance, presumably process maintenance

ME Refers to reprocessing of raffinate to a feed concentrate

7 MFL MgF2 liner (for use in dingot bombs); loosely, also the associated slag used (recycled) in the 

liner

6 Mikro Dust collector (manufactured by Mikro)

6 Mill See Rod mill

Mtns Maintenance

4 Muffle Batch electric (muffle) furnace used to reduce UO3 to UO2

 Mx (note lower case) Uranium (generic for the element?)

MX (note upper case) See MGX

My Radium; also, used to indicate a breath radon measurement (since this measurement was 

used to infer radium body burden)

Mz, MZ Radon-222

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 In the re-extraction process, the OK liquor contained the uranium and the NOK liquor 

contained the dross (after filtration, it became raffinate).

6 Ntns Maintenance (Mtns)?

4 Office, OFF Generic for office

OH Hydroxide (e.g., to neutralize excess HF and recover uranium)

6 OK In the re-extraction process, the OK liquor was the uranium-containing liquid sent to the 

boildown tanks

6 Oliver Filter press (manufactured by Oliver)
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Plant Keyword or code Note

 Orange, orange juice Orange oxide, i.e., UO3

6 Ore, Ore Room, OR Ore processing, storage before use; also handling of residues

OZ Electrically fused dolomite

4, 6, 7 Pack, packing, packaging Appears to be generic for packaging

6 Pangborn Type of dust collector (K-65/Ledoux Sample Lab; ore grinding room; C-3)

6 Peterson Type of filter, often in a hood; at least some was used to filter NOK liquor to produce the raffinate

PH Power House

6E Pickling Soaking derbies in acid to remove surface impurities (scale, oxides)

All Pilot Plant, PP Plant 4’s name (for metallurgical research and developmental UF4 furnace work), after metal 

production moved to 6E; Plants 1 & 2, around the time Plant 6 started up; a slag recovery pilot 

plant in 6E; or an unspecified pilot plant at 6 (must be taken from context)

Pk Packing (packaging) of a product form

All Pl Plant (e.g., Pl. 7 is Plant 7)

Plant 10 The later name of the Plant 4 area (after Plant 4 was dismantled)

2 Plant 51 Building 51 operations in Plant 2

6E PLO Process Lead Operator?

6 Pot, Pot Room Denitration pot room (producing UO3)

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

Project (+ number) Project 89 was Plants 1 and 2; 90, Plant 4 green salt operations; 91, Plant 4 derby production; 

92, Plant 4 recast operations

 Q-11 Pitchblende ore, other high-grade ore

QM-2 Orange oxide (UO3)

6 Raffinate, Raff Residues (mainly the cakes)

6 Railcar Railroad car used to transport ore, K-65 residue, and sometimes scrap

6 Rapping Knocking a dust collector (usually the Pangborn) to loosen collected material for removal

4,7 Reactor Reaction vessel/heater in which UO3 is converted to UO2 with cracked ammonia

4 Recast, Recast Furnace Where derbies were recast into billets

4 Receiving & Shipping Generic? The main Receiving & Shipping area was in Plant 6, but there may have been plantspecific areas 

6 Recovery, REC Nitric oxide recovery (6); slag recovery (7); may have indicated other types of recovery at other 

plants

Regulated Refers to radiologically controlled areas, e.g., the regulated locker room and the general locker 

room 

Regulus Same as Derby

6 Reverter (riverter?) Unclear. Appears to be a type of pump or processor used in the recycling of dust collector 

material

7 RMF Recycle magnesium fluoride (from bomb liner residue)

6E Rockwell Rockwell furnace (UF4-Mg reduction to derby metal)

6 Rod mill Piece of equipment used to grind ore

Rolls Billets

6 Rover Extra pair of hands or swing man? 

S As “Chemical S,” denotes sodium salt (sodium diuranate, Na2U2O7)); in urinalysis records, 

denoted sulfate (work)

4 Salt bath Bath in which a dingot is put to heat it

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

SLAPS Sand (1) cake residue or (2) uranium-containing sands captured from the Japanese

6E Saw Sawing off a portion of the metal billet as a sample

6 SC-5 Uranyl nitrate hexahydrate (UNH)

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

4 Semi-works Term for a facility within a plant, on the scale of or slightly larger than a pilot plant (specifically, 

the dingot operation in Plant 4)

6 Skip hoist Hoist used to raise ore drums, dump them into raw ore bin feeding (ore) rod mill

6E Slag Byproduct from the Mg-UF4 reduction step (mostly MgF2)

7 Slag Processing Plant Same as Slag Separation Plant

[4] Slag Shed Storage of the chipped-off bits of slag?

SLAPS, SLAPSS St. Louis Airport Storage Site

6E Slye Dust collector (manufactured by Slye)

6 Shotgun Laboratory Shotgun sample preparation laboratory

6 Shotgun Samp Prep Shotgun sample preparation (lab)
Document No. ORAUT-TKBS-0005 Revision No. 02 Effective Date: 06/14/2007 Page 179 of 225

Plant Keyword or code Note

6 Soda salt Na2U2O7 (sodium diuranate, a feed material); perhaps other soluble feed material

SP In film badge records, denotes film spoiled in the darkroom (development) process

6 Sperry Filter press (manufactured by Sperry); also, the cake produced (same as or subset of AM-7)

 SR Storeroom 

SS Stainless steel presses, i.e., the plate and frame filters

SSP Usually refers to stainless steel (filter) presses; may indicate Shotgun Sample Prep Lab

ST Uranium metal scrap

6 Storeroom Generic; the main storeroom was in Plant 6

6 Stripper Unclear; probably piece of equipment rather than operator name

All T Generic code term for uranium, including as metal (T metal), in compounds (e.g., TO2), and as 

a daughter indication (e.g., TX1)

6 T-3 Typo for C-3?

4 TA-7 Green salt (UF4)

TA-7R UF4 produced from UO2 in the experimental continuous reactor

Talcum UF4

6 Thawhouse Building where ore was thawed in winter before undrumming

6 Thief Sampling device that opens at a desired depth into the material

6 THP THP ether?

6 Thumper Device to “thump” feed drum to dislodge material during digestor loading

4, 6E TM Top man, Top-off man, etc., in the YM-5 production process

4, 6E Top Refers to the upper or top recasting furnace or the top F machine

4, 6E Top man Top man, Top-off man, etc., in the YM-5 production process

4, 6E, 7 Topping Refers to the operation of adding material to ensure proper fill

Tubealloy Uranium metal (generic term also used roughly to indicate any uranium)

Tumbler, tumbling area UO2 breakup (crushing) area

U In film badge records, denotes “unidentifiable” in context of film condition; elsewhere, usually 

denotes uranium

6, 7 U-CON, U-con Uranium concentrate slurry produced in the Slag Separation (Processing) Plant (7) from reject 

MgF2 material and sent to Plant 6 as a feed material

6 UNH Uranyl nitrate hexahydrate (UO2(NO3)2 ∙ 6 H2O)

Unld Unloading (e.g., TA-7)

6E Utility Utility man or “floater” who filled in as needed

UX1, UX2 Old names for Th-234 and Pa-234m respectively

V In film badge records, denotes “on vacation”

V-4 A form of soda salt used as feed material

W In film badge records, denotes “wet from unknown reason”

All Warehouse, WH Generic: various warehouses were used, with the main one in Plant 6 and later in Plant 7

4 Weigh, check weigh Generic: feed and products were weighed at various points

6E Wilfey Shaker table (manufactured by Wilfey) used to separate high-U slurry from low-U material

X Generic code term for uranium, including in compounds (e.g., XO2)

Yard Open areas outside buildings and guard stations; used to store drums, unload railcars; even 

used postoperations for decontamination

YB-1 UO2 produced from UO3 in the experimental continuous reactor

 Yellow cake The generic form of uranium ore, somewhat preprocessed

4 YM-5 Billet form of uranium metal
Document No. ORAUT-TKBS-0005 Revision No. 02 Effective Date: 06/14/2007 Page 180 of 225

Table A-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 A-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 0.00101 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%

Rn-222 3.82 days 1.58 x 10-1 0.65%

Po-218 3.05 minutes 1.58 x 10-1 0.65%

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%

Ac-228 6.13 hours 8.46 x 10-4 84%

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

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.

Table A-7. Airborne uranium particle size in process areas. 

Area Location or source type Method*

Average 

concentration, 

mg U/m3

Mass median 

particle size, 

microns GSD

Mallinckrodt UO2 Production Area (Rochester 1948b)

UO3-to-UO2 conversion <1 foot of the blow pipe F 2.14 1.2 2.34

Tumbling Vicinity of furnace, tumbler tray rack F .11-.46 1.4-1.45 2.5-3.2

C .64-.93

Further from furnace, tumbler rack F .069-095

C .41-.53

Near skip hoist F 2.8-4.2 1.5-1.9 2.45-2.6

C .59-.69

4 feet from skip hoist F 0.75

C 3.5

Back of the skip hoist F 2.6 2.65

Fiberpack packaging F, C .060-2.2 .72-3.48

2.96-

3.03

Warehouse Closed containers of UO2 F .017-.033 1.1 2.3

General (Non-Mallinckrodt) Foundry (Sanders 1975, Table 2)

Foundry Depleted U --- --- 2.8 ± 2.7** ---

Enriched U --- --- 3.3 ± 2.2** ---

Enriched U --- --- 2.1 ± 2.0** ---

Machining Depleted U, milling dry --- --- 3.0 ± 2.3 ---

Extruding Depleted U --- --- 3.2 ± 2.7 ---

* This refers to the sampling method. F: filter paper (with air sampling pump) . C: cascade impactor

** 88% of these particles were from particles less than 7 μm.
Document No. ORAUT-TKB

S-0005 Revision No. 02 Effective Date: 06/14/2007 Page 181 of 225

Table A-8. Uranium dust concentrations, Plants 4 and 6, in alpha dpm/m3 (Mason 1958a, Tables 1, 2).

Year

Plant 4

Plant 6

Warehousing Ore grinding

UO3 production UO2 production Misc

UO2 

Production

UF4

Production

KB-2 

Production

YM-5 

Production Feed digest Milling Pot room Packaging Load Unload Packaging

1943 2,100 2,380 1,190 2,520 --- --- --- --- --- --- --- --- ---

1944 2,100 2,380 1,190 2,520 --- --- --- --- --- --- --- --- ---

1945 2,100 2,380 1,190 2,520 --- --- --- --- --- --- --- --- ---

1946 2,100 2,380 1,190 2,520 210 13,300 420 12,600 7,770 1,400 5,320 3,150 11,270

1947 2,100 2,380 1,190 2,520 210 13,650 420 12,600 7,770 1,400 5,320 3,150 11,270

1948 2,100 2,380 1,190 2,520 210 13,650 420 12,600 7,770 1,400 5,320 3,150 11,270

1949 420 280 280 770 210 13,650 420 12,600 7,770 1,400 5,320 3,150 11,270

--- --- --- --- 70 350 70 0 4,200 700 1,400 700 350

1950 280 140 210 770 70 350 70 * 770 700 1,400 700 350

--- --- --- --- 28 350 70 --- 350 700 700 350 350

1951 280 210 --- --- 35 350 140 --- 140 70 420 210 350

1952 280 210 --- --- 35 210 350 --- 210 140 420 210 350

1953 --- --- --- --- 63 140 49 --- 210 140 420 210 350

1954 --- --- --- --- 21 140 42 --- 140 140 ** ** **

1955 --- --- --- --- 21 * 56 --- 140 140 --- --- ---

1956 --- --- --- --- 21 --- 28 --- 210 280 --- --- ---

1957 --- --- --- --- 21 --- 56 --- 210 70 --- --- ---

The first two Plant 4 processes were supposedly transferred to Plant 7 in 1952-3, the last two to Plant 6E in 1950-1. Compare to information for Plants 6E and Plant 7 (Table A-9). 

Prior to 1946, the Plant 6 operations were done in Building 51 of Plant 2; all work was transferred to Weldon Spring in March 1957. Note that the figures above reflect chronologically 

the significant improvements in ventilation and dust control made in 1949.

* Discontinued

** Transferred to Plant 7 in October 1952
Document No. ORAUT-TKBS-0005 Revision No. 02 Effective Date: 06/14/2007 Page 182 of 225

Table A-9. Uranium dust concentrations, Plants 6E and 7, in alpha dpm/m3 (Mason 1958a, Tables 3, 

4).

Plant 6E Plant 7 

KB-2 production YM-5 production All operations

Year Average High

Source of 

high reading Average High

Source of 

high reading Average High

Source of 

high reading

1950 7 21 Charging 70 126 Crucible assembly --- ---

1951 21 56 Residue 77 126 Top furnace --- ---

1952 28 70 Residue 84 161 Burnout 35 112 UO2 dumping

1953 21 84 Charging 35 49 Top furnace 28 119 Furnace operation, 

TA-7 packaging

1954 35 210 Residue 63 119 Bottom furnace 35 490 Sampling, cleanup

1955 112 280 Residue 42 105 Bottom furnace 21 77 UO2 dumping

1956 28 112 Capping 35 42 Bottom furnace 28 56 UO2 dumping

1957 28 105 Burnout 49 84 Bottom furnace 21 56 UO2 dumping

1958 56 147 Breakout 84 147 Bottom furnace 35 98 TA-7 packaging

Plant 6: KB-2 is the derby form and YM-5 the billet form of uranium metal.

Plant 7: Operations were moved to Weldon Spring in 1958.

Table A-10. Early air sample data, short-term sample readings in alpha dpm/m3.

Location and operation

1943 1944 1945 1946

Value

No. of 

Samples Value GSD

No. of 

Samples Value GSD

No. of 

Samples Value GSD

No. of 

Samples

Plant 1

Bldg K, Feinc filtration 60 2

Bldg 25, Laboratory 7 2

Plant 2 (Bldgs 51/51A)

Oxide being added to Tank 1A 153 1

Pot Room, general area 301 1 742 2 511 1

Pot Room, aisle between pots 700 1 

Pot Room, (un)loading UO3 2,000 1 2,380 1

Pot Room, chipping, scooping cake 11,300 1

Milling UO3 2,800 3.03 3 3,220 1

Sifting room general area 304 2

Sifting room, sifting 22,50 1

UO2 being weighed 70 1

UO2 unloading 15,540 2 

UO2 packaging 65,800 1

Plant 2 (Shotgun Lab, Bldg 55)

Shotgun residue evaporation 62 2

Shotgun residue grinding 97 1.54 4

Plant 4

UF4 general area 49 3

Unloading UF4 6,160 2 91 1 563 2.36 4

Milling UF4 6,062 2 74 2 630 2

Mixing/blending/jolting UF4, Mg 1,960 1 84 2

Bomb area general 5,320 1.22 3 

Bomb loading 2,100 1

Bomb unloading 1,154 2.79 4 324 2

Biscuit chipping 4,484 1.27 3 42 1 

Biscuit sealing 1,540 1

Slag barreling 210 1.10 4

Crucible burnout 2,100 1

Recast furnace general area 976 2

Recast furnace unloading 700 1

Recast furnace brushing 980 1

Ore milling 720 2.70 4

Ore handling 1,230 2

Plant 6

Ore storage – area 246 4.08 9 

Boxcar 504 4.10 3

The sampling data used to produce the table above are from MED (1943d; 1944b; 1944c; 1944d; 1944e; 1944g; 

1944h; 1945d; 1946d; 1946e; MED 1945o) and MCW (1946b; 1946c; 1946d; 1946e; 1946f). The figure given in the 

Value column is the median, if there were three or more samples; the mean if there were two samples; and the sole 

value if there was one sample.
Document No. ORAUT-TKBS-0005 Revision No. 02 Effective Date: 06/14/2007 Page 183 of 225

Table A-11. Airborne uranium dust concentrations in Plant 4 Areas, 1948: AEC versus Mallinckrodt 

measured data (AEC 1949b, 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 3.34E+03 2.08E+03

Operator B 47.7 30.7 2.39E+03 1.54E+03 3.34E+03 2.15E+03

Furnace Tending 

Operator A 7.5 5.7 3.75E+02 2.85E+02 5.46E+02 3.99E+02

Operator B 9.1 6.1 4.55E+02 3.05E+02 6.37E+02 4.27E+02

TA-7 Unloading

Operator A

- Manual 186 66.8 9.30E+03 3.34E+03 1.30E+04 4.68E+03

- Semi-mechanized 16.4 8.20E+02 1.29E+03

Operator B

- Manual 186 57.2 9.30E+03 2.86E+03 1.30E+04 4.00E+03

- Semi-mechanized 20.7 1.04E+03 1.45E+03

TA-7 Mixing & Packing

Operator A 63 13.1 3.15E+03 6.55E+02 4.41E+03 9.17E+02

Operator B 57 24.6 2.85E+03 1.23E+03 3.99E+03 1.72E+03

Bomb Charging 51 42.4 2.55E+03 2.12E+03 3.57E+03 2.97E+03

Topping 10.4 32.4 5.20E+02 1.62E+03 7.28E+02 2.27E+03

Jolting 51 7.5 2.55E+03 3.75E+02 3.57E+03 5.25E+02

Charge Firing 13.3 13.8 6.65E+02 6.90E+02 9.31E+02 9.66E+02

Derby Unloading

Operator A 5 9.1 2.50E+02 4.55E+02 3.50E+02 6.37E+02

Operator B 5 18.7 2.50E+02 9.35E+02 3.50E+02 1.31E+03

Chipping 26.3 11.5 1.32E+03 5.75E+02 1.84E+03 8.05E+02

Slag Handling 1.6 2.2 8.00E+01 1.10E+02 1.12E+02 1.54E+02

Top Furnace Tending

Operator A 61 36.5 3.05E+03 1.83E+03 4.27E+03 2.56E+03

Operator B 61 23.2 3.05E+03 1.16E+03 4.27E+03 1.62E+03

Bottom Furnace Tending 73 59 3.65E+03 2.95E+03 5.11E+03 4.13E+03

Sawing 15.8 5.4 7.90E+02 2.70E+02 1.11E+03 3.78E+02

Cage Handling 2.7 52 1.35E+02 2.60E+03 1.89E+02 3.64E+03

Office 0.6 4.1 3.00E+01 2.05E+02 4.20E+01 2.87E+02

Shipping & Receiving 1.6 6.8 8.00E+01 3.40E+02 1.26E+02 4.76E+02

Mechanics 5 10.7 2.50E+02 5.35E+02 3.50E+02 7.49E+02

Carpenter 2 4.6 1.00E+02 2.30E+02 1.40E+02 3.22E+02

Porters 0.8 2.9 4.00E+01 1.45E+02 5.60E+01 2.03E+02

Guards 0.4 1.4 2.00E+01 7.00E+01 2.80E+01 9.80E+01

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 1949b). 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

. 

Columns 6 and 7 represent the values obtain by multiplying Columns 4 and 5 respectively by 70 dpm/m3 (assuming 70 

dpm/m3 is equivalent to 50 ug/m3

). 
Document No. ORAUT-TKBS-0005 Revision No. 02 Effective Date: 06/14/2007 Page 184 of 225

Table A-12. Plant 4 measured daily weighted average exposure concentrations. Weighted average concentration, alpha dpm/m3

 

Occupation 5/1956 11/1953 3/1953 6/1950 10/1949 9/1948 5/1948 AEC 1949b AEC 1951a

Magnesium operator 35 70

Lime blender 35 70

Slag man 70 105 210 140

Cage man (handler) 190 3640 189 2940

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) 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

Data from the surveys of 6/50, 10/49, 9/48, and 5/48 is from AEC 1950c; data from the surveys of 3/53 and 11/53 is from AEC1954b; and 

data from the survey of 5/56 is from AEC 1956a. Other data are from the references given in the column headings. 
Document No. ORAUT-TKBS-0005 Revision No. 02 Effective Date: 06/14/2007 Page 185 of 225

Table A-13. Plant 6 measured daily weighted average exposure concentrations. Weighted average concentration, alpha dpm/m3

Occupation

May 

1956

May 

1954

Oct 

1953

Jan 

1953

Jan 

1952

Aug 

1950 1949

Oct-Nov 

1948

May 

1948

Digest area lead operator 6 60 36 62 140 84 686 280

Digest operator 7.3 37 41 52 370 77 399 490

U-Con man #1 7.3

U-Con man #2 14

Feinc operator 6.2 96 38 110 175 154 980 840

Barium operator 38 130 144 126 280

Feed operator 40.8 23 100 150 110 126 910 476

C-3 wash filter operator 79 32 48 120 116 497 476

C-3 adjustments operator 22 420 120 497 476

C-3 centrifuge operator 42 630 52 140 567 476

Ore Room operator 140 170 370 392 350 13,720 4970

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 76 126 273 364

Raffinate operator 216 11 8 170 68 154 273 364

QM-2 (Orange) packager 268 1,961 120 130 130

Furnace operator 12 33 55 96 150 1,400 5320 24500

Furnace room sampler 3150

Reduction area (furn room) lead oper’r 22 25 28 69 54 147 686

LF-9 (Brown) packager 364 11270 39200

Nitric acid recovery operator 20 9.6 19 44 35 99 46 364

Pot Room operator 234 113 45 190 100 336 770 7,770 32200

Metal dissolver #1 204

Metal dissolver #2 21

MGX operator 29 68 52 94

Utility operator 88 129 94 97

Miller (Mill Room) X X 12,600 46200

Pilot Plant group leader 7.5 6.9 3.1 105 91 245

Pilot Plant lead operator 7.7 8.8 6.1 77 116 105 91 245

Pilot Plant technician 1,940 9.2 6 77 116 105 91 245

Production superintendent 7.7 8.8 56 25

Experimental continuous furnace 8,540 X X X

Asst. production superintendent 18 21 26

General/Asst foreman 14 18 30 50

Foreman 17 21 29 58 161

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

Shift foreman 19 25 27 81 96 161

Cloth & Training Grp Lead Operator 23 25 2520

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 30 245

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 25 239 24 

(239)

Laboratory personnel 42 2.9 30 23 21

 MCW Laboratory west section 21 30 245

 MCW Laboratory east section 13 30 245

Powder sample technician 56.5 217? 448

Metal room sampler 420

Outside sampling man 22.5

Sample Room supervisor 41 245 448

Laboratory Office personnel 42 2 5.6
Document No. ORAUT-TKBS-0005 Revision No. 02 Effective Date: 06/14/2007 Page 186 of 225

Weighted average concentration, alpha dpm/m3

Occupation

May 

1956

May 

1954

Oct 

1953

Jan 

1953

Jan 

1952

Aug 

1950 1949

Oct-Nov 

1948

May 

1948

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

Warehouse man -- K-65 sampler 350 270 230 84 189 196

Warehouse man 5.8 10 20 38 46 84 189 196

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

Health Office - personnel 8.1 11 15 14 0 7.0 14

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 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 189

Digest & feed AM 133

Raffinate and C-3 AM 161

Ether & NA House AM 77

Welders, pipefitters, etc. 98 128

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 7.7 33

AEC Engineer 19 9.9 31 7

MCW Office personnel 1.5 2 2.9 0.7 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

The first set of 1948 data (Oct-Nov 1948) is from MCW (MCW 1949d (repeated in AEC 1949b and MCW 1950s), the second set (May 

1948). The 1949 and 1950 data are from MCW (MCW 1950q) and AEC(AEC 1953); the May 1952, January 1953, and the October 1953 

sets of survey data are from AEC (AEC 1954c); the May 1954 survey data are from AEC (AEC 1954d); and the May 1956 survey data are

from AEC (AEC 1956b). For some occupations (mostly office types), the May 1948 concentration was the average in the work area, not a 

DWE, so that the level shown would be higher than what the worker actually experienced.

Notes
Document No. ORAUT-TKBS-0005 Revision No. 02 Effective Date: 06/14/2007 Page 187 of 225

Table A-14. Plant 6E measured daily weighted average exposure concentrations. 

Weighted average concentration, alpha dpm/m3

Occupation 7/1956 3/1955 6/1954 11/1953 4/1953 10/1952

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

The 10/52, 4/53, and 2/54 sets of data are from AEC (AEC 1954d); the 6/54 survey data are from AEC (AEC 1954e); the 3/55 survey data 

are from AEC (1955e); and the 7/56 survey data are from AEC (AEC 1956c).
Document No. ORAUT-TKBS-0005 Revision No. 02 Effective Date: 06/14/2007 Page 188 of 225

Table A-15. Plant 7 measured daily weighted average exposure concentrations. 

Weighted average concentration, alpha dpm/m3

Occupation 7/1956 3/1955 6/1954 11/1953 4-5/1953 10/1952 9/1951

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

The 9/51 survey data are from AEC (AEC 1951a); the 10/52 survey data are from AEC (AEC 1952a); the 4-5/53 and 11/53 

survey data are from AEC (AEC 1954f); the 6/54 survey data are from AEC (AEC 1954g); the 3/55 survey data are from 

AEC (AEC 1955d); and the 7/56 survey data are from AEC (AEC 1956d).

The 5/53 data includes the screw-pulling operation, which was nonroutine.

Table A-16. Plant 7E measured daily weighted average exposure 

concentrations. 

Occupation

Weighted average concentration, 

alpha dpm/m3, 3/1955

Ionium plant operator 0.06

Ionium plant lead operator 0.1

Overall average 0.07

Survey data are from AEC (1955e). The ionium plant was the thorium 

processing plant, in operation from mid-1955 through March 1957. Some benchscale processing was done in the spring of 1955 also.
Document No. ORAUT-TKBS-0005 Revision No. 02 Effective Date: 06/14/2007 Page 189 of 225

Table A-17. 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

“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.
Document No. ORAUT-TKBS-0005 Revision No. 02 Effective Date: 06/14/2007 Page 190 of 225

Table A-18. Job titles and classifications, with geometry factors. 

Plant Job title or classification Notes

Geometry 

classification, %

AP ROT ISO

7 36’ Level operator Subclassification of 7 Furnace operator (rotating jobs): mainly the UO2-to-UF4

reactor area

--- --- ---

6E 4th saw man Mostly same as Saw man 90 10

6 AEC engineer Spent time in Plants 4, 6, 6E, and 7 but office was in 6 10 90

6 AEC Office personnel Spent time only in AEC office in Plant 6 100

All Area mechanic May have worked in all buildings, but appears that he was usually dedicated 

to one building or process step

50 50

6E Asst foreman See Shift foreman 50 50

7 Asst plant superintendent May have been in charge of only Plant 7 or of the whole site 10 90

6 Asst production superintendent 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 Worked in the U products warehouse, apparently 25 75 

6 Bag inspector Checked dust collector bags 75 25

6 Barium operator Worked in the barium salt addition phase of digestion 50 25 25

6E Billet grinder Cleaned and finished billets after recasting and before shipping 90 10

4, 6E Blender See Lime blender 50 50

6 Boiler House operator Presumably worked to provide steam for the boildown processes 100

4 Bomb charger Blended UF4, Mg; charged bomb; dumped out dust collectors 75 25

4 Bomb makeup operator Blended UF4, Mg; charged bomb; dumped out dust collectors 75 25 

4 Bomb makeup/deslagging 

(slag grinding) operator

Performed combined functions of Bomb makeup operator and Slag grinding 

operator

75 25

6E Bottom “F” machine operator Charged bomb (using F machine): UF4-Mg mixture, Mg 75 25

4, 6E Bottom furnace operator/man Removed assembly from recast furnace, removed mold, put in new assembly 75 25

6E Breakout operator/man Removed (broke out) derby from bomb 75 25 

6 Brown furnace operator/

unloader/ packager

See UO2 operator/unloader/packager 75 25

6E Brushing man Same as Chipper 75 25

6E Burnout man Removed broken crucible, knocking lid off 80 20

6 C-3 adjustments operator Cleaned wash precipitate filter press; other duties 50 25 25

6 C-3 centrifuge operator Operated and “plowed off” Bird(?) centrifuge (digestion process) 50 25 25

6 C-3 wash filter operator Operated and cleaned the wash precipitate filter press 50 25 25

6E Cage grinding man Mostly same as Saw man 90 10 

4, 6E Cage operator/man Cleaned and finished billets after recasting; handled scrap in cage 90 10

6E Cage saw man Mostly same as Saw man 90 10

6E Capping man Put the valved top on the Mg-UF4 drum for jolting? 75 25

4 Carpenter May have worked in all buildings, not dedicated to one 25 50 25

4 Casting operator/man Operated the billet casting furnace 75 25

6 C. Eng See Chemical engineer 25 75

4 Ceramic (technician?) Split time: ceramic lab (Ceramic Pilot Plant), Production Research Lab 50 50

4 Charger Same as Bomb charger 75 25

4 Charge firing (man) Furnace operator for bomb furnace 75 25

6E Chemical engineer Worked in production aspects; had production area access 25 75 

4 Chemical technician Did chemical analyses in Analytical Lab; possibly some assays 90 10

4 Chemist Did chemical analyses in Analytical Lab; possibly some assays 90 10

4 Chief chemist Did chemical analyses in Analytical Lab; possibly some assays 90 10 

6 Chief guard Spent time in Plants 4, 6, 6E, and 7 but base was in Plant 6 100

4, 6E Chipper Deslagged and cleaned derbies (and dingots?) 75 25

4, 6 Cleanup man Miscellaneous cleanup activities 50 50

4, 6E,

7 

Clerk 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 Cut and replaced filter cloth for Feinc and similar filters 50 25 25

6 Cloth & Training Group lead 

operator

Apparently dual position: coordinated cloth ops, headed Training Grp 10 90

6 Clothes issue man Issued work and protective clothing in the locker rooms 100

6E Crucible assembler Assembled crucible and mold assembly in billet production process 75 25

6E Crucible loader Loaded the crucible assembly into the recast furnace 75 25

6,6E,7 Decontamination man Did decontamination in all plants, on boxcars, in Plant 6 Decon Room 50 25 25

7 Decontaminator Same as Decontamination man 50 25 25

6E Derby chipper Same as Chipper 75 25

4 Derby unloader Unloaded bomb from furnace 75 25

6 Digest area lead operator Worked on the digestion process (up to extraction) 50 25 25

6 Digest operator Worked on the digestion process (up to extraction) 50 25 25
Document No. ORAUT-TKBS-0005 Revision No. 02 Effective Date: 06/14/2007 Page 191 of 225

Plant Job title or classification Notes

Geometry 

classification, %

AP ROT ISO

4 Dingot operator/forging Prepared, loaded, removed dingot bomb; probably operated furnace 50 25 25

6 Dispensary (personnel) Medical-pharmaceutical personnel; no prod area access assumed 100

6E Electrician May have worked in all buildings, not dedicated to one 25 50 25

4, 6E,

7 

Engineer 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 Worked in the ether house (operating tanks, valves, etc.) 50 50

6 Experimental Continuous 

Furnace

Pilot Plant project for eventual plant-scale use in Plant 7 25 50 25

6 Extraction area lead operator 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 Loaded Mg, UF4 into bomb and sealed it; may have blended them 75 25

6E F machine operator Charged bomb (using F machine): UF4-Mg mixture, Mg 75 25

6 Feed operator 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 Feed sampling Sampling of feed (soda sat, MGX, etc.) 50 25 25

6 Feinc operator Operated Feinc filter; cleaned out cake; washed out feed Niagara 50 25 25

7 Filter operator Operated solids-removal(?) filters in Slag Separation Plant (Bldg 701) 50 50

7 Fire Marshall Did inspections in Plants 4, 6, 6E, and 7, but base was in Plant 6 90 10

All Foreman Apparently dedicated to individual plant, but may have worked in more 50 50 

4 Forge press lead operator Operated the forge press in dingot finishing 75 25

4 Forge press manipulator Same as Forge press operator 75 25

4 Forge press operator Operated the forge press in dingot finishing 75 25

4 Forge press salt bath man Operated the salt bath segment of dingot finishing 75 25

All Fork truck operator/driver Same as lift truck driver; some did work at SLAPS and with railcars 25 75

4 Furnace and saw man Divided duties: see Furnace operator and Saw man 75 25 

4, 6E Furnace loader Loaded bomb or crucible assembly into furnace, depending on plant 75 25

4, 6E Furnace operator Operated recasting (billet) furnace 75 25

6 Furnace operator Operated the UO3-to-UO2 (Rockwell) furnace 75 25

7 Furnace operator Job (Plant 7) rotated 36’ Level, Panel Board, and Sampler & Cleanup tasks: 

see individual job titles

75 10 15

4 Furnace puller Unloaded UO2-to-UF4 furnace? 75 10 15

4 Furnace tender Tended UO2-to-UF4 furnace 75 25

6E Furnace unloader Unloaded bombs from furnace, cleaned out residue(?) 75 25

6, 6E,

7 

General foreman UF4 production (for Plant 6E); or may have spent time in all plants 50 50 

6E Graphite shop personnel Assumed to be doing only clean work manufacturing graphite molds 100

7 Green packager Packaged UF4 (green salt) 75 25

4 Grinding and lead operator Apparently work in the dingot area 50 25 25

All Guard Spent time in Plants 4, 6, 6E, and 7 and SLAPS but base was in 6 75 25

6 Health Office - office personnel Assumed to spend all time in Health Office 100

6 Health Office - other personnel Spent time in Plants 4, 6, 6E, and 7 but base in 6; assume no production area 

access

100

6 Health Office - plant monitor Spent time in Plants 4, 6, 6E, 7 but base in 6; production area access 50 50

6 Health Office - health surveyor Spent time in Plants 4, 6, 6E, 7 but base in 6; production area access 50 50

4, 7 HF operator 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 Worked in the Slag Separation Plant (Bldg 701) 50 50

6 Instrument Shop machinist Worked only in Instrument Shop; some equipment contaminated 10 90

6 Instrument Shop technician 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 Th-230 processing (mostly operating chemical tanks and filters) 75 25

7E Ionium plant operator Th-230 processing (mostly operating chemical tanks and filters) 75 25

4, 6E Jolter “Jolted” (air hammer?) bomb liner, other material to compress, remove 75 25

6 K-65 sampler See Warehouse man: K-65 sampler 90 10

4, 6E KB-2 lead operator Handled all aspects of derby (KB-2 production) 75 15 10

6 Laboratory Office personnel Worked in the Analytical Lab (Building 102) 100

6 Laboratory personnel Worked in the Analytical Lab (Building 102) 90 10 

6 Laundry lead operator Handled operations at the laundry (incl contaminated items) 10 90

6 Laundry operator 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 (raff) Worked in the Ledoux Lab pulverizing, prepping, assaying raffinate 90 10 
Document No. ORAUT-TKBS-0005 Revision No. 02 Effective Date: 06/14/2007 Page 192 of 225

Plant Job title or classification Notes

Geometry 

classification, %

AP ROT ISO

6 Ledoux Lab technician (raff) Worked in the Ledoux Lab pulverizing, prepping, assaying raffinate 90 10

6 Ledoux Lab technician (K-65) Worked in the Ledoux Lab assaying K-65 residue 90 10

6 Ledoux Lab technician (MgF2) Worked in the Ledoux Lab assaying MgF2, incl recycle MgF2 90 10

4 LF-9 (furnace) loader 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) Drove lift truck, handling miscellaneous materials, possibly at all plants, 

including SLAPS

25 75

4, 6E Lime blender Charge blender for derby production 50 50

4 Magnesium operator Handled magnesium storage and disbursal; limited exposure potential? 75 25 

7 Magnesium Room operator Handled magnesium storage and disbursal; limited exposure potential? 100

6E Maintenance man May have worked in all buildings; assume production area access 50 50

6 Maintenance Office clerk Worked only in Maintenance Office 100 

6 Maintenance supervisor Spent time in Plants 4, 6, 6E and 7 but base was in 6 (same person?) 25 75

Manufacturing Generic term for process/storage area work 75 25

4, 7W Metal fab Metal fabrication, i.e., the derby and recast or dingot areas 50 50

6 MCW engineer Spent time in Plants 4, 6, 6E and 7 but base was in 6 10 90

6 MCW Lab (personnel) 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

Spent time in Plants 4, 6, 6E, and 7, but probably only in office and 

construction areas

100

6 MCW Office maintenance Spent time in Plants 4, 6, 6E, and 7, but apparently only in office areas 100

6 MCW Office messenger Spent time in Plants 4, 6, 6E, and 7, but probably only in office areas 100 

6 MCW Office personnel Spent time only in MCW offices in Plant 6 (Bldg 112?) 100

All Mechanic See Area mechanic 50 50

6 Mechanic supervisor See Maintenance supervisor 50 50 

6 Medic Worked in dispensary presumably, but may have gone into production areas 

on occasion

100

6 Metal dissolver (#1, #2) 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 Sampled scrap and other metal for U or content? 75 25

6 MgX operator Processed the MgX feed material 25 50 25

4 Microscopist Worked full-time in the Microscopy Room, probably on U specimens 100 

4, 6 Miller Performed the UF4 pulverizing (4)? Milled UO3 (orange oxide)(6) 50 50

6E Millwright May have worked in all buildings, not dedicated to one 25 50 25

6 Nitric acid recovery operator Worked in the nitric acid recovery area (several buildings) 50 25 25

6 Nurse Worked in dispensary presumably, but may have gone into production areas 

on occasion

100

4, 6E Office employees Assume they are Plants 4 and 6E production(?) office-only personnel 100

6 Ore Room operator Handled ore, residue: storage, lidding, delidding, cleaning drums 70 20 10

6 Outside sampling man Sampled dust collectors; sampled drums (incl. opening and sealing) 50 50 

7 Panel board operator 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 Pickler Worker in area where derbies were “pickled,” i.e., put into acid bath 75 25 

6 Pilot Plant engineer Participated in experimental extraction processes and like activities 25 75

6 Pilot Plant group leader Participated in experimental extraction processes and like activities 75 25

6 Pilot Plant lead operator Participated in experimental extraction processes and like activities 75 25 

6 Pilot Plant technician Unclear what activities this Pilot Plant was carrying on 75 25

6 Pipefitter May have worked in all buildings, not dedicated to one 50 50

4 Plant superintendent May have been in charge of only Plant 4 or of the whole site 10 90

4, 6,

6E, 7 

Porter Some porters may have worked in a janitorial capacity, entering only office 

areas, while others apparently served production areas; may have worked 

in several plants or in only one

100

6 Pot Room operator Worked in the denitration (UO2-to-UF4) pot room 50 25 25

6 Powder sample technician Assume mainly did sampling and worked with small samples 75 25

6E Production clerk Kept records, did inventories; assumed to spend some time in production 

area

100

6E Production machinist Spent all time in graphite shop machining molds and the like 100

6 Production Office clerk Kept records, did inventories; assumed to spend some time in production area 100 

6 Production Office secretary Assumed spent all time in office 100

6 Prod Research Lab personnel Worked in the Production Research Lab 90 10

6 Production superintendent Spent time in Plants 6, 6E, and 7 production areas and Plant 6 offices 50 50 

7 QM-2 dumper Loaded UO3 into trays for the UO3-to-UF4 conversion 75 15 10
Document No. ORAUT-TKBS-0005 Revision No. 02 Effective Date: 06/14/2007 Page 193 of 225

Plant Job title or classification Notes

Geometry 

classification, %

AP ROT ISO

7 QM-2 hoister Same as QM-2 dumper 75 15 10

6 QM-2 loader See UO3 loader 75 25

6 QM-2 (orange) packager “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 Handled the various residues from the filters (drummed, sampled); spent up 

to half time in Pot Room

75 25

6E Recast furnace lead operator Operated the billet casting furnace 50 50

6 Receiving clerk Recorded and inventoried incoming shipments of ore and U products 50 50

7 Record clerk (A, B) 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 Operated UO2 (Rockwell) furnace; assumed to load UO3, unload UO2 75 25

6E Reduction (area) lead operator Operated the bomb (UF4-to-U metal) furnace 50 25 25

6E Reduction furnace operator Operated the bomb (UF4-to-U metal) furnace 50 25 25

4,6E Residue man 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 Worked full-time in the Safety Office 100

7 Safety inspector Did inspections in Plants 4, 6, 6E, and 7 (production area access) 10 80 10

4 Salt bath man Same as Forge press salt bath man 75 25

7 Sampler and cleanup man Subclassification of 7 Furnace operator (rotating jobs): sampling UF4, cleanup 

of furnace platforms

--- --- ---

6 Sample Room supervisor Supervisor operations in the Sample Room (probably in Bldg 111) 75 25

4,6E Saw operator/man Removed billet from quench tank; ground, sawed, and weighed it 90 10 

6 Security office (personnel) Assumed to be other than guards (e.g., clerical) 100

6,6E Shift foreman Assumed to be generic, although may have covered several plants; default 

geometry factors at right may be used

50 50

4 Shipping & Receiving 

(personnel)

Handled the receipt and shipment of U products 50 50

6 Shotgun Lab analyst 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 Blended C-liner and other slag from Plants 4 and 6 for use in bombs 75 25

6E Slag building operator Ground C-liner and other slag, sorted it via shaker tables, drummed it 75 25

4 Slag man, slag grinding 

operator

Mostly the same as Chipper 75 25

6 Soluble feed operator Loaded soda salt (“diuranate”) at appropriate point in digestion process; see 

also Feed operator

75 25

6 Stockroom clerk Stockroom possibly in Bldg 112. Spent time in Receiving (Bldg 101) 100

6 Stockroom foreman Stockroom possibly in Bldg 112. Spent time in Receiving (Bldg 101) 100 

6 Sump recovery operator Worked on U recovery from sump fluids 25 50 25

6E Supervisor Probably same as Technical supervisor 50 50

7 TA-7 hoisting operator (hoister) Hoisted and loaded UO2 into reactor for conversion to UF4 (TA-7) 25 50 25

4,7 TA-7 packager Packaged UF4 (green salt) 75 25

4 TA-7 Pilot Plant (personnel) Assumed to be technicians and operators producing UF4 (green salt) 50 50

4 TA-7 unloader (operator) Unloaded UF4 (green salt) from hydrofluorination reactor 50 50

7 Tables operator Worked in the Slag Separation Plant (701) operating shaker tables 25 75

6E Technical engineer Same as Chemical engineer or (Process) Engineer 25 75

6E Technical superintendent May have spent time in all the plants 25 75

All? Technical supervisor Unclear what duties were; probable frequent production area access 50 50 

Tinner Type of metal worker 25 50 25

4 Top cleaner Cleaned top furnace in billet (YM-5) production; not Top seat man 75 25

6E Top “F” machine operator Charged bomb (using F machine): UF4-Mg mixture, Mg 75 25

6E Top furnace operator/man Operated recasting (billet) furnace, removed crucible parts 50 50

6E Top-off operator/man “Topped off” bomb with slag, bolted on lid 75 25

4 Topper Same as Top-off operator? 75 25

4 Top seat man Involved in billet production, furnace area; same as Top-off operator? 75 25 

6 Trainer Duties not clear. Assumed to train operators, especially in filter work 10 90

6 Trainman Associated with packing/unloading of railcars; perhaps was railcar 

dispatcher/coordinator/inspector; associated with the warehouse

25 50 10

6 Truck operator/driver 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
Document No. ORAUT-TKBS-0005 Revision No. 02 Effective Date: 06/14/2007 Page 194 of 225

Plant Job title or classification Notes

Geometry 

classification, %

AP ROT ISO

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 Same as top furnace man 50 50

6 U-con man (#1, #2) 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 Worked in office and vacuum fusion area; did some hands-on U work 90 10

4 Vacuum fusion technician Worked in office and vacuum fusion area; did some hands-on U work 90 10

4 Vertical lathe (operator) Scalped dingot after casting and before forge-pressing 75 25

6 Warehouse foreman Handled storage of ore and U products 25 75

6 Warehouse man: K-65 samplerNew job, 1953: sampled K-65 residue, plus typical warehouse duties 90 10

6 Warehouse man (other) Handled storage of ore and U products 25 75 

6 Weighmaster Duties not clear. Assumed to perform or approve ore, K-65 weighings 25 75

6,7 Welder May have worked in all buildings, not dedicated to one 10 90

6E YM-5 lead operator Worked on billet (YM-5) casting 80 20

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.

Table A-19. 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-

1958

1957-

1958

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 3640 3640 3640 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

This table is derived from the data given in Table A-12 and from supplementary information in the references.

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. 
Document No. ORAUT-TKBS-0005 Revision No. 02 Effective Date: 06/14/2007 Page 195 of 225

Table A-20. 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 -

1958

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/assistant 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?

This table is derived from the data given in Table A-13 and from supplementary information in the references.

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.

* Outside sampling and U-Con positions began in 1955

** Pilot Plant work began in 1948.

*** Milling work ended by June 1949.

  These jobs apparently ended in 1954 or were subsumed in other job titles

** *** 1961 dpm/m3 in 1955, 268 in 1956-58
Document No. ORAUT-TKBS-0005 Revision No. 02 Effective Date: 06/14/2007 Page 196 of 225

Table A-21. 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 -1958

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

This table is derived from the data given in Table A-14 and from supplementary information in the references.

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.
Document No. ORAUT-TKBS-0005 Revision No. 02 Effective Date: 06/14/2007 Page 197 of 225

Table A-22. Radioactive 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 –

1958

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

This table is derived from the data given in Table A-15 and from supplementary information in the references.

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.

UO3 processing operations continued until July 1958 (Mallinckrodt 1994), but most other operations stopped in 1957. Plant 7E was said to 

be in its startup phase when the only known measurements were taken in March 1955 (AEC 1955e); these may have been taken as part of 

bench-scale processing (AEC 1955c). Thus as AEC (1955e) 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 an end date of 

March 1957 should be taken. 

* These positions began after July 1955. Clearly, this was in the pilot plant (as implied by the information in AEC 1955c). 

** The 36’ Level operator position began in 1953, the panel board operator position in 1952.

**The ionium pilot plant work began after July 1955 and continued until March 1957 (AEC 1955c). 
Document No. ORAUT-TKBS-0005 Revision No. 02 Effective Date: 06/14/2007 Page 198 of 225

Table A-23. Early radon data, 1945. 

% Tolerance

Ventilation 

condition Reference

Bldg 40 storeroom, 1000 lbs ore and active residues 724 F MED 1945f

Bldg 40, center of storeroom between stacks of residue drums <15 F MED 1945j

Bldg 40, residue storage: BZ, residue work -- floor covered with residue <24 MED 1945g

Bldg K, center of storeroom with ore and residues 90 F MED 1945j

Bldg K storeroom, center, 2-3 drums fresh residue and open drums of ore 25 F MED 1945f

(Bldg K) Pilot Plant storeroom with 2000-3000 lbs of ore 50 F MED 1945m

(Bldg K) Pilot Plant, over digestion tank during addition of ore 147 N MED 1945m

(Bldg K) Pilot Plant, center of Pilot Plant during digestion 132 N MED 1945m

(Bldg K) Pilot Plant, over digestion tank 15 minutes after digestion 18 N MED 1945m

(Bldg K) Pilot Plant, N end, away from operation 216 N MED 1945m

Bldg K, N end of room, away from operation <15 N MED 1945h

Bldg K, center of room <22 N MED 1945h

Bldg K, center of room, no operation 26 N MED 1945j

(Bldg K) Pilot Plant, discharge end of Feinc filter during filtration <15 N MED 1945m

Bldg K, S end near Feinc filter, not operating 29 N MED 1945j

Bldg K, S end of plant near Feinc filter, not operating <15 N MED 1945h

Bldg K, vicinity of Feinc filter, no operation <12 W MED 1945f

Bldg K, work desks at NE corner 12 W MED 1945f

Bldg K, work desks, center of room 12 W MED 1945f

Bldg K, alley outside at exhaust <11 MED 1945f

Bldg K alley, directly under stack emitting brown fumes <15 N MED 1945j

Storeroom on RR siding, between 2 tiers of residue drums, 10’ from each 32 MED 1945g

Plant 4, ore storage: center of room with 1800 lbs ore 211 F MED 1945h

Plant 4(?), W end of ore room during full operation 65 F, N MED 1945i

Plant 4, W end or ore milling area near sifter, full operation 58 F, N MED 1945i

Plant 4, center of ore milling room near mill 113 F, N MED 1945i

Plant 4, operator’s hand when unloading ball mill, 300 lbs of ore, fan on 407 F MED 1945f

Plant 4, near mill, 10 min after unloading, fan on <15 FW MED 1945f

Plant 4, opening at top of empty mill <25 MED 1945f

Plant 4, ore milling area: no operation for 3 days 30 N MED 1945h

Plant 4, residue storage: directly above drums of residue 763 MED 1945g

Plant 4, laboratory: center of cage <26 N MED 1945h

Plant 6 ore storage warehouse, SE corner 55 N MED 1945k

Plant 6 ore storage warehouse, center of empty aisle 21 N MED 1945k

Inside [unspecified] lab, no activity <11 MED 1945f

* Fan on (F), windows open (W), natural ventilation (N)

The data above was taken at various times in 1945. The tolerance level was 1.0 x 10-10 Ci/L of radon.

“Bldg K” was apparently Building K1E, also known as a pilot plant.
Document No. ORAUT-TKBS-0005 Revision No. 02 Effective Date: 06/14/2007 Page 199 of 225

Table A-24. Measured radon concentrations, 1947-1949, in units of 1 × 10-10 Ci/L. 

1947 1948 1949

No. Min

Med/

Mean Max No. Min

Med/

Mean Max GSD No. Min

Med/

Mean Max

INDOOR AREAS

Ore Milling 11 0.03 0.39 0.81

Ore Rm: open drums* 0.58 8.50 260

Ore Storage 27 0.46 3.28

Ore Storage 58 0.03 0.30 2.86

Thawing oven, K-65

 Heat off, fan off 28 3.84

 Heat off, fan on 36 0.60

 Heat on, fan on 18 4.06

 Outside, heat on 12 1.09

 Outside, heat off 24 0.07

Thawing oven, K-65** 4 540 622 805

Scalehouse 0.10 21 0.00 4.05 33 44 0.03 0.45 4.4

Scalehouse 2 .60 193 0.03 2.02 32.8 158 0.03 1.03

Scalehouse 5 .09 .27 .84

Scalehouse/Ore Storage/Warehouse 4 .55 .69 .92 5.00

Scalehouse Sample Rm 46 0.0 2.40 13 6 0.22 4.10 19

Scalehouse Sample Rm 3 .06 .09 .12 68 0.03 2.84 25

Digest/Feed 4 .05 .23 .52 57 0 0.39 7.8

Extraction Cells 14 0 1.08 4.1

Centrifuge Area 2 .38 74 1.00 74 0.00 0.76 3.86

Centrifuge Area 141 0 1.13 12 48 0 1.55 12

Feinc/Filter/C-3/Raffinate/Cloth Storage/Niagara 4 .08 .12 .2 9 0.18 1.98

Feinc/Filter/C-3/Raffinate/Cloth Storage/Niagara 56 0.00 0.05 9.74

Feinc/Filter/C-3/Raffinate/Cloth Storage 186 0 12.3 894 172 0 0.75 468

Ledoux Lab 30 0 0.08 0.4

Recovery area (aisle betw Recovery and feed makeup) 4 .05 .16 .31

Sump by M-19, skid of recovery cake beside sump .26

Maintenance Shop (aisle between shop and Feincs) 4 .04 0.11 .18

Dispensary 3 .04 .07 .11

YARDS AND OTHER OUTDOOR AREAS

General Plant 6 35 0 0.19 2.1

Bldg 104/Scalehouse 3 .07 0.17 .36 9 0.07 1.47

Bldg 104 (Pilot Plant) 7 0 0.15 0.6

Warehouse 17 0.10 2.05

Warehouse exhaust 33 0.1 0.91 3.3

Near warehouse, K-65 drum welding 8 0.11 2.84

Labs, guard office 8 0.06 1.38

Parking lot area 9 0.13 1.56

Scalehouse intake/exhaust 3 0.12 2.06

Scalehouse exhaust 18 0.13 3.08

Scalehouse exhaust 1 0.93

Scalehouse exhaust 24 0 2.2 49

Ore storage intake/exhaust 6 0.68 2.44

Ore storage exhaust 2 0.31 1.77

Railcar, ore, fans on* 2 0.5 <1.00 7

Drums outside Scalehouse 4 0.0 0.39 0.90

K-65 drums in sun** 2 637 654 670

SLAPS

In K-65 storage shed 5 4.2 3.2

At K-65 storage shed 19

15 ft from K-65 shed 2 0.04 0.99 24 93

25-30 ft, K-65 shed 2 7.8 1.6 6.6

50 ft from K-65 shed 3 3.8 2.1 4.4

50-120 ft, K-65 shed 13 2.1 5.0

100 ft from K-65 shed 2.4

150 ft from K-65 shed, at guard shack 3 0.00 0.21 1.1 14 1.4

180-250 ft, K-65 shed 5 1.1 5.8

200 ft from K-65 shed 0.94
Document No. ORAUT-TKBS-0005 Revision No. 02 Effective Date: 06/14/2007 Page 200 of 225

Values in this and the next table are taken from AEC (AEC 1948b; AEC 1949; AEC 1948j; AEC 1949j) and MCW (various, MCW 1950b; 

MCW 1948I; MCW 1948e; MCW 1948j; MCW 1948e; MCW 1951c; MCW 1949f).

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 were given in full, a lognormal distribution was used for the analysis and the 

geometric mean and geometric standard deviation (GSD) are given 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.

Control readings were taken in the Dispensary. Measurements were taken at face level whenever possible.

* The single asterisked items are an exception to the note above. For the Ore Room case, the value in the Min column is the point at 

which the concentration drops below 1 x 10-10 Ci/L (25 minutes), the value in the Mean column is the average concentration over the 

elapsed time (i.e., the 25 minutes), and the value in the Max column is the value at time zero. For the railcar case, the value in the Min 

column is the average of readings at about 3.5 hours (stable concentration), the value in the Mean column is the point at which the 

concentration drops below 1 x 10-10 Ci/L (about 20 minutes), and the value in the Max column is the average value at time zero. More 

information can be found in MCW (MCW 1949m) and MCW (MCW 1948h) respectively.

** The double asterisked items correspond to samples taken of the standard 30-gallon K-65 drums sitting inside sealed 55-gallon drums 

(to facilitate sampling) after heating in the thawing oven or being allowed to heat in the sun (MCW 1949r); these thus represent the 

maximum radon evolved over 2-4 hours of heating, concentrated in the larger drum head space but not actually in the breathing zone.
Document No. ORAUT-TKBS-0005 Revision No. 02 Effective Date: 06/14/2007 Page 201 of 225

Table A-25. Measured radon concentrations, 1950-1957, in units of 1 × 10-10 Ci/L. 1950 1951 1952 1953 1954 1955 1956 1957

No.

Med/

Mean GSD No.

Med/

Mean GSD No.

Med/

Mean GSD No.

Med/

Mean GSD No. Min

Med/

Mean Max GSD Min

Med/

Mean Max No.

Med/

Mean GSD No.

Med/

Mean GSD

INDOOR AREAS

Scalehouse 35 0.09 5.9 95 0.09 6.0 66 0.01 5.0

Scalehouse 1.02

Scalehouse/Ore 

Storage/Warehouse

74 0.01 5.4 5 0.01 1.00

Scalehouse Sample 

Room

0.09 0.24 0.50

Digest/Feed 2 0.05 11 17 0.12 5.2 57 0.03 3.8 18 0.01 5.1 3 0.03 5.5

Extraction Cells 33 0.36 5.9 92 0.37 5.3 59 0.28 3.8 101 0.26 5.8 33 0.01 7.0 3 0.01 8.7

Centrifuge Area 8 0.11 2.3 35 0.06 4.8 95 0.12 5.0 55 0.05 5.6 51 0.07 6.5 7 0.01 5.2 3 0.01 1.0

Feinc/Filter/C-3/

Raffinate/Cloth 

Storage/Niagara 

34 34 0.18 3.7 94 0.14 4.1 60 0.10 4.8 114 0.14 6.2 <.01 0.50 1.7 41 0.07 5.8 5 0.01 6.8

Orange Packing 1 0.12 0.12 0.12

Pot Room 2 0.02 2.2

Ledoux Lab 6 0.10 1.3 10 0.02 3.4

Shotgun Lab 2 0.04 5.8

Other Lab

(Research/Control/Xray/MY)

3 0.01 1.0 2 0.04 8.0

6E Breakout 3 0.01 1.0

6E Recast 4 0.01 1.0

Recovery area 1 0.11 0.11 0.11

Decontamination Room 1 0.01 0.01 0.01

Metal Dissolver Bldg 4 0.01 2.6

NA House 3 0.01 16

Ether House 2 0.02 2.7

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 1.0

Maintenance Shop 3 0.04 6.9

Smoking Room 1 0.01 0.01 0.01

Production Office 1 0.01 0.01 0.01

Dispensary 2 0.04 6.1

YARDS AND OTHER OUTDOOR AREAS

Ether House/ Bldg 109 1 0.01 0.01 0

Drum storage outside 

Bldg 115

3 0.30 8.6

Values in this and the next table are taken from AEC (AEC 1948b; AEC 1949; AEC 1948j; AEC 1949j) and MCW (MCW various, MCW 1950b; MCW 1948I; MCW 1948e; MCW 1948j; MCW 1948e; MCW 

1951c; MCW 1949f).

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 were given in full, a lognormal distribution was used for the analysis and the geometric 

mean and geometric standard deviation (GSD) are given 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
Document No. ORAUT-TKBS-0005 Revision No. 02 Effective Date: 06/14/2007 Page 202 of 225

Table A-26. RESERVED

Table A-27. Contamination levels and associated dose rates from work clothing. Surface contamination measurements for various types of work clothing

Spot Whole garment or group

Item Max Average Max Average Contamination notes

Regulated coveralls 80% had ≥ 1-2 in2

spot >3000 cpm

Most coveralls 

had more than 

1 spot >1 

mr/hr

60% had ≥ 1 

mrep/hr over 

whole garment

1.5 mrep/hr; range, 0.2 to 

12; 100% had avg >100 

cpm per 2 in2

Apparent U spots on 70%; 

area was from 10 cm2 to 

30% of total area 

Nonregulated coveralls 30% have more than 

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”

Gloves, contaminated 

on the inside

47 mrep/hr Range 23-47 mrep/hr Beta dose rate, on contact 

with the inside

Shielding of beta radiation by various materials

Item Distance Reduction factor Std deviation

Coveralls, 9-oz denim 5 inches–3 ft .78 7.5%

Gloves, neoprenecovered cotton.50 3 gloves were measured

Measurements of surface contamination on clothing are from Utnage (1958b); measurements of beta shielding and exposure rates inside 

gloves are from AEC (AEC 1950j). Measurements were taken after wearing but before washing; gloves contaminated on the inside were 

taken at random from workers actually using the gloves and were not discarded gloves. 

The beta reduction factor is the ratio of shielded dose rate to unshielded dose rate. The source was a sheet of uranium metal, 18” x 24,” in 

equilibrium with UX1 and UX2.

The “smocks” appear to be cover clothing.

Table A-28. RESERVED.

Table A-29A. RESERVED. 

Table A-29B. RESERVED.

Table A-29C. RESERVED. 

Table A-29D. RESERVED. 

Table A-29E. RESERVED.

Table A-29F. RESERVED. 

Table A-30. RESERVED. 

Table A-31A. RESERVED. 

Table A-31B. RESERVED. 

Table A-31C. RESERVED. 

Table A-31D. RESERVED. 

Table A-31E. RESERVED. 

Table A-31F. RESERVED. 
Document No. ORAUT-TKBS-0005 Revision No. 02 Effective Date: 06/14/2007 Page 203 of 225

Table A-32. Exposure rates from K-65 residue and Q-11 ore, in mR/hr (AEC 1949b, 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 railcar

One drum group centerline 

perpendicular to long side of 

railcar (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 railcar, over one group (A); 

over empty center (B)

Contact 21 11

5 railcars in a row 

containing unspecified 

amount of K-65

Along a line perpendicular to the 

axis of the line of railcars, even 

with the center of the middle car

10 50

25 22

43 13

50 10.2

The total length of a railcar was given as 57 feet.
Document No. ORAUT-TKBS-0005 Revision No. 02 Effective Date: 06/14/2007 Page 204 of 225

Table A-33. Measured dose rates in various work areas. 

Exposure,

% tolerance

Exposure, 

mrep/hr or 

mR/hr* Year Reference

PLANTS 1 and 2

Soda salt, near-contact: gamma; beta 20; 20 2.5; 13 1943 MED 1944f

Soda salt, 6”: gamma; beta 10; 20 1.2; 13 1943 MED 1944f

Vanadium sludge, over 1 month old, 10 cm from 216 jar: gamma; beta 1.5; 20 .19; 13 1943 MED 1944f

High-grade pitchblende ore, 55-gal drum, 1000 lbs ore: 60” 3.8 1945 Rochester 1945

High-grade pitchblende ore, 55-gal drum, 1000 lbs ore: 45” 6.1 1945 Rochester 1945

High-grade pitchblende ore, 55-gal drum, 1000 lbs ore: 30” 12 1945 Rochester 1945

High-grade pitchblende ore, 55-gal drum, 1000 lbs ore: 15” 29 1945 Rochester 1945

High-grade pitchblende ore, 55-gal drum, 1000 lbs ore: 6” 66 1945 Rochester 1945

High-grade pitchblende ore, 55-gal drum, 1000 lbs ore: contact 100 1945 Rochester 1945

High-grade pitchblende ore, 15-gal drum, 443 lbs ore: 60” 2.4 1945 Rochester 1945

High-grade pitchblende ore, 15-gal drum, 443 lbs ore: 45” 3.6 1945 Rochester 1945

High-grade pitchblende ore, 15-gal drum, 443 lbs ore: 30” 7.1 1945 Rochester 1945

High-grade pitchblende ore, 15-gal drum, 443 lbs ore: contact 90 1945 Rochester 1945

Ether extraction tanks 25 3.1 1943 MED 1944f

Residue drier (beta only) 120 75 1943 MED 1944f

“SM sub 2”: gamma; beta 42; 40 5.2; 25 1943 MED 1944f

“SN sub 4”: gamma only 30 3.8 1943 MED 1944f

“SB sub 3”: gamma; beta 35; 40 4.4; 25 1943 MED 1944f

Tank 1A cake from SN sub 2: gamma; beta 10; 120 1.2; 13 1943 MED 1944f

Tank 7 press cake, 10 days old, 10 cm from the edge: gamma; beta 20; 400 2.5; 250 1943 MED 1944f

308 [material] from Tank 7 press cake residue, 10 days old: beta only 400 250 1943 MED 1944f

Residue Chem. 4217, 2 months old, 10 cm from 2-lb bottle: gamma; beta 5; 30 .62; 10 1943 MED 1944f

Over top of Shed #1 recovery cake: gamma, beta 12; 30 1.5; 19 1943 MED 1944f

Acid press cake from tank treatment, 10” above: gamma; beta --- 6.3; 300 1944 MED 1945b

Barrel of residue after storage for 6 months, top: gamma; beta --- 6.2;50 1944 MED 1945b

Barrel of residue after storage for 6 months, 6’ from side: gamma, beta --- 6.2; 75 1944 MED 1945b

Caustic precipitate before it enters the drier, 5-6”: gamma; beta --- 6.2, 450 1944 MED 1945b

Plate of residue being dried in oven, 3,” beta only --- 250 1944 MED 1945b

Trays in drier, pile, in which NG & SNG residues are dried: 2 “ above: gamma; 

beta --- 5; 50 1944 MED 1945b

Trays in the drier, along the axis about 1’: gamma, beta --- 6.2, 250 1944 MED 1945b

Flat dish of residue on top of above crucible covering, 3”: gamma; beta --- 0; 1000 1944 MED 1945b

Concrete crucible covering, residue cooling for 4 hours, 3” above: beta --- 400 1944 MED 1945b

Flat dish of slag dust, 3,” beta only --- 0 1944 MED 1945b

Floor of Hood #6, 4,” where most 4th extractions spilled: gamma; beta 50; 30 6.2; 19 1944 MED 1944m

Most contaminated part of hood floor where shotgun samples dried: gamma; 

beta 70; 40 8.8; 25 1944 MED 1944m

Shotgun prep lab table top with daily change of brown paper, 4”: gamma; beta --- <13; <25 1944 MED 1945b

Lip of evaporating dish with 3rd-extraction liquor, .5”: gamma; beta 50; 30 6.2; 19 1944 MED 1944m

Lip of evapor’g dish w/ 4th-extraction liquor, .5” away, 2” above liquor: gamma; 

beta 100; 600 13; 375 1944 MED 1944m

Lip of evaporating dish with shotgun sample, .5”: gamma; beta 60; 100 7.5; 62 1944 MED 1944m

Lip of evaporating dish containing shotgun sample, 7”: gamma; beta 30; 20 3.8; 13 1944 MED 1944m

Lip of evaporating dish containing 5.17-g shotgun sample, 10”: beta only 20 13 1944 MED 1944m

Lip of evaporating dish, cleaned, that had held shotgun sample, .5”: gamma; 

beta 30; 20 2.5; 25 1944 MED 1944m

3.36 g shotgun sample, 3” above: gamma; beta 200; 1600 25; 1000 1944 MED 1944m

5.15 g shotgun sample in bottom of evaporating dish, 4”: gamma; beta 130; --- 16; >375 1944 MED 1944m

4.31 g shotgun sample, 7”: gamma; beta 100; 400 13; 250 1944 MED 1944m

5.15 g shotgun sample in bottom of evaporating dish, 7”: beta only 700 440 1944 MED 1944m

5.15 g shotgun sample in bottom of evaporating dish, 11”: beta only 340 215 1944 MED 1944m

5.15 g shotgun sample in dish, 11,” through rubber glove: beta only 170 106 1944 MED 1944m

Surface of rubber glove above: beta only 20 13 1944 MED 1944m

Outside bottom of evaporating dish with 3.36 g sample, .5”: gamma; beta 50; 140 6.2; 88 1944 MED 1944m

Bottom of cleaned evaporating dish (had held shotgun sample), 2”: gamma; 

beta 20; 40 2.5; 25 1944 MED 1944m

Glass shipping bottle with 3.08-g shotgun sample, 1”: gamma; beta 60; 100 7.5; 62 1944 MED 1944m

Shotgun package containing three bottles: gamma; beta --- 2.5; 0 1944 MED 1945b

Bottom of package with 5 shotgun samples to be shipped, 2”: gamma; beta 100; 0 13; 0 1944 MED 1944m
Document No. ORAUT-TKBS-0005 Revision No. 02 Effective Date: 06/14/2007 Page 205 of 225

Exposure,

% tolerance

Exposure, 

mrep/hr or 

mR/hr* Year Reference

PLANT 4

Sublining material, 6” from pail: gamma; beta 42; >400 5.2; >250 1943 MED 1944f

Insulating brick from recast furnace: beta only >400 >250 1943 MED 1944f

Freshly cast metal: beta only 10 6.2 1943 MED 1944f

Green salt, form normally handled, 1”: beta only --- 100 1944 MED 1945b

Green salt, contact with tray of green salt: beta only --- 108 1948 MCW 1948a

Green salt, exposure to any large surface (tray, open can, etc), 1’: beta only --- 64 1948 MCW 1948a

Above bench on which green salt is packed, 4”: gamma, beta --- <13; <25 1944 MED 1945b

Bomb step, exposure to large surfaces of green salt, 1’: beta only --- 50 1948 MCW 1948b

Bomb step, loading bomb, full bomb, 1’: beta only --- 53 1948 MCW 1948b

Bomb step, unloading bomb, knockout grate, 1’: beta only --- 3.6 1948 MCW 1948a

Bomb step, at edge of unloader jolter, 1’: beta only --- 3 1948 MCW 1948a 

Top of table where ingot is pushed out of bomb, beta only --- 100 1944 MED 1945b

Bomb step, derby transport table, 1’: beta only --- 5.3 1948 MCW 1948a

Bomb step, contact with unchipped derby: beta only --- 116 MCW 1948a

Bomb step, chipping, 1’ (chipper location): beta only --- 42 1948 MCW 1948a

Top of chipper’s table, beta only --- 125 1944 MED 1945b

Contact with the chipper’s table: beta only --- 465 1948 MCW 1948a

Ingot storage room, 4’ from pile of ripe metal, beta only --- 25 1944 MED 1945b

Ingot storage and packing room, 3’ from ore scrap pile in crib, beta only --- 25 1944 MED 1945b

Recasting step, cleaning crucible lids and stoppers, 1’ above tray: beta only --- 3000 1948 MCW 1948B

Recasting step, cleaning crucible lids and stoppers, chest high: beta only --- 280 1948 MCW 1948a 

Recasting furnace platform, 6” above, beta only --- 168 1944 MED 1945b

Recasting step, above top of furnace immediately after opening, 1’: beta only --- 1450 1948 MCW 1948a

Recasting step, above top of furnace, baffle & crucible lid removed, 1’: beta 

only --- 650 1948 MCW 1948a

Recasting step, contact with cleaned crucible lid: beta only --- 155 1948 MCW 1948a 

Recasting step, chest high while cleaning top of furnace: beta only --- 135 1948 MCW 1948a

Recasting step, cleaning furnace top, edge of top cleaning hood: beta only --- 220 1948 MCW 1948a

Recasting step, inside bottom of furnace before removing rolls, 1’: beta only --- 12 1948 MCW 1948a

Recasting step, inside bottom of furnace after cleaning, 1’: beta only --- 16 1948 MCW 1948a

Recasting step, average for general area of bottom furnace work: beta only --- 16 1948 MCW 1948a

Recasting step, above drum of D-7, 1’: beta only --- 140 1948 MCW 1948a

Recasting step, above D-7 table, 1’: beta only --- 270 1948 MCW 1948a 

Billets stored by desk and scale, gamma+beta 90-270 1953 MCW 1953e

Saw with no uranium billet in it, gamma+beta 100-150 1953 MCW 1953e

Empty furnace, lid off, top center//edge//1 foot, gamma+beta

300-

400//100//70 1953 MCW 1953e

Furnace lid and brush, center//edge, gamma+beta 1200//100 1953 MCW 1953e

Furnace bottom plate (operator foot position), gamma+beta 800 1953 MCW 1953e

Empty furnace, bottom opening, edge//1 foot, gamma+beta 300//100-250 1953 MCW 1953e

Used crucible before loading 380 1953 MCW 1953e 

Sample storage room (Plant 4), beta only --- 25 1944 MED 1945b

Storage and packaging, average for general area: beta only --- 3 1948 MCW 1948a

Storage and packaging, facing table of derbies, 1’ above table edge: beta only --- 21 1948 MCW 1948a

Storage and packaging, surface of cleaned derby: beta only --- 85 1948 MCW 1948a

Storage and packaging, box of rolls, 1’ above: beta only --- 5 1948 MCW 1948a

Slag from reaction bomb, 4,” beta only --- 250 1944 MED 1945b

Bomb step, edge of the slag barrel: beta only --- 66 1948 MCW 1948a 

Pile of slag in sample aging room, 2”: gamma; beta --- <25, 200 1944 MED 1945b

Piece of slag (~30 in3), 1.5”: gamma; beta --- 2.5; 100 1944 MED 1945b

PLANT 6 

Ore barrels, direct contact with a group --- 200 1947 AEC 1947a

Ore drum, 4” away through 2” of steel(?) shielding --- 9.5 1948 AEC 1948a

Removing lids from ore drums, hands --- 25-38 1948 AEC 1948d

Ore milling feeder tube, 3” away, gamma --- 53 1948 AEC 1948a

Ore milling hopper general vicinity, first platform --- 15 1948 AEC 1948a

Ore milling hopper, 8” away through 1/8” steel plate shielding --- 40 1948 AEC 1948a

Ore milling hopper, 3” away, first platform --- 20 1948 AEC 1948a

Ore milling hopper, 35” from revolving ore drum at first platform guardrail --- 7 1948 AEC 1948a

Ore milling gas blower, 15” from revolving ore drum --- 9.5 1948 AEC 1948a

By Tank M-2 on platform, 5 hrs after addition of 6000 lbs ore: 1 foot 100 1946 MCW 1946a
Document No. ORAUT-TKBS-0005 Revision No. 02 Effective Date: 06/14/2007 Page 206 of 225

Exposure,

% tolerance

Exposure, 

mrep/hr or 

mR/hr* Year Reference

By Tank M-2, on platform (catwalk): 6”; 18”; 36” from tank 230; 120; 90 1947 AEC 1949f

By Tank M-3 on platform while adding ore (2000 lbs in tank): 1 foot 70 1946 MCW 1946a

By Tank M-3, on platform (catwalk): 6”; 18”; 36” from tank 220; 120; 90 1947 AEC 1949f

Under Tank M-3, containing GLC: 1 foot 120 1946 MCW 1946a

By Tank M-4, on platform (catwalk): 6”; 18”; 36” from tank 80-200; 50-60; 

40

1947 AEC 1949f

By Tank M-72, containing wash water: 1 foot 50 1946 MCW 1946a

By Tank M-1, on platform: about 250 gal GLC (3 days old), 1 foot 90 1946 MCW 1946a

By Tank M-1, on platform (catwalk): 6”; 18”; 36” from tank 220; 180; 90 1947 AEC 1949f

By Tank M-83: three-quarters full of GLC (2 days old), 1 foot 220 1946 MCW 1946a

By Tank M-83: 6” from tank 500 1947 AEC 1949f

At operator position between skid and Tank M-14, on platform: 1 foot 120 1946 MCW 1946a

Above skid of GLC on M-14: 1 foot 280 1946 MCW 1946a

Between Tanks M-14 and M-20, midway, on platform (catwalk) 350 1947 AEC 1949f

By Tank M-14, on platform (catwalk): 6”; 18”; 36” from tank 450; 400; 250 1947 AEC 1949f

By Tank M-19, on platform:reslurrying skid cake batch: 1 foot 120 1946 MCW 1946a

By Tank M-19, on platform: 18” from tank 300 1947 AEC 1949f

By Tank M-19, on top of tank: 1 foot 250 1946 MCW 1946a

In front of Feinc filter while filtering reslurry batch: 1 foot 210 (2) 1946 MCW 1946a

At Feinc filter GI-4/GI-9: front (walkway); side; back; top (all at 6”) 400; 350; 500; 

900

1947 AEC 1949f

Near Feinc filter, on platform: gamma min//avg//max 5//16//50 1948 MCW 1948E

Feinc cloth change, < 3 feet from Feinc: gamma, beta 30; 69 1951 MCW 1951a

Feinc cloth change, general vicinity: gamma, beta 12; Negligible 1951 MCW 1951a

Leach (cell) cloth change, < 3 feet: gamma, beta 16; 158 1951 MCW 1951a 

Leach (cell) cloth change, general vicinity: gamma, beta 10; 40 1951 MCW 1951a

Wash (cell) cloth change, < 3 feet:gamma, beta 42; 150 1951 MCW 1951a

Wash (cell) cloth change, general vicinity: gamma, beta 17; 48 1951 MCW 1951a

Old material work in old cloth room: average gamma, beta 5; 20 1951 MCW 1951a 

Carrying, laundering Feinc, Leach, & Wash cloths, old cloth room: gamma, 

beta

3; 25 1951 MCW 1951a

Carrying, repairing Feinc blankets, old cloth room: gamma, beta 6; 30 1951 MCW 1951a

Washing Feinc blanket with brush on floor: carrying//changing -- gamma, beta 28; 105//28; 

150

1951 MCW 1951a

Changing strings, Feinc, Leach, and Wash filters in cells: typical gamma, beta 30;75 1951 MCW 1951a

Cloth storage room, gamma 0.3 - 2.0 1955 MCW 1957

Feinc filter plates (removed for cleaning, with some residue caked on them) <1-5; <10 1950 MCW 1950b

Floor surface under Feinc platform: 1 foot 30 (8) 1946 MCW 1946a

Floor around Feinc filters: 1 foot 110 (7) 1946 MCW 1946a

Feinc filtrate residue, contact (gamma; beta) >300; >500 1948 AEC 1948k

Operator position while filling drums of GLC: about 1 foot 110 (2) 1946 MCW 1946a

Skids of acid press cake (2-3 days old): 1 foot 200 (5) 1946 MCW 1946a

55-gal drum: 521 lbs of K-65 less than 24 hours old, side contact (gamma) 275 1949 MCW 1949g

GLC (3 days old), centered among 4 barrels; between 2 barrels 150; 120 1946 MCW 1946a

4 drums “aged” K-65, 6 feet --- 28 1947 AEC 1947c

2 drums “aged” K-65, 6 feet --- 23 1947 AEC 1947c

Sump recovery skids, all full (1 foot) 80 (5) 1946 MCW 1946a

Dempster body of Chem 6BC; Chem 6BP (1 foot) 30; 40 1946 MCW 1946a

Scalehouse sample room, sampling K-65: gamma; beta 2.0-4.0; 2.5-5.0 1955 MCW 1955n

Scalehouse sample room, carrying K-65 sample pans: gamma; beta 6.0; 12.5 1955 MCW 1955n

Sample Prep Room, pile of K-65 in hood: 1 foot 60 1946 MCW 1946a

Sample Prep Room, 3-gal bottle of K-65 at equilibrium; 3 2-qt bottles at equil 

(1 foot)

160; 100 1946 MCW 1946a

Sample Prep Room, operator position at Ro-Tap 50 1946 MCW 1946a

Sample Prep Room, operator position while milling GLC 130 1946 MCW 1946a

Sample Prep Room, near drying oven: 1 foot 20 1946 MCW 1946a

Laboratory sample room: center; vault 0; 80 1946 MCW 1946a

Analytical Lab, Dry Sample Prep Room, making up shotgun samples: gamma, 

beta

3; 1 1956 MCW 1956l

Shotgun Lab, scooping samples into vats: gamma, beta 3; 8 1956 MCW 1956l

Shotgun Lab, draining residue, 1st ether contact, max (NLO sample): gamma, 

beta

5; 8 1956 MCW 1956l
Document No. ORAUT-TKBS-0005 Revision No. 02 Effective Date: 06/14/2007 Page 207 of 225

Exposure,

% tolerance

Exposure, 

mrep/hr or 

mR/hr* Year Reference

Shotgun Lab, draining residue, later ether contacts, max: gamma, beta 6; 50 1956 MCW 1956l

Shotgun Lab, moving residue to hot plate, max: gamma, beta 6; 25 1956 MCW 1956l

Shotgun Lab, removing material from muffle furnace, max: gamma, beta 8; 30 1956 MCW 1956l

Shotgun Lab, removing material from mold, max: gamma, beta 8; 25 1956 MCW 1956l

Shotgun Lab, paint and place in box: gamma, beta 8; 10 1956 MCW 1956l

Shotgun Lab, counting samples: gamma, beta 2; 2 1956 MCW 1956l

Shotgun Lab, lab assay work 1.5; ND 1956 MCW 1956l

Shotgun Lab, general area work 1; ND 1956 MCW 1956l

Fork truck driver position while loading GLC(rail?)car 140 1946 MCW 1946a

UO2 packing area, cylindrical boxes containing UO2, cover on, beta 4.5 1948 Rochester 1948b

UO2 packing area, cylindrical boxes containing UO2, cover off, beta 22.5 1948 Rochester 1948b

UO2 packing area, inside 55-gal drum half-filled with UO2, beta 27 1948 Rochester 1948b

Warehouse, on 25 cylindrical boxes containing UO2, cover on, beta 15 1948 Rochester 1948b

PLANT 6E

Lid cleaner of the furnace enclosures: gamma, beta 170; 7500 1955 MCW 1955t

Crucible, large unpainted, contact: gamma, beta 5; 3000 1955 MCW 1955s

Crucible, large painted, contact: gamma, beta 4; 1000 1955 MCW 1955s

Crucible, small unpainted, contact: gamma, beta 4; 700 1955 MCW 1955s

Crucible, small painted, contact: gamma, beta 2; 560 1955 MCW 1955s

Painting used crucible: gamma, beta 1; 300 1955 MCW 1955s

Painting used crucible, body or arm: gamma, beta 2; 500 1956 MCW 1956k

Painting used crucible, hand: gamma, beta 4; 1500 1956 MCW 1956k

Assembling used crucible: gamma, beta 1; 70 1955 MCW 1955s

Assembling crucible, body at enclosure armhole: gamma, beta (mR/wk, 

mrep/wk)

36; 2500 1956 MCW 1956n

Assembling crucible, body: gamma, beta 2; 200 1956 MCW 1956k

Assembling crucible, arm: gamma, beta (in mR/week, mrep/week) 74; 17500 1956 MCW 1956n

Assembling crucible, arm: gamma, beta 6; 1800 1956 MCW 1956k

Assembling crucible, hand: gamma, beta (in mR/week, mrep/week) 78; 25000 1956 MCW 1956n

Assembling crucible, hand: gamma, beta 4; 1500 1956 MCW 1956k

General crucible assembly area: gamma, beta ND; 30 1955 MCW 1955s

General crucible assembly area work: gamma, beta 0.5; 3 1956 MCW 1956k

Crucible assembly operator, wrist: gamma, beta (mR/day, mrep/day) 50; 740 1956 MCW 1956m

Adjusting crucible jack: gamma, beta 1; 25 1956 MCW 1956k

Assembled used crucible, before loading: gamma, beta 1; 20 1955 MCW 1955s

Loading high-carbon/CST, max, body: gamma, beta 1; 20 1956 MCW 1956k

Loading high-carbon/CST, max, arm: gamma, beta 2; 120 1956 MCW 1956k

Loading high-carbon/CST, max, hand: gamma, beta 6; 300 1956 MCW 1956k

Loading crucible: gamma, beta 1; 20 1955 MCW 1955s

Lining up crucible: gamma, beta 1; 48 1955 MCW 1955s

Handling crucible lid: gamma, beta 1; 300 1955 MCW 1955s

Top insulation block: gamma, beta 1; 75 1955 MCW 1955s

Fixing stuck (crucible removal) sleeve: gamma, beta 2; 400 1955 MCW 1955s

General furnace enclosure area: gamma, beta 0.5; 8 1955 MCW 1955s

General furnace enclosure area work: gamma, beta 1; 100 1956 MCW 1956k

Crucible loader (top man), wrist: gamma, beta (mR/day, mrep/day) 0; 90 1956 MCW 1956m

Cleaning furnace enclosure, body at enc armhole: gamma, beta (mR/wk, 

mrep/wk)

23; 196 1956 MCW 1956n

Cleaning furnace enclosure, body at enclosure armhole: gamma, beta 1; 100 1956 MCW 1956k

Cleaning furnace enclosure, arm: gamma, beta (in mR/week, mrep/week) 23; 488 1956 MCW 1956n

Cleaning furnace enclosure, arm: gamma, beta 1; 300 1956 MCW 1956k

Cleaning furnace enclosure, hand: gamma, beta (in mR/week, mrep/week) 25; 628 1956 MCW 1956n

Cleaning furnace enclosure, hand: gamma, beta 1; 240 1956 MCW 1956k

Vacuum-cleaning burners: gamma, beta 1; 120 1955 MCW 1955s

Setting crucible over to cool: gamma, beta 1; 25 1955 MCW 1955s

Unlidding crucible: gamma, beta 1; 25 1955 MCW 1955s

Removing crucible lid, body: gamma, beta 3; 215 1956 MCW 1956k

Removing crucible lid, arm: gamma, beta (in mR/week, mrep/week) 22; 506 1956 MCW 1956n

Removing crucible lid, arm: gamma, beta 4; 1600 1956 MCW 1956k

Removing crucible lid, hand: gamma, beta (in mR/week, mrep/week) 23; 685 1956 MCW 1956n

Removing crucible lid, hand: gamma, beta 8; 2400 1956 MCW 1956k

Removing crucibles and slugs from enclosure: gamma, beta 2; 400 1955 MCW 1955s
Document No. ORAUT-TKBS-0005 Revision No. 02 Effective Date: 06/14/2007 Page 208 of 225

Exposure, 

% tolerance

Exposure, 

mrep/hr or 

mR/hr* Year Reference

General burnout area: gamma, beta 1; 8 1955 MCW 1955s

General burnout area work: gamma, beta 0.5; 3 1956 MCW 1956k

Gen’l burnout area work, body at enclo armhole: gamma, beta (mR/wk, 

mrep/wk)

22; 200 1956 MCW 1956n

Burnout operator, wrist: gamma, beta (mR/day, mrep/day) 90; 215 1956 MCW 1956m

SLAPS (all as “gamma; beta” except where indicated)

At contact with aged raffinate heaps 0.5-3.3;0.3-0.8 1949 MCW 1949g

Fresh raffinate heap, 1 foot 70; 1.6 1949 MCW 1949g

Fresh raffinate heap, chest height 33; 1.3 1949 MCW 1949g

Fresh raffinate heap, on bulldozer cab pushed into heap 2.7; 1.0 1949 MCW 1949g

“C” material heap, top of pile at waist height 11.5; 1.6 1949 MCW 1949g

“C” material heap, bulldozer cab pushed into heap (ratioed by raff bulldozer 

result)

7.7; 0.8 1949 MCW 1949g

Aged BC (barium sulfate cake) heap, top of pile at waist height 10; 3-10 1949 MCW 1949g

Aged BC heap bulldozer cab pushed into heap (ratioed by raff bulldozer result) 8.9; 0.13 1949 MCW 1949g

Fresh BC (barium sulfate cake) heap, top of pile at waist height 35; 2.5 1949 MCW 1949g

Fresh BC heap bulldozer cab pushed into heap (ratioed by raff bulldozer 

result)

2.5; 0.5 1949 MCW 1949g

K-65 storage shed, 2 feet (gamma only) 17.6 2.2 1948 MCW 1950f

K-65 storage shed, 3 feet (gamma only) 12.4 1.6 1948 MCW 1950f

K-65 storage shed, 6 feet (gamma only) 9.0 1.1 1948 MCW 1950f

K-65 storage shed, 9 feet (gamma only) 7.5 0.94 1948 MCW 1950f

K-65 storage shed, 12 feet (gamma only) 6.2 0.78 1948 MCW 1950f

K-65 storage shed, 15 feet (gamma only) 4.3 0.54 1948 MCW 1950f 

K-65 storage shed, 18 feet (gamma only) 3.6 0.45 1948 MCW 1950f

K-65 storage shed, 21 feet (gamma only) 2.6 0.33 1948 MCW 1950f

K-65 storage shed, 24 feet (gamma only) 1.8 0.23 1948 MCW 1950f 

K-65 storage shed, 27 feet (gamma only) 1.3 0.16 1948 MCW 1950f

K-65 storage shed, 30 feet (gamma only) 0.8 0.10 1948 MCW 1950f

K-65 storage shed, 150 feet (gamma only) -- guard shack location 0.032 0.004 1948 MCW 1950f

PLANT 4, measured beta dose rates analyzed for the design of Plant 6E

Meter reading, 

mrep/week

Worker film 

badge avg 

beta, mrep/wk Year Reference

Green salt, loading and unloading 816 480 1948 MCW 1948b

Green salt, box pulling 300 415 1948 MCW 1948b

Green salt, milling 432 530 1948 MCW 1948b

Green salt, extra man 300 510 1948 MCW 1948b

Green salt, lead operator 300 360 1948 MCW 1948b

Bomb step, jolting 600 440 1948 MCW 1948b

Bomb step, charging 600 370 1948 MCW 1948b

Bomb step, topping 600 380 1948 MCW 1948b

Bomb step, furnace tender 500 380 1948 MCW 1948b

Bomb step, unloading 600 460 1948 MCW 1948b

Bomb step, chipping 1900 845 1948 MCW 1948b

Bomb step, lime mixing --- 415 1948 MCW 1948b

Bomb step, magnesium mixing --- 300 1948 MCW 1948b

Recasting, furnace tending 500 665 1948 MCW 1948b

Recasting, top furnace tending 2125 1220 1948 MCW 1948b

Recasting, bottom furnace tending 1050 800 1948 MCW 1948b

Recasting, sawing >500 535 1948 MCW 1948b

Recasting, weighing and packing 613 580 1948 MCW 1948b

PLANT 6, gamma dose rates at area film badge monitoring locations

Meter reading, 

mR/hr

Average, 

mR/hr Year Reference

Digest area: upper platform 2.5 3.8 1955 MCW 1955c

Digest area: lower platform 3 5 1955 MCW 1955c

Cell: lower C-3 area (M-14, M-203); upper C-3 area (Olivers); leach area 

(Olivers)

2.5; 10; 3 2.6; 14.2; 3.1 1955 MCW 1955c

C-3: lower platform at M-214; upper platform 2; 2 3.8; 2.4 1955 MCW 1955c

Centrifuge area 2 3.4 1955 MCW 1955c

Feinc area: lower platform; upper platform; east area 2; 2; 4.5 3.1; 2.0; 6.5 1955 MCW 1955c

Cell: Feinc Niagara area 1.5 1.7 1955 MCW 1955c
Document No. ORAUT-TKBS-0005 Revision No. 02 Effective Date: 06/14/2007 Page 209 of 225

* The mrep/hr should be assumed for mixed beta-gamma or for beta alone; the mR/hr should be assumed for gamma only (i.e., either when 

stated explicitly or when it can be assumed, such as from the side of a tank or bottle). Slashes indicate multiple values. ND means not 

detectable.

Many figures are given in units of per cent of tolerance, defined in MCW (1949f) as “the amount of exposure that a person can receive 8 

hrs/day for an indefinite period of time,” presumably corresponding to 100 mR per eight-hour day for gammas and 500 mrep per eight-hour 

day for betas at the relevant time period. The exposure rates in mR/hr or in mrep/hr were in most cases converted from the original units of 

R per 8 hours. MED (1945b) gave beta exposure rate figures in “beta units,” which as MED (1945c) explained, were the same as “T units,”

with a T unit being 0.25 “r/hr” of beta radiation. This unit was taken to be 0.25 rep in the table above.

Parentheses indicate the number of measurements, if more than one, used to form the average reported.

Data from AEC 1949f were taken at waist height except where indicated.
Document No. ORAUT-TKBS-0005 Revision No. 02 PC-1 Effective Date: 05/25/2009 Page 210 of 225

Table A-34. Neutron dose rates and doses from the alpha-neutron reaction sources and from a RaBe 

source. 

Alpha-neutron source information and dose rates

Form Source

Target 

element(s)

Weight in 

container

Dose rate at 

1 foot, rem/hr

Dose rate at 

3 feet, rem/hr

Annual dose, 

rem

U3O8 U natural mix O 100 lbs 1.29E-07 1.44E-08 4.31E-05

U natural mix + daughters O 100 lbs 1.10E-06 1.22E-07 3.67E-04

UO3, UO2 U natural mix O 75 lbs 9.71E-08 1.08E-08 3.24E-05

U natural mix + daughters O 75 lbs 8.27E-07 9.19E-08 2.76E-04

UF4 U natural mix F 75 lbs 6.92E-06 7.69E-07 2.31E-03

U natural mix + daughters F 75 lbs 5.91E-05 6.57E-06 1.97E-02

Na2U2O7

(soda salt)

U natural mix

O, Na 75 lbs 2.61E-07 2.90E-08 8.70E-05

U natural mix + daughters O, Na 75 lbs 2.13E-06 2.37E-07 7.11E-04

ThF4 Th-230/Th232 mix F 198 g 4.08E-04 4.53E-05 1.36E-01

Th-230/Th-232 mix + 

daughters F 198 g 8.97E-04 9.96E-05 2.99E-01

Th(NO3)4 Th-230/Th232 mix O 309 g 4.74E-06 5.26E-07 1.58E-03

Th-230/Th-232 mix + 

daughters O 309 g 1.04E-05 1.16E-06 3.47E-03

RaBe source information and dose rates

RaBe Ra-226 --- 100 mg Ra 1.60E-02 1.77E-03 4.44E+00

Annual organ doses (based on ThF4 for the alpha-neutron case; Ra-226 for the RaBe case)

Organ

Alpha-neutron 

dose, rem

RaBe dose, 

rem

Bladder 2.09E-01 5.00E+00

Bone marrow 1.16E-01 3.37E+00

Bone surface 1.34E-01 3.15E+00

Breast 3.10E-01 5.08E+00

Colon 1.34E-01 4.17E+00

Esophagus 1.17E-01 4.02E+00

Lung 1.62E-01 4.32E+00

Gonads (ovaries) 1.19E-01 4.15E+00

Gonads (testes) 3.39E-01 5.50E+00

Liver 1.78E-01 4.54E+00

Remainder organs 2.57E-01 4.13E+00

Skin 9.64E-09 4.33E+00

Stomach 2.20E-01 4.96E+00

Thyroid 2.95E-01 5.15E+00

Data used to calculate the alpha-neutron dose rates (except for the assumed amount of U, Th, or Ra material) is from 

Shleien (1992), Tables 8.4.1, 7.3, and 7.2; Salmon and Hermann (1993), Table 3; DOE (2001a), Table 6-5; Foderaro (1978); 

and Mlekodaj (2002). Data used to calculate the RaBe dose rates are from Shleien (1992), Tables 7.5 and 8.4.1, and 

Foderaro (1978). In both cases, the dose conversion factors were taken from NIOSH (2007). The quantities listed 

correspond to one drum of material for the uranium forms; to one day’s processing for the thorium forms (on a 70-day basis); 

and to the known RaBe source for the Ra form.

In the alpha-neutron case, the neutron energy was taken as 1.5 MeV because it is the approximate maximum energy for Th232, the other isotopes also emit neutrons in the range 1.0-2.0, and the flux-to-dose conversion factor varies slowly in this 

range. For the annual dose, it was assumed that 1 hour /day was spent at 1 foot from the container and three hours/day 

spent at 3 feet, every working day for a year. In the RaBe case, the neutron energy was taken as 4.0 MeV and it was 

assumed that 1 hour/day was spent at 1 foot from the unshielded source and 1 hour/day spent at 3 feet, every working day 

for a year. (This is conservative since the source was likely to be at least partially shielded when in use and completely 

shielded when not in use.) 

In both cases, the maximum organ dose was based on the maximum annual whole-body dose taken to be the ambient dose. 

For the alpha-neutron source case, an exposure geometry of 75% AP/25% ROT was assumed, as per Table A-18 (for 

“ionium plant operator”); for the RaBe case, 90% AP/10% ROT was assumed, as per Table A-18 (for “Shotgun Lab analyst”).
Document No. ORAUT-TKBS-0005 Revision No. 02 Effective Date: 06/14/2007 Page 211 of 225

Table A-35. External gamma dose rates from processing AM-7 residue to concentrate 

thorium. 

Radionuclide

Gamma energy

MeV Freq Ci R/h@1 ft Annual Dose, R

Th-230 0.142 0.0007 24.4 2.08E-04 6.93E-02

Th-230 0.143 0.00045 24.4 1.35E-04 4.49E-02

Th-232 0.126 0.0004 0.001 4.32E-09 1.44E-06

Ra-226 0.186 0.0328 0.158 8.26E-05 2.75E-02

Total --- --- 48.959 4.25E-04 1.42E-01

Ratio, Ra/Th --- 3.24E-03 2.41E-01 2.41E-01

Ratio, Ra/Total --- 3.23E-03 1.94E-01 1.94E-01

Note: Assumptions for the annual dose are as for Table A-34, i.e., 1 hours per day were spent at one foot 

from the unshielded source and 3 hours per day at 3 ft for 250 working days.

Table A-36. Reserved. 
Document No. ORAUT-TKBS-0005 Revision No. 02 Effective Date: 06/14/2007 Page 212 of 225

Table A-37. Monitoring experience from the decontamination/demolition period, 1960 -1961 ( MCW 

1961b).

General outside air concentration during decontamination and demolition

Airborne activity, in units of 10-13 uCi/cc

Date sampled Uranium Gross alpha No. samples

Dec 1960 4.0 6.8 16

Jan 1961 7.4 6.5 48

Feb 1961 1.1 1.6 104

Mar 1961 1.4 1.7 141

Apr 1961 2.0 6.9 115

May 1961 2.0 4.0 97

Jun 1961 3.3 5.6 69

Jul 1961 1.6 2.8 48

Aug 1961 1.9 4.1 77

Sep 1961 2.2 2.9 83

Oct 1961 3.8 5.0 77

Nov 1961 12.0 6.7 21

Weighted average 2.6 4.0 ---

Operational area air concentrations, as airborne alpha activity in uCi/cc

Operation or Area Average High

Cutting and removing iron structures, equipment 8.7 E-11 7.7 E-10

Loading iron, equipment by hand and by lift 8.7 E-11 5.3 E-10

Demolishing wood, concrete, or masonry 2.7 E-11 9.2 E-11

Mechanically loading debris and rubble 1.8 E-11 1.1 E-10

Removing tar and gravel roofs 2.5 E-11 1.5 E-11

Dry-sandblasting the insides of buildings 3.6 E-10 7.4 E-10

General area outside blogs during inside sandblasting 2.7 E-11 9.5 E-11

Dry-sandblasting outside surfaces 4.2 E-11 1.1 E-10

Shoveling residues into small quarters 2.4 E-10 ---

Jack hammering Unite from tanks 8.2 E-10 1.8 E-09

Outside hood during removal of Unite inside hood 3.4 E-12 ---

Cleaning sand, residues from inside vessel after sandblasting 6.4 E-11 ---

Steam cleaning of no salvageable materials 2.0 E-12 ---

Source sampling with wire brush inside pipe after solvent soaking 8.8 E-10 ---

Stack exhaust area, from dust collector-hood setup 3.3 E-11 ---

General air in building during period of inactivity 1.4 E-11 ---

Concentrations of uranium in urine, in mg/L

Sample type Average Range

Equipment removal 

phase

Pre-exposure sample 0.007 0.002 - 0.013

Monday before work sample 0.010 0.005 - 0.008

Friday after work sample 0.025 0.003 - 0.078

Decontamination and 

demolition

Pre-exposure sample 0.005 0.001 - 0.014

Monday before work sample 0.019 0.003 - 0.058

Friday after work sample 0.036 0.006 - 0.271

Termination sample 0.017 0.006 - 0.027

Special operations Sandblasting 0.031 0.007 - 0.080

Torch-cutting stainless steel 0.032 ---

Cutting, wrecking a rod mill 0.098 0.061 - 0.135

Dismantling vacuum, dust collector 0.065 ---

Stripping wiring, metal from blogs 0.045 0.007 - 0.158

Film badge data

Gamma Beta

User

Annual 

average, rem

Highest quarterly 

reading, rem

Highest 

monthly 

reading, rem

Annual average, 

rem

Highest 

monthly 

reading, rem

Contractor 0.04 0.15 0.08 0.3 0.29

AEC 0.1 --- --- 0.6 ---

MCW 0.2 --- --- 0.8 ---

AEC/MCW --- --- 0.14 --- 0.15

Visitor --- --- 0.05 --- ---
Document No. ORAUT-TKBS-0005 Revision No. 02 Effective Date: 06/14/2007 Page 213 of 225

Table A-38. Summary of post-decontamination exposure data as measured by two survey groups. 

Plant

Bldg 

Remaining

1981 1986 1990 Overall averages, 1981

Radon 

daughters, 

WL

Radon Yearly 

Avgas, 

WLM/yr

Radon 

Yearly 

Avgas, 

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 phi/L 0.4 – 37 phi/L, avgas

daytime; 69 phi/L max

25-1 0.001 0.01 Radon daughters 

in air

<0.01 WL 0.0009-0.07 WL, avgas

daytime concentration

25-2 0.0009 0.01 Gamma from Ra, 

U daughters

8 or/hr 8-290 or /yr @ 1 m above 

floor

38

40

2 50 0.0003

51 0.0005 0.02 0.01

51A 0.02 0.01

52A 0.07

52 0.0007-

0.001

0.00

6 100 0.02

101 0.10 0.01

6E 116-1 0.00 0.04

116-2 0.00

116B 0.01

117-1 0.00

117-2 0.00

7 700 0.00

704 0.03 0.00

705 0.00

706 0.00

708 0.01

Measurements are from ORNL (1981) and Applied Nuclear Safety (1986; 1990).
Document No. ORAUT-TKBS-0005 Revision No. 02 Effective Date: 06/14/2007 Page 214 of 225

Table A-39. Measured data used to produce source terms for decontamination and postdecontamination exposure calculations. Internal exposure External exposure

Plant Case

Dose 

potential

Surface alpha, 

dpm/100 cm2

Volume 

alpha, phi/g

Maximum 

average 

gamma, mR/hr

Typical 

average 

gamma, mR/hr

Maximum 

average beta, 

prep/hr

Typical 

average beta, 

prep/hr

4 Decon Low 4.4E+03 1.0E+04 8.0E-02 2.4E-02 1.3E+01 7.5E+00

Moderate 8.3E+03 4.2E+04 3.0E-01 9.0E-02 2.4E+01 1.4E+01

High 1.7E+04 8.1E+04 2.0E+00 6.0E-01 4.7E+01 2.7E+01

6 Decon Low 1.2E+02 1.0E+01 1.0E-01 1.2E-02 1.0E+00 1.2E-01

Moderate 3.9E+03 7.7E+02 2.5E-01 2.9E-02 2.0E+00 2.4E-01

High 6.0E+04 3.0E+03 3.0E+00 1.7E+00 3.9E+01 4.6E+00

6E Decon Low 4.0E+02 0.0E00 1.0E-01 5.0E-02 2.0E-01 4.3E-02

Moderate 2.7E+03 3.8E+02 5.0E-01 1.0E-01 1.3E+00 2.8E-01

High 3.5E+04 1.6E+03 1.4E+00 5.0E-01 1.2E+01 2.5E+00

7 Decon Low 8.5E+01 0.0E+00 5.0E-02 2.0E-02 1.0E-01 1.7E-02

Moderate 2.8E+03 1.9E+02 1.0E-01 4.0E-02 1.7E+00 2.9E-01

High 1.2E+05 2.6E+02 2.5E-01 1.0E-01 1.8E+01 3.0E+00

1 Post-deacon --- --- --- 8.7E-02 2.1E-02 8.9E+00 9.0E-01

2 Post-deacon --- --- --- 2.1E-02 1.0E-02 9.7E+00 2.3E-01

6E Post-deacon Low 1.2E+03 0.0E+00 5.4E-02 1.7E-02 5.0E-01 1.0E-01

Moderate 3.0E+03 3.8E+02 8.1E-02 4.2E-02 7.5E-01 2.5E-01

High 4.5E+03 1.6E+03 1.6E-01 8.3E-02 1.5E+00 5.0E-01

7 Post-deacon Low 1.0E+03 0.0E+00 8.1E-03 3.2E-03 2.5E-01 1.0E-01

Moderate 2.8E+03 1.9E+02 1.6E-02 8.1E-03 5.0E-01 2.5E-01

 High 7.5E+03 2.6E+02 3.2E-02 1.6E-02 1.0E+00 5.0E-01

The data in this table was extracted from MCW (1958), MCW(1959), MCW (1961a), and ORNL (1981). 

For the Plants 6E and 7 post-decontamination dose rates, the beta/gamma ratio was assumed to be the same as for the corresponding 

decontamination maximum case. For both decontamination and post-decontamination cases, the High maximum was taken as the maximum 

average or representative maximum for the respective plant; the High typical value was the average of the non-zero average levels, hence is 

conservative. Similarly, the Moderate maximum and the Low maximum were the respective middling and low averages. The Moderate and 

Low typical values were rationed from the High typical values.

For the Plants 1 and 2 case, the maximum averages were derived from the highest representative measurements in each building and the 

maximum building in each plant was chosen to be representative of all buildings in that plant. Also for Plants 1 and 2, the measured betagamma dose rate was assumed to be all beta, for conservatism and in default of other information; the gamma measurements are based on 

true gamma-only readings taken with a different instrument than was used to measure the beta-gamma level. The beta-gamma 

measurements were taken at less than 1 inch (the MCW references) or at 1 cm (ORNL 1981), while the gamma measurements were 

generally at 3 feet (1 m).
Document No. ORAUT-TKBS-0005 Revision No. 02 PC-1 Effective Date: 05/25/2009 Page 215 of 225

Table A-40. Annual internal exposures during the decontamination and post-decontamination periods. 

Decontamination Annual average exposure, 1959-1961

Plant 4 Plant 6 Plant 6E Plant 7

Inhalation, phi

Low 8.39E+02 1.98E+01 6.86E+01 1.46E+01

Moderate 1.77E+03 6.67E+02 4.66E+02 4.82E+02

High 3.48E+03 1.03E+04 6.02E+03 1.97E+04

Radon, WLM

Low 1.43E-02 3.73E-04 1.30E-03 2.76E-04

Moderate 2.69E-02 1.25E-02 8.76E-03 9.09E-03

High 5.36E-02 1.95E-01 1.14E-01 3.73E-01

Ingestion, phi

Low 1.75E+01 4.13E-01 1.43E+00 3.04E-01

Moderate 3.68E+01 1.39E+01 9.72E+00 1.00E+01

High 7.25E+01 2.15E+02 1.25E+02 4.11E+02

Post-decontamination

Annual average exposure, 1962-1995 (1959-1995 for Plants 1 and 2)

Plant 1 Plant 2

1959-1965 1966-1975 1976-1985 1986-1995 1959-1965 1966-1975 1976-1985 1986-1995

Inhalation, phi 3.77E+01 3.78E+01 3.80E+01 3.77E+01 3.77E+01 3.78E+01 3.80E+01 3.77E+01

Radon, WLM 4.25E-01 4.17E-01 4.04E-01 3.92E-01 1.42E+00 1.39E+00 1.35E+00 1.31E+00

Ingestion, pCi 7.84E-01 7.87E-01 7.92E-01 7.85E-01 7.84E-01 7.87E-01 7.92E-01 7.85E-01

Post-decontamination Plant 6E Plant 7

1959-1965 1966-1975 1976-1985 1986-1995 1959-1965 1966-1975 1976-1985 1986-1995

Inhalation, pCi

Low 9.12E+00 5.72E+00 5.75E+00 5.70E+00 4.75E+00 4.77E+00 4.79E+00 4.75E+00

Moderate 1.71E+01 1.72E+01 1.73E+01 1.72E+01 1.47E+01 1.48E+01 1.49E+01 1.48E+01

High 3.35E+01 3.36E+01 3.37E+01 3.36E+01 3.76E+01 3.77E+01 3.79E+01 3.76E+01

Radon, WLM

Low 4.54E-03 4.45E-03 4.31E-03 4.18E-03 3.78E-03 3.71E-03 3.60E-03 3.48E-03

Moderate 1.13E-02 1.11E-02 1.08E-02 1.04E-02 1.06E-02 1.04E-02 1.01E-02 9.75E-03

High 1.70E-02 1.67E-02 1.62E-02 1.57E-02 2.84E-02 2.78E-02 2.70E-02 2.61E-02

Ingestion, pCi

Low 1.90E-01 1.19E-01 1.20E-01 1.19E-01 9.90E-02 9.93E-02 9.99E-02 9.90E-02

Moderate 3.57E-01 3.58E-01 3.60E-01 3.57E-01 3.07E-01 3.08E-01 3.10E-01 3.07E-01

High 6.97E-01 6.99E-01 7.03E-01 7.00E-01 7.83E-01 7.86E-01 7.90E-01 7.84E-01

Inhalation and radon data in this table was calculated with the RESRAD-BUILD computer code (ANL 2003), using measured data from MCW 

(1958; 1959; 1961a). Both surface contamination and bulk floor and wall contamination were taken into account in calculating the inhalation 

and radon contributions. Ingestion data were calculated using the ingestion model of OCAS (2004).

“High” exposure potential represents those working in the most contaminated areas, i.e., the former process areas; “Moderate”, those 

accessing less contaminated areas or infrequently accessing former process areas; and “Low”, those accessing uncontaminated areas. 

Plants 1 and 2 did not have enough data for these classifications to be applied.

The Plants 1 and 2 inhalation and ingestion figures were taken to be the same as the Plant 7 figures since the post-D&D contamination 

figures as given in ORNL (1981) for them were similar to the data for Plant 7. ORNL (1981) showed that while most radon daughter readings 

were similar in all of the plants, there was one significantly higher reading in both Plant 1 and Plant 2. Thus the Plants 1 and 2 readings are 

the Plant 7 readings multiplied by the factor of difference for the respective buildings (15 for Plant 1 and 50 for Plant 2). This should be quite 

conservative since the measured Plant 1 and 2 readings were maxima. 

Intake quantities listed above in Table A-40 (both inhalation and ingestion) represent total alpha activity. Isotopic distribution should be 

based on either the likely source term or selected to be favorable to claimants based on the tabulation below (total alpha source term factors). 

For areas in which only uranium ores were present, the intake can be based on the isotopic distribution of uranium in equilibrium with its 

progeny (shown below). If the work area cannot be determined, or if exposure to uranium raffinates was likely, then the total alpha activity 

can be distributed as 94.6% Th-230 and 5.4% Ac-227. This distribution is based on the isotopic composition of the total alpha activity of the 

ionium product which was determined to be 94.6% th-230 and 5.4% th-227 (Ac-227 progeny) (Kuhlman 1955) and results in an internal dose 

that is bounding of that which would be calculated based on the possible raffinate mixtures potentially present.

Total Alpha Source Term Factors – uranium ore.

Source Uranium in equilibrium

Ratio to Uranium

Fraction of Total 

Gross Alphaa 

Uranium 1 0.4

Th-230 0.49 0.2

Ra-226 0.49 0.2

Pb-210 0.49 0.2

Po-210 0.49 0.2

Pa-231 0.02 0.01

Ac-227 0.02 0.01 a Total alpha fraction based on ratio of each nuclide activity to the total alpha emitting fraction (Uranium + Th-230 + Ra-226 + Po-210 + Pa231) and assuming that the Pb-210 activity is equal to that of Po-210 and the Ac-227 activity is equal to that of Pa-231.
Document No. ORAUT-TKBS-0005 Revision No. 02 Effective Date: 06/14/2007 Page 216 of 225

Table A-41. Annual external photon organ doses, decontamination and post-decontamination periods 

(rem per year). Decontamination years (1959-1961) 

Plant 4 Plant 6 Plant 6E Plant 7

Organ Type Low Moderate High Low Moderate High Low Moderate High Low Moderate High

Bladder Min 5.6E-02 2.1E-01 1.4E+00 4.9E-02 1.2E-01 3.0E+00 9.2E-02 2.9E-01 1.1E+00 4.0E-02 8.1E-02 2.0E-01

Mode 1.1E-01 4.1E-01 2.7E+00 9.7E-02 2.4E-01 5.8E+00 1.8E-01 5.7E-01 2.1E+00 7.9E-02 1.6E-01 4.0E-01

Max 2.9E-01 1.1E+00 7.2E+00 2.5E-01 6.4E-01 1.5E+01 4.7E-01 1.5E+00 5.5E+00 2.1E-01 4.2E-01 1.0E+00

Bone (red 

marrow)

Min 3.8E-02 1.4E-01 9.5E-01 3.4E-02 8.4E-02 2.0E+00 6.2E-02 2.0E-01 7.2E-01 2.7E-02 5.5E-02 1.4E-01

Mode 7.4E-02 2.8E-01 1.9E+00 6.6E-02 1.6E-01 4.0E+00 1.2E-01 3.9E-01 1.4E+00 5.4E-02 1.1E-01 2.7E-01

Max 2.0E-01 7.3E-01 4.9E+00 1.7E-01 4.3E-01 1.0E+01 3.2E-01 1.0E+00 3.7E+00 1.4E-01 2.8E-01 7.1E-01

Bone 

(surface)

Min 6.7E-02 2.5E-01 1.7E+00 6.0E-02 1.5E-01 3.6E+00 1.1E-01 3.5E-01 1.3E+00 4.9E-02 9.7E-02 2.4E-01

Mode 1.3E-01 4.9E-01 3.3E+00 1.2E-01 2.9E-01 7.0E+00 2.2E-01 6.9E-01 2.5E+00 9.5E-02 1.9E-01 4.8E-01

Max 3.5E-01 1.3E+00 8.6E+00 3.1E-01 7.7E-01 1.8E+01 5.7E-01 1.8E+00 6.6E+00 2.5E-01 5.0E-01 1.3E+00

Breast 

(female)

Min 5.9E-02 2.2E-01 1.5E+00 5.2E-02 1.3E-01 3.1E+00 9.7E-02 3.1E-01 1.1E+00 4.3E-02 8.5E-02 2.1E-01

Mode 1.2E-01 4.3E-01 2.9E+00 1.0E-01 2.6E-01 6.1E+00 1.9E-01 6.1E-01 2.2E+00 8.3E-02 1.7E-01 4.2E-01

Max 3.0E-01 1.1E+00 7.6E+00 2.7E-01 6.7E-01 1.6E+01 5.0E-01 1.6E+00 5.8E+00 2.2E-01 4.4E-01 1.1E+00

Colon Min 5.0E-02 1.9E-01 1.3E+00 4.5E-02 1.1E-01 2.7E+00 8.3E-02 2.7E-01 9.6E-01 3.6E-02 7.3E-02 1.8E-01

Mode 9.9E-02 3.7E-01 2.5E+00 8.7E-02 2.2E-01 5.3E+00 1.6E-01 5.2E-01 1.9E+00 7.1E-02 1.4E-01 3.6E-01

Max 2.6E-01 9.7E-01 6.5E+00 2.3E-01 5.7E-01 1.4E+01 4.3E-01 1.4E+00 4.9E+00 1.9E-01 3.7E-01 9.4E-01

Esophagus Min 3.9E-02 1.5E-01 9.8E-01 3.5E-02 8.7E-02 2.1E+00 6.4E-02 2.1E-01 7.5E-01 2.8E-02 5.7E-02 1.4E-01

Mode 7.7E-02 2.9E-01 1.9E+00 6.8E-02 1.7E-01 4.1E+00 1.3E-01 4.0E-01 1.5E+00 5.5E-02 1.1E-01 2.8E-01

Max 2.0E-01 7.5E-01 5.0E+00 1.8E-01 4.5E-01 1.1E+01 3.3E-01 1.1E+00 3.8E+00 1.5E-01 2.9E-01 7.3E-01

Eye Min 6.0E-02 2.3E-01 1.5E+00 5.3E-02 1.3E-01 3.2E+00 9.9E-02 3.2E-01 1.1E+00 4.4E-02 8.7E-02 2.2E-01

Mode 1.2E-01 4.4E-01 2.9E+00 1.0E-01 2.6E-01 6.3E+00 1.9E-01 6.2E-01 2.3E+00 8.5E-02 1.7E-01 4.3E-01

Max 3.1E-01 1.2E+00 7.7E+00 2.7E-01 6.9E-01 1.7E+01 5.1E-01 1.6E+00 5.9E+00 2.2E-01 4.5E-01 1.1E+00

Gonads 

(ovaries)

Min 4.8E-02 1.8E-01 1.2E+00 4.2E-02 1.1E-01 2.5E+00 7.9E-02 2.5E-01 9.1E-01 3.5E-02 6.9E-02 1.7E-01

Mode 9.4E-02 3.5E-01 2.3E+00 8.3E-02 2.1E-01 5.0E+00 1.5E-01 4.9E-01 1.8E+00 6.8E-02 1.4E-01 3.4E-01

Max 2.5E-01 9.2E-01 6.1E+00 2.2E-01 5.4E-01 1.3E+01 4.0E-01 1.3E+00 4.7E+00 1.8E-01 3.6E-01 8.9E-01

Gonads 

(testes)

Min 6.3E-02 2.3E-01 1.6E+00 5.5E-02 1.4E-01 3.3E+00 1.0E-01 3.3E-01 1.2E+00 4.5E-02 9.1E-02 2.3E-01

Mode 1.2E-01 4.6E-01 3.1E+00 1.1E-01 2.7E-01 6.5E+00 2.0E-01 6.5E-01 2.3E+00 8.9E-02 1.8E-01 4.4E-01

Max 3.2E-01 1.2E+00 8.0E+00 2.9E-01 7.1E-01 1.7E+01 5.3E-01 1.7E+00 6.1E+00 2.3E-01 4.7E-01 1.2E+00

Liver Min 5.2E-02 1.9E-01 1.3E+00 4.6E-02 1.1E-01 2.8E+00 8.5E-02 2.7E-01 9.9E-01 3.7E-02 7.5E-02 1.9E-01

Mode 1.0E-01 3.8E-01 2.5E+00 9.0E-02 2.2E-01 5.4E+00 1.7E-01 5.3E-01 1.9E+00 7.3E-02 1.5E-01 3.7E-01

Max 2.7E-01 1.0E+00 6.6E+00 2.4E-01 5.9E-01 1.4E+01 4.4E-01 1.4E+00 5.1E+00 1.9E-01 3.8E-01 9.6E-01

Lung Min 5.1E-02 1.9E-01 1.3E+00 4.5E-02 1.1E-01 2.7E+00 8.4E-02 2.7E-01 9.7E-01 3.7E-02 7.4E-02 1.8E-01

Mode 1.0E-01 3.8E-01 2.5E+00 8.9E-02 2.2E-01 5.3E+00 1.6E-01 5.3E-01 1.9E+00 7.2E-02 1.4E-01 3.6E-01

Max 2.6E-01 9.8E-01 6.6E+00 2.3E-01 5.8E-01 1.4E+01 4.3E-01 1.4E+00 5.0E+00 1.9E-01 3.8E-01 9.5E-01

Remainder 

organs

Min 4.6E-02 1.7E-01 1.1E+00 4.1E-02 1.0E-01 2.4E+00 7.5E-02 2.4E-01 8.7E-01 3.3E-02 6.6E-02 1.7E-01

Mode 9.0E-02 3.4E-01 2.2E+00 7.9E-02 2.0E-01 4.8E+00 1.5E-01 4.7E-01 1.7E+00 6.5E-02 1.3E-01 3.2E-01

Max 2.4E-01 8.8E-01 5.9E+00 2.1E-01 5.2E-01 1.3E+01 3.9E-01 1.2E+00 4.5E+00 1.7E-01 3.4E-01 8.5E-01

Skin Min 4.9E-02 1.8E-01 1.2E+00 4.3E-02 1.1E-01 2.6E+00 8.0E-02 2.6E-01 9.3E-01 3.5E-02 7.1E-02 1.8E-01

Mode 9.6E-02 3.6E-01 2.4E+00 8.5E-02 2.1E-01 5.1E+00 1.6E-01 5.0E-01 1.8E+00 6.9E-02 1.4E-01 3.5E-01

Max 2.5E-01 9.4E-01 6.3E+00 2.2E-01 5.6E-01 1.3E+01 4.1E-01 1.3E+00 4.8E+00 1.8E-01 3.6E-01 9.1E-01

Stomach Min 5.7E-02 2.1E-01 1.4E+00 5.1E-02 1.3E-01 3.0E+00 9.4E-02 3.0E-01 1.1E+00 4.1E-02 8.3E-02 2.1E-01

Mode 1.1E-01 4.2E-01 2.8E+00 9.9E-02 2.5E-01 6.0E+00 1.8E-01 5.9E-01 2.1E+00 8.1E-02 1.6E-01 4.0E-01

Max 2.9E-01 1.1E+00 7.3E+00 2.6E-01 6.5E-01 1.6E+01 4.8E-01 1.5E+00 5.6E+00 2.1E-01 4.2E-01 1.1E+00

Thymus Min 6.2E-02 2.3E-01 1.6E+00 5.5E-02 1.4E-01 3.3E+00 1.0E-01 3.3E-01 1.2E+00 4.5E-02 9.0E-02 2.2E-01

Mode 1.2E-01 4.6E-01 3.0E+00 1.1E-01 2.7E-01 6.5E+00 2.0E-01 6.4E-01 2.3E+00 8.8E-02 1.8E-01 4.4E-01

Max 3.2E-01 1.2E+00 8.0E+00 2.8E-01 7.1E-01 1.7E+01 5.2E-01 1.7E+00 6.1E+00 2.3E-01 4.6E-01 1.2E+00

Thyroid Min 6.9E-02 2.6E-01 1.7E+00 6.1E-02 1.5E-01 3.7E+00 1.1E-01 3.7E-01 1.3E+00 5.0E-02 1.0E-01 2.5E-01

Mode 1.4E-01 5.1E-01 3.4E+00 1.2E-01 3.0E-01 7.2E+00 2.2E-01 7.2E-01 2.6E+00 9.8E-02 2.0E-01 4.9E-01

Max 3.6E-01 1.3E+00 8.9E+00 3.2E-01 7.9E-01 1.9E+01 5.9E-01 1.9E+00 6.8E+00 2.6E-01 5.2E-01 1.3E+00

Uterus Min 4.8E-02 1.8E-01 1.2E+00 4.2E-02 1.1E-01 2.6E+00 7.9E-02 2.5E-01 9.1E-01 3.5E-02 6.9E-02 1.7E-01

Mode 9.4E-02 3.5E-01 2.3E+00 8.3E-02 2.1E-01 5.0E+00 1.5E-01 4.9E-01 1.8E+00 6.8E-02 1.4E-01 3.4E-01

Max 2.5E-01 9.2E-01 6.2E+00 2.2E-01 5.5E-01 1.3E+01 4.1E-01 1.3E+00 4.7E+00 1.8E-01 3.6E-01 8.9E-01

Whole-body 

dose rate

Min 7.6E-02 2.9E-01 1.9E+00 6.7E-02 1.7E-01 4.1E+00 1.3E-01 4.0E-01 1.5E+00 5.5E-02 1.1E-01 2.8E-01

Mode 1.5E-01 5.6E-01 3.7E+00 1.3E-01 3.3E-01 7.9E+00 2.4E-01 7.8E-01 2.8E+00 1.1E-01 2.2E-01 5.4E-01

Max 3.9E-01 1.5E+00 9.8E+00 3.5E-01 8.7E-01 2.1E+01 6.4E-01 2.1E+00 7.5E+00 2.8E-01 5.7E-01 1.4E+00

Post-decontamination years (1962-1995)

Plant 1 Plant 2 Plant 6E Plant 7

Organ Type Low Moderate High Low Moderate High Low Moderate High Low Moderate High

Bladder Min --- --- 5.5E-02 --- --- 1.9E-02 3.8E-02 7.6E-02 1.5E-01 6.5E-03 1.5E-02 3.0E-02

Mode --- --- 1.1E-01 --- --- 3.7E-02 7.5E-02 1.5E-01 3.0E-01 1.3E-02 2.9E-02 5.8E-02

Max --- --- 2.8E-01 --- --- 9.6E-02 2.0E-01 3.9E-01 7.8E-01 3.3E-02 7.6E-02 1.5E-01

Bone (red 

marrow)

Min --- --- 3.7E-02 --- --- 1.3E-02 2.6E-02 5.1E-02 1.0E-01 4.4E-03 1.0E-02 2.0E-02

Mode --- --- 7.3E-02 --- --- 2.5E-02 5.1E-02 1.0E-01 2.0E-01 8.7E-03 2.0E-02 3.9E-02

Max --- --- 1.9E-01 --- --- 6.5E-02 1.3E-01 2.6E-01 5.3E-01 2.3E-02 5.2E-02 1.0E-01
Document No. ORAUT-TKBS-0005 Revision No. 02 PC-1 Effective Date: 05/25/2009 Page 217 of 225

Post-decontamination years (1962-1995)

Plant 1 Plant 2 Plant 6E Plant 7

Organ Type Low Moderate High Low Moderate High Low Moderate High Low Moderate High

Bone 

(surface)

Min --- --- 6.6E-02 --- --- 2.3E-02 4.6E-02 9.1E-02 1.8E-01 7.8E-03 1.8E-02 3.6E-02

Mode --- --- 1.3E-01 --- --- 4.4E-02 9.0E-02 1.8E-01 3.6E-01 1.5E-02 3.5E-02 7.0E-02

Max --- --- 3.4E-01 --- --- 1.2E-01 2.4E-01 4.7E-01 9.4E-01 4.0E-02 9.2E-02 1.8E-01

Breast 

(female)

Min --- --- 5.8E-02 --- --- 2.0E-02 4.0E-02 8.0E-02 1.6E-01 6.9E-03 1.6E-02 3.1E-02

Mode --- --- 1.1E-01 --- --- 3.9E-02 7.9E-02 1.6E-01 3.1E-01 1.3E-02 3.1E-02 6.1E-02

Max --- --- 3.0E-01 --- --- 1.0E-01 2.1E-01 4.1E-01 8.2E-01 3.5E-02 8.0E-02 1.6E-01

Colon Min --- --- 5.0E-02 --- --- 1.7E-02 3.4E-02 6.8E-02 1.4E-01 5.9E-03 1.3E-02 2.7E-02

Mode --- --- 9.7E-02 --- --- 3.3E-02 6.7E-02 1.3E-01 2.7E-01 1.2E-02 2.6E-02 5.2E-02

Max --- --- 2.6E-01 --- --- 8.7E-02 1.8E-01 3.5E-01 7.0E-01 3.0E-02 6.9E-02 1.4E-01

Esophagus Min --- --- 3.9E-02 --- --- 1.3E-02 2.7E-02 5.3E-02 1.1E-01 4.6E-03 1.0E-02 2.1E-02

Mode --- --- 7.6E-02 --- --- 2.6E-02 5.2E-02 1.0E-01 2.1E-01 8.9E-03 2.0E-02 4.1E-02

Max --- --- 2.0E-01 --- --- 6.7E-02 1.4E-01 2.7E-01 5.4E-01 2.3E-02 5.3E-02 1.1E-01

Eye Min --- --- 5.9E-02 --- --- 2.0E-02 4.1E-02 8.2E-02 1.6E-01 7.0E-03 1.6E-02 3.2E-02

Mode --- --- 1.2E-01 --- --- 4.0E-02 8.1E-02 1.6E-01 3.2E-01 1.4E-02 3.1E-02 6.3E-02

Max --- --- 3.1E-01 --- --- 1.0E-01 2.1E-01 4.2E-01 8.4E-01 3.6E-02 8.2E-02 1.6E-01

Gonads 

(ovaries)

Min --- --- 4.7E-02 --- --- 1.6E-02 3.3E-02 6.5E-02 1.3E-01 5.6E-03 1.3E-02 2.5E-02

Mode --- --- 9.2E-02 --- --- 3.1E-02 6.4E-02 1.3E-01 2.5E-01 1.1E-02 2.5E-02 5.0E-02

Max --- --- 2.4E-01 --- --- 8.2E-02 1.7E-01 3.3E-01 6.7E-01 2.9E-02 6.5E-02 1.3E-01

Gonads 

(testes)

Min --- --- 6.2E-02 --- --- 2.1E-02 4.3E-02 8.5E-02 1.7E-01 7.3E-03 1.7E-02 3.3E-02

Mode --- --- 1.2E-01 --- --- 4.1E-02 8.4E-02 1.7E-01 3.3E-01 1.4E-02 3.3E-02 6.5E-02

Max --- --- 3.2E-01 --- --- 1.1E-01 2.2E-01 4.4E-01 8.7E-01 3.8E-02 8.5E-02 1.7E-01

Liver Min --- --- 5.1E-02 --- --- 1.7E-02 3.5E-02 7.0E-02 1.4E-01 6.0E-03 1.4E-02 2.7E-02

Mode --- --- 1.0E-01 --- --- 3.4E-02 6.9E-02 1.4E-01 2.7E-01 1.2E-02 2.7E-02 5.4E-02

Max --- --- 2.6E-01 --- --- 8.9E-02 1.8E-01 3.6E-01 7.2E-01 3.1E-02 7.1E-02 1.4E-01

Lung Min --- --- 5.0E-02 --- --- 1.7E-02 3.5E-02 6.9E-02 1.4E-01 6.0E-03 1.4E-02 2.7E-02

Mode --- --- 9.9E-02 --- --- 3.4E-02 6.8E-02 1.4E-01 2.7E-01 1.2E-02 2.7E-02 5.3E-02

Max --- --- 2.6E-01 --- --- 8.8E-02 1.8E-01 3.6E-01 7.1E-01 3.1E-02 7.0E-02 1.4E-01

Remainder 

organs

Min --- --- 4.5E-02 --- --- 1.5E-02 3.1E-02 6.2E-02 1.2E-01 5.3E-03 1.2E-02 2.4E-02

Mode --- --- 8.8E-02 --- --- 3.0E-02 6.1E-02 1.2E-01 2.4E-01 1.0E-02 2.4E-02 4.8E-02

Max --- --- 2.3E-01 --- --- 7.9E-02 1.6E-01 3.2E-01 6.4E-01 2.7E-02 6.2E-02 1.2E-01

Skin Min --- --- 4.8E-02 --- --- 1.6E-02 3.3E-02 6.6E-02 1.3E-01 5.7E-03 1.3E-02 2.6E-02

Mode --- --- 9.4E-02 --- --- 3.2E-02 6.5E-02 1.3E-01 2.6E-01 1.1E-02 2.5E-02 5.1E-02

Max --- --- 2.5E-01 --- --- 8.4E-02 1.7E-01 3.4E-01 6.8E-01 2.9E-02 6.7E-02 1.3E-01

Stomach Min --- --- 5.6E-02 --- --- 1.9E-02 3.9E-02 7.7E-02 1.5E-01 6.7E-03 1.5E-02 3.0E-02

Mode --- --- 1.1E-01 --- --- 3.7E-02 7.6E-02 1.5E-01 3.0E-01 1.3E-02 3.0E-02 5.9E-02

Max --- --- 2.9E-01 --- --- 9.8E-02 2.0E-01 4.0E-01 7.9E-01 3.4E-02 7.8E-02 1.6E-01

Thymus Min --- --- 6.1E-02 --- --- 2.1E-02 4.2E-02 8.4E-02 1.7E-01 7.2E-03 1.6E-02 3.3E-02

Mode --- --- 1.2E-01 --- --- 4.1E-02 8.3E-02 1.6E-01 3.3E-01 1.4E-02 3.2E-02 6.4E-02

Max --- --- 3.1E-01 --- --- 1.1E-01 2.2E-01 4.3E-01 8.6E-01 3.7E-02 8.5E-02 1.7E-01

Thyroid Min --- --- 6.8E-02 --- --- 2.3E-02 4.7E-02 9.4E-02 1.9E-01 8.1E-03 1.8E-02 3.7E-02

Mode --- --- 1.3E-01 --- --- 4.6E-02 9.3E-02 1.8E-01 3.7E-01 1.6E-02 3.6E-02 7.2E-02

Max --- --- 3.5E-01 --- --- 1.2E-01 2.4E-01 4.8E-01 9.7E-01 4.2E-02 9.5E-02 1.9E-01

Uterus Min --- --- 4.7E-02 --- --- 1.6E-02 3.3E-02 6.5E-02 1.3E-01 5.6E-03 1.3E-02 2.5E-02

Mode --- --- 9.3E-02 --- --- 3.2E-02 6.4E-02 1.3E-01 2.5E-01 1.1E-02 2.5E-02 5.0E-02

Max --- --- 2.4E-01 --- --- 8.3E-02 1.7E-01 3.3E-01 6.7E-01 2.9E-02 6.5E-02 1.3E-01

Whole-body 

dose rate

Min --- --- 7.5E-02 --- --- 2.6E-02 5.2E-02 1.0E-01 2.1E-01 8.9E-03 2.0E-02 4.0E-02

Mode --- --- 1.5E-01 --- --- 5.0E-02 1.0E-01 2.0E-01 4.0E-01 1.7E-02 3.9E-02 7.9E-02

Max --- --- 3.9E-01 --- --- 1.3E-01 2.7E-01 5.3E-01 1.1E+00 4.6E-02 1.0E-01 2.1E-01

Data in this table were calculated using measured data from MCW (1958; 1959; 1961a) as given in Table A-39 above. Both maximum average and 

typical average dose rates at 1 meter (3 feet) were taken into account in calculating the effective gamma (photon) whole-body exposure rates; it was 

assumed that a worker spent two hours per day (500 hours per year) in the area of maximum dose rate and six hours per day (1500 hours per year) 

in the area of typical dose rate. The base gamma annual exposures are shown as the whole-body exposure rates above. 

The Min, Mode, and Max values were taken to correspond to the effective dose rates at 1 m, 1 foot (30 cm), and 1 cm from the contaminated surface, 

as per NIOSH direction.

Organ dose conversion factors for the ambient dose equivalent (H*(10) to HT) from NIOSH (2007) were applied to produce the organ doses from 

photons. A geometry of 50% AP and 50% ROT was used and it was assumed that 100% of the exposure was due to photons in the 30-250 keV 

range.

“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. Plants 1 and 2 did not have enough data for these classifications to be applied.
Document No. ORAUT-TKBS-0005 Revision No. 02 Effective Date: 06/14/2007 Page 218 of 225

Table A-42. Annual external electron organ doses during the decontamination and postdecontamination periods (rem per year). Decontamination years (1959-1961)

Plant 4 Plant 6 Plant 6E Plant 7

Organ Type Low Moderate High Low Moderate High Low Moderate High Low Moderate High

Skin Min 2.0E+00 3.8E+00 7.4E+00 7.8E-02 1.6E-01 3.0E+00 1.9E-02 1.2E-01 1.1E+00 8.7E-03 1.5E-01 1.5E+00

Mode 3.7E+00 6.8E+00 1.3E+01 1.4E-01 2.8E-01 5.5E+00 3.4E-02 2.2E-01 2.0E+00 1.6E-02 2.7E-01 2.8E+00

Max 1.8E+01 3.3E+01 6.4E+01 6.8E-01 1.4E+00 2.6E+01 1.6E-01 1.1E+00 9.6E+00 7.6E-02 1.3E+00 1.3E+01

Breast Min 8.0E-01 1.5E+00 2.9E+00 3.1E-02 6.1E-02 1.2E+00 7.5E-03 4.8E-02 4.3E-01 3.4E-03 5.8E-02 6.0E-01

Mode 1.4E+00 2.7E+00 5.2E+00 5.5E-02 1.1E-01 2.2E+00 1.3E-02 8.7E-02 7.8E-01 6.2E-03 1.0E-01 1.1E+00

Max 6.9E+00 1.3E+01 2.5E+01 2.7E-01 5.3E-01 1.0E+01 6.5E-02 4.2E-01 3.7E+00 3.0E-02 5.0E-01 5.2E+00

Testes Min 3.9E-01 7.2E-01 1.4E+00 1.5E-02 3.0E-02 5.9E-01 3.7E-03 2.4E-02 2.1E-01 1.7E-03 2.9E-02 2.9E-01

Mode 7.1E-01 1.3E+00 2.6E+00 2.7E-02 5.4E-02 1.1E+00 6.6E-03 4.3E-02 3.8E-01 3.0E-03 5.1E-02 5.3E-01

Max 3.4E+00 6.3E+00 1.2E+01 1.3E-01 2.6E-01 5.1E+00 3.2E-02 2.1E-01 1.8E+00 1.5E-02 2.5E-01 2.5E+00

Post-decontamination years (1962-1995)

Plant 1 Plant 2 Plant 6E Plant 7

Organ Type Low Moderate High Low Moderate High Low Moderate High Low Moderate High

Skin Min 5.3E-02 1.0E-01 6.7E-01 1.4E-02 2.8E-02 6.0E-01 4.6E-02 8.7E-02 1.7E-01 3.2E-02 7.2E-02 1.4E-01

Mode 9.6E-02 1.9E-01 1.2E+00 2.5E-02 5.0E-02 1.1E+00 8.3E-02 1.6E-01 3.1E-01 5.7E-02 1.3E-01 2.6E-01

Max 4.6E-01 9.0E-01 5.8E+00 1.2E-01 2.4E-01 5.2E+00 4.0E-01 7.5E-01 1.5E+00 2.8E-01 6.3E-01 1.3E+00

Breast Min 2.1E-02 4.1E-02 2.6E-01 5.4E-03 1.1E-02 2.4E-01 1.8E-02 3.4E-02 6.8E-02 1.2E-02 2.8E-02 5.7E-02

Mode 3.7E-02 7.3E-02 4.7E-01 9.8E-03 2.0E-02 4.2E-01 3.3E-02 6.1E-02 1.2E-01 2.2E-02 5.1E-02 1.0E-01

Max 1.8E-01 3.5E-01 2.3E+00 4.7E-02 9.4E-02 2.0E+00 1.6E-01 2.9E-01 5.9E-01 1.1E-01 2.5E-01 4.9E-01

Testes Min 1.0E-02 2.0E-02 1.3E-01 2.7E-03 5.3E-03 1.2E-01 8.9E-03 1.7E-02 3.3E-02 6.1E-03 1.4E-02 2.8E-02

Mode 1.8E-02 3.6E-02 2.3E-01 4.8E-03 9.6E-03 2.1E-01 1.6E-02 3.0E-02 6.0E-02 1.1E-02 2.5E-02 5.0E-02

Max 8.8E-02 1.7E-01 1.1E+00 2.3E-02 4.6E-02 1.0E+00 7.7E-02 1.4E-01 2.9E-01 5.3E-02 1.2E-01 2.4E-01

Data in this table was calculated using measured data from MCW (1958; 1959; 1961a) as given in Table A-39 above. Both maximum average and 

typical average dose rates at about 1 cm were taken into account in calculating the effective electron (beta) base exposure rates; it was assumed 

that a worker spent two hours per day (500 hours per year) in the area of maximum dose rate and six hours per day (1500 hours per year) in the area 

of typical dose rate. The base annual exposures are shown as the skin exposures above.

The Min, Mode, and Max values were taken to correspond to the effective dose rates at 1 cm, 1 foot (30 cm), and 1 meter from the contaminated 

surface, as per NIOSH direction.

Organ dose conversion factors for conversion from skin dose to breast and testes doses were taken from ICRP (996); these are 0.49 for the breast 

and 0 .24 for the testes. In addition, a factor of .8 was used to account for clothing coverage for the breast and testes doses.

“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. Plants 1 and 2 did not have enough data for these classifications to be applied.

Table A-43. Annual dust inhalation, ingestion, and radon daughter intakes during the SLAPS 

postoperations period. 

Inhalation intakes, pCi/yr Ingestion intakes, pCi/yr Radon intakes, WLM/yr

1959-1960 1961-1966 1959-1960 1961-1966 1959-1960 1961-1966

Bulldozer/crane operators 2.27E+04 --- 4.72E+02 --- 4.75E+00 ---

Other material workers (e.g., riggers) 2.27E+04 --- 4.72E+02 --- 4.75E+00 ---

Health & Safety workers 1.13E+04 --- 2.36E+02 --- 2.38E+00 ---

AEC 1.13E+04 3.79E-04 2.36E+02 7.89E-06 2.38E+00 2.28E-01

Guards 2.86E+03 7.89E-04 1.53E+02 1.64E-05 1.54E+00 1.54E+00
Document No. ORAUT-TKBS-0005 Revision No. 02 Effective Date: 06/14/2007 Page 219 of 225

Table A-44. Annual external beta doses, gamma doses and gamma organ doses during the SLAPS 

postoperations period (rem per year). 

1959-1960 1959-1960 1961-1966 1959-1960 1961-1966

Organ

Bulldozer driver, 

rigger

Health & Safety, 

AEC

Health & Safety, 

AEC Guard Guard

Bladder 1.32E+00 6.60E-01 6.34E-02 4.28E-01 1.32E-01

Bone (red marrow) 1.30E+00 6.50E-01 6.24E-02 4.21E-01 1.30E-01

Bone (surface) 2.14E+00 1.07E+00 1.03E-01 6.92E-01 2.14E-01

Breast (female) 1.47E+00 7.34E-01 7.05E-02 4.75E-01 1.47E-01

Colon 1.31E+00 6.53E-01 6.27E-02 4.23E-01 1.31E-01

Esophagus 1.27E+00 6.33E-01 6.08E-02 4.10E-01 1.27E-01

Eye 1.60E+00 8.00E-01 7.68E-02 5.18E-01 1.60E-01

Gonads (ovaries) 1.33E+00 6.65E-01 6.38E-02 4.31E-01 1.33E-01

Gonads (testes) 1.39E+00 6.96E-01 6.68E-02 4.51E-01 1.39E-01

Liver 1.39E+00 6.95E-01 6.67E-02 4.50E-01 1.39E-01

Lung 1.49E+00 7.44E-01 7.14E-02 4.82E-01 1.49E-01

Remainder organs 1.34E+00 6.70E-01 6.43E-02 4.34E-01 1.34E-01

Skin 1.52E+00 7.60E-01 7.30E-02 4.92E-01 1.52E-01

Stomach 1.38E+00 6.89E-01 6.61E-02 4.46E-01 1.38E-01

Thymus 1.42E+00 7.10E-01 6.82E-02 4.60E-01 1.42E-01

Thyroid 1.84E+00 9.19E-01 8.82E-02 5.95E-01 1.84E-01

Uterus 1.24E+00 6.21E-01 5.96E-02 4.03E-01 1.24E-01

Gamma dose rate 2.50E+00 1.25E+00 1.20E-01 8.10E-01 2.50E-01

Beta dose ratea 5.00E-01 2.50E-01 2.40E-02 1.62E-01 5.00E-02

a. Beta doses are not included in the organ doses listed above.
Document No. ORAUT-TKBS-0005 Revision No. 02 Effective Date: 06/14/2007 Page 220 of 225

ATTACHMENT B

U. S. Department of Labor correspondence to NIOSH concerning dose reconstruction for 

members of the Mallinckrodt-designated Special Exposure Cohort
Document No. ORAUT-TKBS-0005 Revision No. 02 Effective Date: 06/14/2007 Page 221 of 225
Document No. ORAUT-TKBS-0005 Revision No. 02 Effective Date: 06/14/2007 Page 222 of 225
Document No. ORAUT-TKBS-0005 Revision No. 02 Effective Date: 06/14/2007 Page 223 of 225

ATTACHMENT C

Notes on How the “Tolerance” or “Preferred” Level for 

Insoluble Uranium Compounds in Air Was Calculated in 1948

(From AEC 1949b) 

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/g 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) = 2.54 × 10-4 μCi per 10 m3

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

1.4 = factor to convert half-life to mean life

4. μCi per m3

: 

= 2.54 × 10-5 

= 56 dpm per m3
Document No. ORAUT-TKBS-0005 Revision No. 02 Effective Date: 06/14/2007 Page 224 of 225

ATTACHMENT C

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 1949b 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/week.
Document No. ORAUT-TKBS-0005 Revision No. 02 Effective Date: 06/14/2007 Page 225 of 225

ATTACHMENT D

Example Calculation of Daily Weighted Average Exposure (DWE)

Calculation for Continuous Experimental Furnace Operators (from MCW 1950s)

Operation

Minutes 

per event

# events, 

day shift

# events, 

night shift

Total 

time, day 

shift

Total 

time, 

night shift

Measured alpha 

conc, dpm/m3

Conc times 

total time, 

day shift

Conc times 

total time, 

night shift

Filling hopper 1.75 10 10 18 18 131000 (avg of 3)

H 153000 L 102000

2,358,000 2,358,000

Removing bottle 3 16 15 48 45 1640 (avg of 3)

H 3440 L 265

78,720 73,800

Dumping bottles into drum 3.5 3 3 10 10 181000 (avg of 2)

H 236000 L 125000

1,810,000 1,810,000

Sampling the drum 1 3 3 3 3 407 (avg of 3)

H 569 L 234

1,221 1,221

General air, furnace area 276 259 238 (avg of 6)

H 908 L 8.5

65,688 61,642

General air, sampling area 60 60 175 (avg of 4)

H 191 L 131

10,500 10,500

General air, smoking room 50 40 314 (avg of 3)

H 378 L 248

15,700 12,560

General air, locker room 15 5 4.2 (avg of 5)

H 6.5 L 0.7

63 21

General air, change room 15 15 68 (avg of 5)

H 101 L 19

1,020 1,020

General air, lunchroom 30 40 1.27 (avg of 8)

H 2.0 L 0.56

38.1 50.8

TOTAL --- --- --- 525 495 --- 4,340,950 4,328,815

Average --- --- --- 505 --- 

Average alpha concentration = 4340950

505

 = 8596 dpm/m3

Average “times tolerance” exposure = 8596

70

 = 123

Note: H and L denote the high and low values respectively.