ORAU TEAM

Dose Reconstruction

Project for NIOSH

Oak Ridge Associated Universities I Dade Moeller I MJW Technical Services

Page 1 of 89

DOE Review Release 12/12/2012

Document Title:

Site Profile for Nuclear Materials and 

Equipment Corporation, Apollo and Parks 

Township, Pennsylvania

Document Number: ORAUT-TKBS-0041

Revision: 02

Effective Date: 11/26/2012

Type of Document: TBD

Supersedes: Revision 01 PC-1

Subject Expert(s): Dennis L. Strenge

Approval: Signature on File Approval Date: 11/09/2012

Dennis L. Strenge, Document Owner

Concurrence: Signature on File Concurrence Date: 11/13/2012

John M. Byrne, Objective 1 Manager

Concurrence: Signature on File Concurrence Date: 11/09/2012

Edward F. Maher, Objective 3 Manager

Concurrence: Signature on File Concurrence Date: 11/14/2012

Kate Kimpan, Project Director

Approval: Signature on File Approval Date: 11/26/2012

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-0041 Revision No. 02 Effective Date: 11/26/2012 Page 2 of 89

PUBLICATION RECORD

EFFECTIVE 

DATE

REVISION

NUMBER DESCRIPTION

09/26/2008 00 Approved new technical basis document for Nuclear Materials and 

Equipment Corporation (NUMEC) in Apollo, Pennsylvania. 

Incorporates formal internal and NIOSH review comments. Training 

required: As determined by Task Manager. Initiated by Dennis L. 

Strenge.

06/02/2009 01 Revised to incorporate the Parks Township facility. Incorporates 

formal internal and NIOSH review comments. Constitutes a total 

rewrite of the document. Training required: As determined by 

Objective Manager. Initiated by Dennis L. Strenge.

02/26/2010 01 PC-1 Page change revision to revise minimum detectable concentration 

for plutonium in urine value based on updated analysis of bioassay 

data. This resulted in changes to Section 5.2 on pages 44 and 45 

and Table 5-4 on page 45. Updated references on pages 43, 44, 51, 

82-83. Added reference on page 81 and updated other references 

on pages 82-83 in the Reference Section. No changes occurred as a 

result of formal internal review. Incorporates formal NIOSH review 

comments. Training required: As determined by Objective Manager. 

Initiated by Dennis L. Strenge. Approval:

Signature on File 02/11/2010

Dennis L. Strenge, Document Owner

Signature on File 02/11/2010

John M. Byrne, Objective 1 Manager

Signature on File 02/17/2010

Edward F. Maher, Objective 3 Manager

Signature on File 02/17/2010

Kate Kimpan, Project Director

Signature on File 02/26/2010

James W. Neton, Associate Director for Science

11/26/2012 02 Revision initiated due to a change in guidance of dose during 

residual period and change in guidance on assignment of medical 

X-ray dose. Sections 3.0, 7.4.1, 7.4.2 were revised. Table 7-1 was 

deleted and Tables 7-2 (now 7-1), 7-3 (now 7-2) and 7-4 (now 7-3) 

were revised. A glossary was added. Incorporates formal internal 

and NIOSH review comments. Constitutes a total rewrite of the 

document. Training required: As determined by Objective Manager. 

Initiated by Dennis L. Strenge.
Document No. ORAUT-TKBS-0041 Revision No. 02 Effective Date: 11/26/2012 Page 3 of 89

TABLE OF CONTENTS

SECTION TITLE PAGE

Acronyms and Abbreviations ............................................................................................................... 8

1.0 Introduction ............................................................................................................................ 11

1.1 Purpose...................................................................................................................... 11

1.2 Scope......................................................................................................................... 11

1.3 Special Exposure Cohorts .......................................................................................... 12

1.3.1 Apollo Site Class............................................................................................. 12

1.3.2 Parks Township Site Class ............................................................................. 12

2.0 Site Descriptions .................................................................................................................... 13

2.1 Apollo Site .................................................................................................................. 13

2.1.1 General Description and Operational History .................................................. 13

2.1.2 Facilities.......................................................................................................... 14

2.1.3 Process Descriptions ...................................................................................... 14

2.1.3.1 General Process Operations............................................................. 15

2.1.3.2 Uranium Process Activities ............................................................... 15

2.1.3.3 High-Temperature Gas-Cooled Reactor Critical Assembly 

Fuel Elements .................................................................................. 17

2.1.3.4 Uranium Nitrate Solution for the University of California ................... 17

2.1.3.5 Incinerator ........................................................................................ 17

2.1.3.6 Thorium Operations.......................................................................... 18

2.1.3.7 Research Activities in the Early Years .............................................. 18

2.1.4 Source Term ................................................................................................... 18

2.1.5 Remediation, Decontamination, and Decommissioning................................... 19

2.1.5.1 Shallow Land Disposal Area............................................................. 20

2.1.5.2 Administrative Building ..................................................................... 21

2.2 Parks Township Site................................................................................................... 21

2.2.1 General Description and Operational History .................................................. 21

2.2.2 Facilities.......................................................................................................... 24

2.2.3 Process Descriptions ...................................................................................... 25

2.2.3.1 Building A – Plutonium Plant............................................................. 25

2.2.3.2 Building B – Multipurpose Fabrication Building ................................. 26

2.2.3.2.1 The Hafnium Facility........................................................ 27

2.2.3.2.2 The Machine Shop .......................................................... 27

2.2.3.2.3 The Metals Plant ............................................................. 27

2.2.3.3 Building C – Highly Enriched Uranium Processing Facility................ 27

2.2.4 Source Term ................................................................................................... 28

2.2.5 Remediation, Decontamination, and Decommissioning................................... 32

3.0 Occupational Medical Dose.................................................................................................... 33

4.0 Environmental Occupational Dose ......................................................................................... 34

4.1 Environmental Internal Dose ...................................................................................... 34

4.2 Environmental External Dose ..................................................................................... 34

5.0 Occupational Internal Dose.................................................................................................... 34

5.1 Internal Exposure Sources ......................................................................................... 34

5.2 In Vitro Bioassay......................................................................................................... 36

5.2.1 Plutonium Urine and Feces Bioassay.............................................................. 37
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5.2.1.1 Plutonium Urine Minimum Detectable Concentrations and 

Frequencies...................................................................................... 37

5.2.1.2 Plutonium Urine Analytical Procedure............................................... 37

5.2.1.3 Plutonium Fecal Minimum Detectable Concentrations and 

Frequencies...................................................................................... 37

5.2.2 Uranium Urine and Feces Bioassay................................................................ 39

5.2.2.1 Uranium Urine Analytical Procedure................................................. 39

5.2.2.2 Urine MDCs and Frequencies........................................................... 41

5.2.2.3 Uranium Fecal Minimum Detectable Concentrations and 

Frequencies...................................................................................... 42

5.2.3 Thorium Exposures......................................................................................... 43

5.2.4 Mixed Fission Products................................................................................... 43

5.2.5 Unmonitored Radionuclides from Recycled Uranium ...................................... 44

5.3 In Vivo Counting ......................................................................................................... 44

5.4 Apollo Process Uranium Air Sampling Studies ........................................................... 46

5.5 Reports of Overexposures and Incidents.................................................................... 48

5.5.1 Apollo Site Incidents ....................................................................................... 48

5.5.2 Parks Township Site Incidents ........................................................................ 48

5.6 Uranium and Plutonium Air Sampling Programs......................................................... 49

6.0 Occupational External Dose................................................................................................... 50

6.1 External Exposure Sources ........................................................................................ 50

6.2 Workplace Radiation Fields ........................................................................................ 51

6.2.1 Beta Radiation ................................................................................................ 51

6.2.2 Photon Radiation ............................................................................................ 51

6.2.3 Neutron Exposures ......................................................................................... 51

6.3 Dosimeter Technology................................................................................................ 52

6.3.1 Beta/Gamma Dosimeters................................................................................ 52

6.3.2 Neutron Dosimeters ........................................................................................ 52

6.3.3 Limits of Detection .......................................................................................... 52

6.3.4 Radiological Records ...................................................................................... 54

6.4 Limitations in Measured Dose..................................................................................... 55

6.4.1 Low-Energy Photons....................................................................................... 55

6.4.2 Neutron Radiation........................................................................................... 57

6.5 Dose Reconstruction Recommendations.................................................................... 58

6.5.1 Recorded Dose Practices ............................................................................... 58

6.5.2 Adjustments to Recorded Dose....................................................................... 58

6.5.2.1 Beta Dose Adjustments .................................................................... 58

6.5.2.2 Photon Dose Adjustments ................................................................ 59

6.5.2.3 Neutron Weighting Factor Adjustments ............................................ 59

6.5.3 Missed and Unmonitored Dose ....................................................................... 59

6.5.3.1 Shallow Dose and Deep Dose.......................................................... 59

6.5.3.2 Neutron Dose ................................................................................... 59

6.5.4 Uncertainty...................................................................................................... 60

6.5.5 Radiation Dose Fraction.................................................................................. 60

7.0 Estimation of Exposure to Residual Activity ........................................................................... 61

7.1 External Dose from Residual Activity in the Workplace............................................... 61

7.1.1 Apollo Site....................................................................................................... 61

7.1.2 Parks Township Site ....................................................................................... 62

7.2 External Ambient Dose from Residual Activity ............................................................ 63

7.3 Occupational Medical Dose ........................................................................................ 63

7.4 Internal Dose from Residual Activity in the Workplace................................................ 63
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7.4.1 Apollo Site....................................................................................................... 63

7.4.2 Parks Township Site ....................................................................................... 65

8.0 Attributions and Annotations .................................................................................................. 68

References ........................................................................................................................................ 69

Glossary ............................................................................................................................................ 77

ATTACHMENT A, BREATHING-ZONE AIR SAMPLE SURVEY DATA............................................. 87
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LIST OF TABLES

TABLE TITLE PAGE

2-1 Apollo site area descriptions .................................................................................................. 14

2-2 Apollo site source and SNM possession limits ....................................................................... 19

2-3 Parks Township site area descriptions ................................................................................... 22

2-4 Parks Township Building A source and SNM possession limits.............................................. 29

2-5 Parks Township Building B source and SNM possession limits.............................................. 30

2-6 Parks Township Building C source and SNM possession limits.............................................. 31

2-7 Parks Township other facility source and SNM possession limits........................................... 31

5-1 Fuel types, chemical form, isotope, and enrichment of NUMEC process material .................. 35

5-2 Uranium source term information ........................................................................................... 35

5-3 Activity composition of Hanford reference fuel-grade plutonium mixture................................. 36

5-4 Plutonium and americium urine bioassay MDC, frequency, and period.................................. 38

5-5 Plutonium fecal bioassay MDC, frequency, and period .......................................................... 39

5-6 Uranium urine bioassay MDC, frequency, and period ............................................................ 41

5-7 Uranium fecal bioassay MDC, frequency, and period............................................................. 43

5-8 Lung-counting MDAs of uranium based on enrichment in picocuries ..................................... 45

5-9 In vivo MDAs for 239Pu, 241Am................................................................................................. 45

5-10 HASL report summary............................................................................................................ 46

6-1 Workplace potential exposures .............................................................................................. 51

6-2 Dosimetry for external whole-body, wrist, and extremity exposures ....................................... 53

6-3 MDLs and potential missed photon, beta, or neutron dose..................................................... 54

6-4 Annual occupational radiation exposures at the Apollo site (Breuer 1977, 1978) ................... 54

6-5 Plutonium photon (and beta) energy factors........................................................................... 57

6-6 Recorded dose practices........................................................................................................ 58

6-7 Interpretation of reported data................................................................................................ 58

6-8 Beta, photon, and neutron radiation energies and percentages for IREP input....................... 60

7-1 External annual dose from residual surface contamination at the Apollo site ......................... 62

7-2 Uranium air concentration and annual intake in the Apollo site residual period ...................... 64

7-3 Plutonium air concentration and annual intake in the Parks Township site residual 

period.....................................................................................................................................67

7-4 Activity fraction for plutonium material at the Parks Township plutonium facility ..................... 67

7-5 Uranium air concentration and annual intake in the Parks Township residual period.............. 68

A-1 HASL-82 weighted BZA survey, December 1959................................................................... 87

A-2 HASL-92 weighted BZA survey, June 1960 ........................................................................... 88

A-3 HASL-103 weighted BZA survey, October 1960..................................................................... 88

A-4 HASL-106 weighted BZA survey, December 1960................................................................. 89

A-5 HASL-114 weighted BZA survey, May 1961........................................................................... 89
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LIST OF FIGURES

FIGURE TITLE PAGE

2-1 Apollo site layout....................................................................................................................14

2-2 B&W Shallow Land Disposal Area ......................................................................................... 21

2-3 Parks Township site layout..................................................................................................... 24

5-1 Correlation of fecal bioassay with air sampling....................................................................... 40

6-1 Gamma spectrum external to plutonium chemical processing line and plutonium 

ceramics line..........................................................................................................................56

6-2 Gamma spectrum external to plutonium ceramics line and plutonium melt glovebox ............. 56

6-3 Dosimeter filter ratios .............................................................................................................56

6-4 Relative contribution by energy group to personnel exposure during plutonium fuel 

fabrication from 1,000 MWd/t plutonium................................................................................. 57

7-1 Reported plutonium air concentrations................................................................................... 66
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ACRONYMS AND ABBREVIATIONS

ADU ammonium diuranate

AEC U.S. Atomic Energy Commission

ANL-E Argonne National Laboratory-East

AWE Atomic Weapons Employer

B&W Babcock & Wilcox (Company)

BZA breathing-zone air

CFR Code of Federal Regulations

Ci curie

cm centimeter

CP Chemical Processing

CRP Chemical Reprocessing

d day

DCF dose conversion factor

DHHS U.S. Department of Health and Human Services

DOE U.S. Department of Energy

dpm disintegrations per minute

DU depleted uranium

EEOICPA Energy Employees Occupational Illness Compensation Program Act of 2000

EU enriched uranium

F fast (absorption type) °

F degrees Fahrenheit

Fab fabrication area

FFTF Fast Flux Test Facility

FP fission product

fpm feet per minute

ft foot

g gram

GA general air

gal gallon

GSD geometric standard deviation

HASL Health and Safety Laboratory

HEPA high-efficiency particulate air

HEU highly enriched uranium

HTGR High-Temperature Gas-Cooled Reactor

hr hour

ICRP International Commission on Radiological Protection

in. inch

IREP Interactive RadioEpidemiological Program

keV kiloelectron-volt, 1,000 electron-volts

kg kilogram

L liter

lb pound
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LEU low-enriched uranium

LLRW low-level radioactive waste

LOD limit of detection

M moderate (absorption type)

m meter

mCi millicurie

MDA minimum detectable activity (or amount)

MDC minimum detectable concentration

MDL minimum detectable level

MeV megaelectron-volt, 1 million electron-volts

mg milligram

min minute

mL milliliter

MOX mixed oxide

MPC maximum permissible concentration

mrad millirad

mrem millirem

MWd megawatt-day

nCi nanocurie

NIOSH National Institute for Occupational Safety and Health

NRC U.S. Nuclear Regulatory Commission

NTA nuclear track emulsion, type A

NU natural uranium

NUMEC Nuclear Materials and Equipment Corporation

ORISE Oak Ridge Institute for Science and Education

OW open window

pCi picocurie

PL plastic

POC probability of causation

R roentgen

R&D research and development

RU recycled uranium

s second

S slow (absorption type)

SEC Special Exposure Cohort

SLDA Shallow Land Disposal Area

SNM Special Nuclear Material

SRDB Ref ID Site Research Database Reference Identification (number)

SRS Savannah River Site

t ton (metric)

TCE trichloroethylene

TLD thermoluminescent dosimeter

TRU transuranic

U.S.C. United States Code

WB whole body
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wt % weight %

yr year

ZPPR Zero Power Plutonium (later Physics) Reactor

μCi microcurie

μg microgram

μR microroentgen

§ section or sections
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1.0 INTRODUCTION

1.1 PURPOSE

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 historical 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.

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

Employees Occupational Illness Compensation Program Act of 2000 [EEOICPA; 42 U.S.C. § 7384I(5) 

and (12)]. EEOICPA, as amended, provides for employees who worked at an AWE facility during the 

contract period and/or during the residual period.

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 2010a):

• 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 2010a). 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.

1.2 SCOPE

This site profile provides specific information about documentation of historical practices at the 

Nuclear Materials and Equipment Corporation (NUMEC) Apollo and Parks Township sites. This site 

profile for NUMEC presents information useful for reconstruction of doses NUMEC employees 

received. 
Document No. ORAUT-TKBS-0041 Revision No. 02 Effective Date: 11/26/2012 Page 12 of 89

1.3 SPECIAL EXPOSURE COHORTS

This section describes the classes in the Special Exposure Cohort (SEC) for the NUMEC sites in 

Apollo and Parks Township, Pennsylvania. 

The SEC classes cover employees who worked at one or both of the Apollo and Parks Township 

facilities. This site profile covers both NUMEC facilities and can be used to perform partial dose 

reconstructions for individuals who worked at either or both facilities.

1.3.1 Apollo Site Class

An SEC class has been identified that includes all AWE employees who were monitored or should 

have been monitored for exposure to ionizing radiation while working at the NUMEC site in Apollo,

Pennsylvania, from January 1, 1957, through December 31, 1983, for a number of days aggregating 

at least 250 work days or in combination with work days within the parameters established for one or 

more other classes of employees in the SEC (Leavitt 2007). 

NIOSH has determined, and the Secretary of the U.S. Department of Health and Human Services 

(DHHS) has concurred, that it is not feasible to reconstruct doses for the following situations:

• Uranium internal exposure before 1960 for lack of bioassay monitoring;

• Thorium and plutonium internal exposures for lack of monitoring data, process description, and 

source term data;

• Potential ambient radiation dose from stack releases;

• Dose from radium-beryllium and polonium-beryllium neutron source fabrication operations; 

and 

• Internal doses if the bioassay data were based on the NUMEC Apollo contactor, Controls for 

Environmental Pollution, from 1976 through 1983, because of concerns about data quality.

Although the combined petition evaluation report for petitions SEC-00047 and SEC-00080 (NIOSH 

2007a) focused on the inability to estimate dose for the above situations during the period from 

January 1, 1957, through December 31, 1983, partial doses can be estimated for workers for whom 

applicable monitoring data are available. The DHHS designation for the worker class indicates that it 

is possible to reconstruct occupational medical dose and components of the internal dose (uranium 

doses starting from 1960). Therefore, individuals with nonpresumptive cancers can be considered for 

partial dose reconstruction (Leavitt 2007).

1.3.2 Parks Township Site Class

An SEC class has been identified that includes all AWE employees who worked at the NUMEC facility 

in Parks Township, Pennsylvania, from June 1, 1960, through December 31, 1980, for a number of 

workdays aggregating to at least 250 workdays occurring either solely under this employment or in 

combination with workdays within the parameters established for one or more other classes of 

employees in the SEC (Leavitt 2008).

NIOSH has determined, and the Secretary of DHHS has concurred, that it is not feasible to 

reconstruct doses for the following situations: 

• Thorium internal exposures for lack of monitoring data and process descriptions; 
Document No. ORAUT-TKBS-0041 Revision No. 02 Effective Date: 11/26/2012 Page 13 of 89

• Internal exposures for work with irradiated fuel and fabrication of radiation sources for lack of 

monitoring data, process descriptions, and source term data; and

• Internal doses if the bioassay data were based on the NUMEC Apollo contactor, Controls for 

Environmental Pollution, from 1976 through 1980, because of concerns about data quality.

Although the SEC petition evaluation report for petition SEC-00108 (NIOSH 2008a) focused on the 

inability to estimate dose for the above situations during the period from June 1, 1960, through 

December 31, 1980, partial doses can be estimated for workers for whom applicable monitoring data 

are available. The DHHS designation for the worker class indicates that it is possible to reconstruct 

occupational medical dose and components of the internal dose when adequate monitoring data are

available. Therefore, individuals with nonpresumptive cancers can be considered for partial dose 

reconstruction (Leavitt 2008).

2.0 SITE DESCRIPTIONS

2.1 APOLLO SITE 

2.1.1 General Description and Operational History

The NUMEC Apollo nuclear fuel site operated under license SNM-145 and Source Material License 

C-3762 that the U.S. Atomic Energy Commission (AEC; a DOE predecessor agency) issued in 1957. 

From 1958 to 1983, the Apollo site was used for small-scale research and production of low-enriched 

uranium (LEU), highly enriched uranium (HEU), and thorium fuels. By 1963, the majority of the Apollo 

facility was dedicated to production of uranium fuel. The major activities at NUMEC Apollo included 

(1) the conversion of LEU hexafluoride (<5% 235U by weight) to uranium oxide (UF6 to UO2) for use in 

light-water-moderated reactors; (2) the conversion of HEU to produce HEU (>20% 235U) nuclear fuel 

for use in the naval reactors program; and (3) the processing of unirradiated uranium scrap (including 

LEU and HEU) from the AEC in the 1960s (B&WNES 1997). 

In 1967, the Atlantic Richfield Company bought the Apollo facility from the original owner of NUMEC. 

In 1971, the facility was purchased by the Babcock & Wilcox Company (B&W), which ran the uranium 

fuel facility and nuclear laundry until production stopped in 1983. Decommissioning support activities 

began in 1978 and the Apollo site ceased all operations in 1983. Early decommissioning activities 

included site characterization, demolition of certain building structures, and selected soil remediation. 

In 1992, the U.S. Nuclear Regulatory Commission (NRC) approved the Apollo site decommissioning 

plan and decommissioning was complete in 1995 (B&WNES 1997). 

The Apollo facility had one main bay (known as the East Bay), and three smaller attached bays known 

as the West Bay, the Box Shop, and the Annex. The site also included a Laundry Building and a 

Small Block Building in the parking lot. The Laundry Building was used for washing protective 

clothing from the nuclear facilities and the Small Block Building was used for storage of processing 

equipment. These buildings were on the east side of the site between Warren Avenue and the 

Kiskiminetas River. The Apollo Office Building was outside the site area across Warren Avenue. The 

Office Building contained a laboratory that was used to analyze radioactive and nonradioactive 

product. A small portion of the building basement housed operations that manufactured instruments 

for the production of nuclear fuels. Although the Office Building was not an original part of the Apollo 

Decommissioning Project, it was included as part of the project in the spring of 1993 after it was 

determined that some floorboards and drain lines contained uranium contamination (B&WNES 1997). 

The parking lot area was bounded by the Kiskiminetas River on the west, Warren Avenue on the east, 

and the offsite area on the north. Figure 2-1 is a general layout of the Apollo site. 
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Figure 2-1. Apollo site layout (Author unknown 2004).

2.1.2 Facilities

The Apollo site was divided into production and process areas and clean areas. Personnel were 

required to enter through the main entrance near the parking lot. Before exiting through the main 

entrance, personnel were to shower if they had entered or worked in a production or process area. 

There were two emergency exits. Entrance into production areas was through the change room with 

the exception of the filter press section of the plant waste treatment area (NUMEC 1963). Production 

and process areas are listed in Table 2-1. 

2.1.3 Process Descriptions

Brief summaries for the principal operations are provided below; additional details can be found in the 

series of AEC Feasibility and Health and Safety Laboratory (HASL) reports in the cited references 

(Forscher 1963; AEC 1960a,b,c, 1961a,b). Inherent in all the operations was nuclear criticality safety 

that governed not only the operations and storage but also the movement of material in the facility. A 

review of the available literature showed that no criticality accidents occurred during Apollo site 

operations.

Table 2-1. Apollo site area descriptions.

Building or 

area location Description Operations/radionuclides

Period of 

operation

CF-1 Ceramic fabrication UO2, ThO2, (metal, powder, and 

oxide)

1957–1970

ThO2: 1963–1970

CF-2 Ceramic fabrication Uranium metal (HEU and DU) UO2, 

and U3O8

Early 1959–1972

PC-1 Process chemistry HEU, EU, DU, (NH4)2U2O7), UO3, 

UF6, UF4, uranyl nitrate, UO2, and 

U3O8

1957–1983

HEU: 1957–1978

LEU: 1957–1983

PC-2 Process chemistry HEU, EU, DU, uranyl nitrate, UO2,

UF6, UF4, and U3O8

1957–1983
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Building or 

area location Description Operations/radionuclides

Period of 

operation

PC-3 Process chemistry HEU, EU, DU, uranyl nitrate, UO2,

UF6, UF4, and U3O8

1957–1983

CP-1 Chemical processing HEU, EU, DU, uranyl nitrate, UO2,

UF6, UF4, and U3O8

1957–1983

CP-2 Chemical processing HEU, EU, DU, uranyl nitrate, UO2,

UF6, UF4, and U3O8

1957–1983

CRP-1 Chemical reprocessing HEU, EU, DU, uranyl nitrate, UO2,

UF6, UF4, and U3O8

1957–1983

CRP-2 Chemical reprocessing HEU, EU, DU, U nitrate, UO2, UF6, 

UF4, and U3O8

1957–1983

CRP-3 Chemical reprocessing Beryllium handling equipment, HEU, 

EU, DU, uranyl nitrate, UO2, UF6, 

UF4, and U3O8

1957–1983

GF-1 1957–1983

QC Quality control 1957–1983

PS 1957–1983

A Vault Process security material. 

Controlled by CP-2

HEU, EU, DU, uranyl nitrate, UO2,

UF6, UF4, and U3O8

1957–1983

C Vault Receiving and shipping 

storage area

1957–1983

E Vault Storage of SNM 1957–1983

F Vault Storage for SNM of all 

enrichments

1957–1983

G Vault Solution storage area, inprocess storage of materials 

for CP-1

HEU, EU, DU, uranyl nitrate, UO2,

UF6, UF4, and U3O8

1957–1983

H Vault Storage of SNM 1957–1983

Waste 

Treatment Area

Filter press section HEU, EU, DU, uranyl nitrate, UO2,

UF6, UF4, U3O8, FPs, and TRU 

elements

1957–1983

GPH Room 2 Health and safety counting 

room

HEU, EU, DU, uranyl nitrate, UO2,

UF6, UF4, U3O8, FPs, and TRU 

elements

1957–1983

Office Building Off site Uranium 1957–1983

Laundry 

Building

Co-60, plutonium, thorium, and 

uranium

1960–1983 

(closed 1984)

Block Building Storage of processing equipment 1957–1983

Incinerator Area 62 (Hoskinson H-100) 30–35 lb/hr, 300 g 235U limit/batch 1960–1983

Sources: Author unknown (2004); NUMEC (1963); B&WNES (1997); Caldwell (1966, 1968a).

2.1.3.1 General Process Operations 

NUMEC Apollo process operations were varied. Because the Apollo site was a research center and 

an all-purpose process center, a variety of radioactive materials and Special Nuclear Materials (SNM) 

were processed. Most of the work performed involved work for commercial reactors but much of the 

work was in the development of better fuel configurations for the burgeoning Navy nuclear program. 

Although this work is not weapons related, the SNM and radiation exposures from this work during the 

AEC operational years (1957 to 1983) are to be included in radiation dose reconstruction in

compliance with NIOSH policy. Some examples are listed below to illustrate the breadth of the SNM 

operations that occurred at the Apollo site.

2.1.3.2 Uranium Process Activities 

A general description of processes for various enrichments of uranium at Apollo is outlined in AEC 

HASL Survey Reports 82, 92, 103, 106, and 114 (Occupational Exposure to Radioactive Dusts
Document No. ORAUT-TKBS-0041 Revision No. 02 Effective Date: 11/26/2012 Page 16 of 89

reports) which cover the period from December 1959 to January 1961 and the Procedures for 

Recovery of Uranium Scrap from 1962 (AEC 1960a,b,c, 1961a,b; NUMEC 1962). The following 

paragraphs provide a description of the process. 

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

Room. The UF6 gas was hydrolyzed to an aqueous solution of UO2F2 and HF. This solution was 

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

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

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

small polyethylene containers to Ceramics Fabrication for further processing (AEC 1960a, p. 3). The 

HASL-92 (AEC 1960b, p. 2) survey from August 1960 noted that the filter cake was dried by a rotary 

kiln rather than the earlier fry pan method. The HASL-114 (AEC 1961b, p. 2) survey from June 1961 

noted that a calciner was added for reduction of ADU to U3O8. The dried cake was discharged 

directly from the kiln into a container, eliminating the manual transfer. 

NOTE: While not specifically stated in the HASL reports, the removal of the filter cake from the 

pressure filter might have been a manual operation. No information could be found to indicate that 

the filter scraping was automated or that it was performed in a glovebox or under a filter hood. 

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

transferred to the Ceramic Laboratory for additional fabrication. HASL-92 (AEC 1960b) noted that the 

UO3 was reduced to UO2 in a rotary kiln rather than the reduction furnace that had been used earlier. 

Ceramics fabrication was performed in the Ceramics Fabrication Area where UO2 was hammer-milled 

in a ventilated enclosure and then moved to the blender glovebox where aerowax was added and the 

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

cake was placed in a glovebox where it was granulated by hand with screens to give the desired 

particle size. The UO2 was loaded into shallow metal pans called “firing boats” and sintered. Sintered 

UO2 was classified, weighed, and packaged. 

Hammer milling, blending, granulating, pellet pressing, centerless grinding, inspecting, and packaging 

were performed in the Ceramics Fabrication Area. 

Uranium-graphite pellet production was a temporary contract activity. U3O8 and graphite powder were 

weighed, blended in a twin shell blender, and prepressed in a Drake press in a large polyethylene tent 

in the ceramic laboratory. The prepressed slugs were granulated. The granulated powder was fed 

into the hopper of the Colton press and compact-pressed. The resultant pellets were cured, 

inspected, and packed for shipping. All steps after prepressing were performed in the Ceramics 

Fabrication Area.

R&D in techniques for coating uranium and UO2 particles with metallic elements such as niobium, 

chromium, tungsten, and molybdenum by halide reduction and vapor deposition was performed. 

According to the 1960 and 1961 HASL survey reports (AEC 1960a,b,c,d, 1961a,b), an approximately 

0.25-in. layer of UO2 powder was placed in a furnace tube (1-in. inner diameter by 2 ft) and heated to 

the desired temperature. A ratio of niobium pentachloride to hydrogen was introduced. The 

pentachloride was reduced by the hydrogen such that niobium metal deposited on the UO2 particles. 

Vibration of the system enabled uniform particle coating. 

All sampling for uranium content and accountability was conducted in the Wet Chemistry, 

Spectrographic, and Metallographic Laboratories. 

Recovery of uranium (U3O8) from uranium-zirconium scrap chips in oil was performed in the Chemical 

Reprocessing Rooms. The 1960 HASL-82 (AEC 1960a) survey report described the recovery of U3O8

from uranium-zirconium scrap as follows: 
Document No. ORAUT-TKBS-0041 Revision No. 02 Effective Date: 11/26/2012 Page 17 of 89

The oil was drained and the chips were transferred to wire mesh baskets and degreased with 

trichloroethane in a ventilated degreasing tank. The chips were then dissolved in hydrofluoric acid in 

chemical fume hoods. The batch was heated and hydrogen peroxide was added to oxidize the 

insoluble UF4 to soluble UF6. The batch was filtered and the filtrate was reduced in chemical fume 

hoods, with insoluble UF4 precipitating preferentially from the solution. The solutions were filtered and 

the UF4 collected in a common filter. The UF4 was then converted to U3O8 by ignition. 

According to Procedures for Recovery of Uranium Scrap (NUMEC 1962), NUMEC dissolved the scrap 

in two designated areas, CRP-2 and CRP-3. The product solutions from the dissolution methods 

were processed to generate insoluble UF4, and were ultimately converted to the final product of U3O8

or UO2.

According to the HASL-92 (AEC 1960b) survey report, the experimental development of recovering 

U3O8 (93%) by solvent extraction was under development at the Apollo site (AEC 1960b). In addition, 

UF4 was being converted to U3O8 by ignition, with the U3O8 granulated manually through screens. A 

facility for the processing of HEU was established on the second floor near the scrap recovery area. 

According to the HASL-103 (AEC 1960c) survey report, a cascade-solvent extraction uraniumzirconium recovery process was under construction. A similar type of extraction process was already 

in operation for Chemical Reprocessing (CRP), which consisted of leaching, feed preparation, solvent 

extraction, ammoniation of strip solution, precipitation, filtration of ADU slurry, kiln drying, and 

packaging as UO3. 

2.1.3.3 High-Temperature Gas-Cooled Reactor Critical Assembly Fuel Elements 

A March 25, 1960, letter (Katine 1960) recommended approval of NUMEC Feasibility Report to 

fabricate 3,000 graphite fuel elements to be used in the General Atomics High-Temperature GasCooled Reactor (HTGR) critical assembly. The job was to involve between 95 and 120 kg of 93%-

enriched U3O8 supplied by another company. The total 235U content of 2,850 fuel elements was to be 

79.339 kg. A letter dated April 1960 to Shapiro (NUMEC) from Wesley Johnson, Deputy Manager 

(AEC), indicated approval of Feasibility Report for the General Atomic HTGR critical assembly fuel 

elements (Katine 1960). 

2.1.3.4 Uranium Nitrate Solution for the University of California 

A letter report dated June 9, 1961 (Katine 1961), to J. E. McLaughlin, Director, Radiation Physics 

Division, HASL, describes a trip to the Apollo site on June 7, 1961, to observe equipment for 

producing uranyl nitrate solution for the University of California. A vague description of NUMEC 

processes and facilities was reported. The report mentions nitric acid and aluminum nitrate solutions 

employed in the solvent extraction process. 

2.1.3.5 Incinerator

Combustible contaminated solid wastes were incinerated in Area 62 of the Apollo site. The system 

consisted of a Hoskinson H-100 incinerator with a main burner in the firebox and an afterburner in the 

stack just above the firebox. Both burners used natural gas for fuel (Caldwell 1968a). 

Packages of contaminated waste were labeled with the 235U content and were burned at a rate of 30

to 35 lb/hr. Ashes were collected in stainless-steel 1-gal containers. After cooling, the ash can was 

placed into a closed container and transferred to an ash handling glovebox. The ashes were sifted, 

sampled for uranium content, and transferred to a clean, lidded 1-gal pail which was transferred to 

one of the plant vaults (Caldwell 1968a).

The operator was protected during charging by a positive inflow of air through the charging door. Ash 

collection was enclosed in an exhausted box. All ash handling was restricted to a glovebox at 
Document No. ORAUT-TKBS-0041 Revision No. 02 Effective Date: 11/26/2012 Page 18 of 89

negative pressure (Caldwell 1968a). Before 1968, this might not have been the case and work in this 

area was perhaps the highest for intakes. 

Combustible gases passed through the afterburner to a water-operated, venturi-type fume scrubber. 

This separated the fly ash from the gas stream. The gases were passed through a packed tower (for 

removal of fine particulates) and discharged through a 15-ft stack. Exhaust air from the ash handling 

glovebox passed through a prefilter and a high-efficiency particulate air (HEPA) filter before discharge 

through a roof stack (Caldwell 1968a). 

2.1.3.6 Thorium Operations

According to Forscher (1963), which cites the 1963 Feasibility Report No. 47 for ThO2, NUMEC was 

to complete fabrication of 626 pellets of ThO2 with no nuclear criticality considerations necessary. 

NUMEC was to purchase 30 kg of ThO2 from Davison Chemical Division of W.R. Grace Company.

NUMEC correspondence (Forscher 1963) with the AEC Oak Ridge Operations Office indicates the 

following plans for the fabrication of ThO2 pellets: 

1. 30 kg of ThO2 would be transferred to the CF-1 Fabrication area. 

2. Working batches of 5 kg would be processed. All powder transfers and handling would be in 

ventilated gloveboxes with a face velocity of 100 fpm. Material would be handled wearing 

latex gloves. 

3. The powder would be slugged to 4-5 g/cm3

, then granulated through 14-mesh screen. 

4. Each batch of powder would be blended in a “V”-type blender in a ventilated glovebox. 

5. The ThO2 pellets would be pressed using a hand press and/or automatic press in a hood 

with a face velocity of 100 fpm. 

6. The ThO2 pellets would be sintered in a hydrogen atmosphere with the out-gases of the 

furnace passing through a filtered exhaust ventilation system. 

7. All pellets would be centerless ground in a ventilated hood. 

8. The final product would be packaged in sausages with each sausage packaged in a 

polyethylene bag. 

Air sampling was performed to characterize thorium exposures during this period. According to a 

health protection program review in 1964, thorium operations involving the blender and weighing hood 

were resulting in excessive airborne concentrations (Thornton and Johnson 1964). 

2.1.3.7 Research Activities in the Early Years

There were research projects at the Apollo site that involved mostly the fabrication of new types of 

fuel in support of the Naval Reactors Branch through the Knolls Atomic Power Plant and Bettis Atomic 

Research Laboratory. The research involved chemical process development with various forms of 

uranium compounds and metal.

2.1.4 Source Term

Three main sources describe the amounts and types of radioactive material that were handled at the 

Apollo site: (1) Federal and State of Pennsylvania licenses for the possession and use of radioactive 
Document No. ORAUT-TKBS-0041 Revision No. 02 Effective Date: 11/26/2012 Page 19 of 89

materials; (2) descriptions and reviews of proposed experiments or jobs handling radioactive material 

in the form of HASL reports and process feasibility reports, which contain information on

radionuclides, quantities, and recommended safety precautions for the described activity; and 

(3) inventory/material handling (accountability) records.

The use of SNM was governed by AEC regulations and licenses, under license SNM-145 and Source 

Material License C-3762 issued by the AEC in 1957 (Docket No 70-135). Some possession limits at 

different periods are listed for the Apollo site in Table 2-2.

Table 2-2. Apollo site source and SNM possession limits.

Areas Source/chemical or physical form

Maximum 

possession

Processing areas, laboratories,

and vaults

U-235 enrichment >5% 5,000 kg

U-235 enrichment ≤5% 75,000 kg

Plutonium as fully clad or encapsulated material 500 kg

Mass Spectrometry Laboratory Uranium in any enrichment 350 g

Plutonium in any form 0.5 g

LLRW storage areas Within fenced areas in approved storage containers 35 g U-235

In buildings meeting safeguards and security requirements 50 kg U-235

Nuclear Decontamination 

Corporation 

Any byproduct material 20 mCi

Any source material 20 g

Any SNM 20 mCi

Source: Reitler (1972).

The Apollo site radiological source term included uranium, thorium, plutonium, and fission and 

activation products (Reitler 1972). At present, no definitive information is available to relate 

measurement of one component of the source term (e.g., plutonium) to another unmonitored 

component (e.g., americium) for a given area or process. Much of the work was R&D, so unique 

source terms could be encountered in a particular job. 

Uranium. Uranium in the form of metal, oxide, and carbide was used for Apollo fabrication, reactor 

fuel, and research studies in the gloveboxes and laboratories (AEC 1960a,b,c, 1961a,b). The typical 

amounts of uranium in use in an area ranged from milligrams to hundreds of kilograms. Occasional 

work involving other uranium chemical forms, such as UF6 or uranyl nitrate was conducted. Uranium 

forms included depleted uranium (DU), natural uranium (NU; i.e., natural enrichment), HEU (up to 

93%), as well as 232U, 233

U, and 236U. Uranium from recycling operations would have included smaller 

activities of nonuranium isotopes, such as 99Tc, 237Np, and 239Pu.

Thorium. Thorium dioxide use was similar to uranium use. The total mass of thorium used on site 

was less than that of uranium overall, but the thorium activity in an area at a given time could have 

been greater or less than uranium activity. Thorium dioxide was obtained from virgin thorium sources. 

The use of reclaimed or reconstituted thorium was not acceptable (Forscher 1963). 

2.1.5 Remediation, Decontamination, and Decommissioning

The HEU processing area on the second floor of the Apollo East Bay underwent remediation from 

1978 until July 1991. All remaining equipment, ventilation systems, piping, and power lines from the 

area were dismantled and disposed of (B&WNES 1997). 

The LEU processing area in the Apollo East Bay was remediated between 1983 and 1984. During 

this period, the LEU processing equipment was removed and disposed of. By October 1984, all 

equipment was removed and sent to the Chem-Nuclear low-level radioactive waste disposal facility in 

Barnwell, South Carolina (B&WNES 1997). 
Document No. ORAUT-TKBS-0041 Revision No. 02 Effective Date: 11/26/2012 Page 20 of 89

The Laundry Building was remediated between 1984 and 1991. In 1984, the processing equipment, 

nonessential utilities, and miscellaneous support systems were volume-reduced, packaged, and sent 

to Chem-Nuclear. The Laundry Building trench, which served as a sump drain for washing machine 

wastewater, was removed in April 1989 (B&WNES 1997). 

All equipment in the Box Shop was removed in 1976. The Small Block Building was demolished and 

stored in the parking lot until accepted at the on-site processing plant (B&WNES 1997). 

As of August 23, 1978, NUMEC had completed decommissioning of its HEU processing at the Apollo

site. All process and related equipment were removed by this date. NUMEC indicated that access to 

the area was limited to authorized personnel. In 1982, the NRC conducted a confirmatory survey to 

identify HEU that might have been present. The report indicates uranium contamination levels in 

grams of uranium to surface area. This report was generated to account for HEU inventory during 

decommissioning. The total grams remaining on and in the floors, walls, pad and ceiling were

estimated to be 35,548.55 g of total uranium with about 23,743.27 g of 235U (Martin 1982). 

Decontamination efforts were completed from 1984 to June 1992 for the Apollo site. In June 1992, 

the NRC approved the Apollo decommissioning plan. Decommissioning occurred from June 1992 to 

1995. NRC staff reviewed B&W groundwater monitoring data, final termination survey, and a 

confirmatory survey in 1996. On April 14, 1997, after notifying the Pennsylvania Department of 

Environmental Protection, the NRC issued a letter to B&W terminating the Apollo license (PDEP 

2008). 

2.1.5.1 Shallow Land Disposal Area

In October 1995, the NRC placed the adjacent B&W Shallow Land Disposal Area (SLDA) on a 

separate license. Until 1970, the SLDA was used as a disposal facility for the Apollo site (and 

possibly the Parks Township site) with about 700,000 ft3 of waste buried in trenches (PDEP 2008). At 

present, options for the disposition of the SLDA are in discussion with the NRC and the former 

licensee. The site might become a Formerly Utilized Sites Remedial Action Program site. As of 2007,

a feasibility study was being conducted by the U.S. Army Corps of Engineers (USACE 2006). The

area is shown in Figure 2-2. 
Document No. ORAUT-TKBS-0041 Revision No. 02 Effective Date: 11/26/2012 Page 21 of 89

Figure 2-2. B&W Shallow Land Disposal Area (USACE 2006).

2.1.5.2 Administrative Building

The Apollo office building had been used for office space since the mid-1950s. Portions of the 

building had been used for an analytical laboratory and to develop and manufacture calibration 

sources in the 1960s and early 1970s. Both operations had been terminated by 1972.

On February 23, 1995, confirmatory surveys were conducted of the administrative building by the 

NRC. This included soil sampling, surface scans, and exposure rate surveys of the soil beneath the 

former concrete slab of the basement of the administrative building and of the surface after the area 

was backfilled and leveled (survey conducted May 23, 1995). All results met the guidelines for 

unrestricted release. 

2.2 PARKS TOWNSHIP SITE

2.2.1 General Description and Operational History

The NUMEC Parks Township site (sometimes referred to as the Advanced Material Center) was on 

115 acres in Parks Township, Armstrong County, Pennsylvania, along U.S. Highway 66 about 3 miles 

southeast of Leechburg and 30 miles northeast of Pittsburgh. The site expanded in stages throughout 

the 1960s. The original Building A was 20,000 ft2

. From 1961 through 1970, a major expansion of 

Building A (the plutonium facility) was completed in five separate expansions (61,000 ft2

). This 

included construction of fabrication areas Fab 2 through Fab 9 and the Drum Warehouse as well as 

termination and remediation of a drum storage area. The main structures were Buildings A, B, and C, 

the outside Storage Areas, and the Trailer Storage Area (Author unknown 2004). 

Building A was constructed in 1959 and 1960 and was authorized to operate in 1961. The Hafnium 

Facility (part of the Building B complex) was constructed in 1960 and was operational in 1961, and the 

plutonium annex was completed and in operation in 1963 (for production of 238Pu sources). The 

Metals Facility (Building B complex) was constructed in 1962 and was operational in 1963. The 

Machine Shop (Building B complex) was constructed in 1964. Building C, the Type II Uranium 
Document No. ORAUT-TKBS-0041 Revision No. 02 Effective Date: 11/26/2012 Page 22 of 89

Facility, was constructed in 1972 in the existing incinerator building (built in 1969), and preproduction 

of Type II fuel began in 1973 (Author unknown undated a). 

The initial functions of the Parks Township facilities were fabrication of plutonium fuel, preparation of 

HEU fuel, and production of zirconium/hafnium bars under AEC and later NRC License SNM-414 

received in March 1961, which allowed the handling of plutonium already on the site. The Parks 

Township site made fuel for the DOE Fast Flux Test Facility (FFTF) at the Hanford Site in the 1970s 

and early 1980 (Author unknown 2004). The FFTF fuel was a mixture of PuO2 and depleted UO2. 

The site also made fuel plates for the DOE Zero Power Plutonium (later Physics) Reactor (ZPPR) in 

the late 1960s and ZPPR-III fuel wafers (Author unknown 2004). Activities included plutonium scrap 

recovery, DU fabrication, HEU fuel manufacturing, source manufacturing (primarily 60Co, PuBe, and 

AmBe), irradiated fuel sample examination, laboratory operations, and supporting nuclear power site 

operations. The activities were conducted in Buildings A, B, and C (Author unknown 2004). 

Production and process areas are summarized in Table 2-3 and discussed further in Section 2.6. 

Although many of the dates of operation in Table 2-3 indicate activity through 1980, some of the 

processes were probably terminated earlier; exact dates could not be determined from the available 

information. Some of the processing areas are shown in Figure 2-3, which depicts the Parks 

Township site layout. 

Table 2-3. Parks Township site area descriptions.

Building or area location Operations Radionuclides

Period of 

operation

Building A–Plutonium 

processing facility

Fabrication of plutonium

reactor fuel pellets, 

blankets, rods 

PuO2; Pu nitrate and oxalates 

(AmBe, PuBe, 1959–1970); alpha, 

beta, and gamma sources

1960–1980

A–Fab 1 Plutonium conversion, fuel 

fabrication for FFTF 

Plutonium nitrate, plutonium oxide, 

depleted UO2

1960–1980

A–East Side of Fab 1 Routine repair and 

maintenance of 

contaminated equipment

All 1960–1980

A–Fab 2 Fuel fabrication for ZPPR Plutonium nitrate, plutonium oxide, 

depleted UO2

1962–1980

A–Fab 3 Manufacturing operations, 

metallography, quality 

control of FFTF fuel

Plutonium nitrate, plutonium oxide 1963–1980

A–Fab 4 Alpha, beta, gamma, and 

neutron source fabrication

AmBe double encapsulated, PuBe 

compacted powder, Ir-192, Cs-137, 

Be-7, Po-210, Co-60, Am-241, Pu238/239, PuO2, plutonium and 

americium metal

1963–1980

A–Fab 5 Scrap recovery Plutonium – various forms 1963–August 

1, 1967, 

Analytical laboratory work All, small quantities of radioactive 

samples

1979–1980

A–Fab 6 Scrap recovery Plutonium – various forms 1968–1973

A–Fab 7 Fuel rod quality control 

tests, nonradioactive 

processes

All, clean and contaminated items 1968–1980

A–Fab 8 Storage All, clean and contaminated items 1970–1980

A–Fab 9 FFTF fuel pin finishing Encapsulated nuclear material 1970–1980

Building A–Hot Cell Room Examination of irradiated 

samples, high-activity 

source fabrication

PuBe compacted powder, Co-60, 

Ir-192

1960–1969

Storage of sources Sealed sources, Ir-192, Cs-137, 

Be-7, Po-210, Co-60, Am-241, Pu238/239, PuO2,

1969–1980
Document No. ORAUT-TKBS-0041 Revision No. 02 Effective Date: 11/26/2012 Page 23 of 89

Building or area location Operations Radionuclides

Period of 

operation

Plutonium Facility Trailer 

Storage Area

Storage of large quantities 

of plutonium and uranium

Plutonium (nonpyrophoric), DU, 

NU, and EU (to 5% 235U)

1961–1980

Building B–Multipurpose 

fabrication building

DU, NU, thorium, 

plutonium

DU metal or alloy, U3O8, 238Pu, 

ThO2

1961–1980

Building B

Hafnium Plant

Metal production Hafnium and Zr-Be alloy 

(nonradioactive) 

1961–1980

Building B

Plutonium Annex

Conversion of 238Pu nitrate 

to oxide

238Pu nitrate, 238Pu oxalate, PuO2

powder or alloys

1963–1980

Building B–Hot Cell Room Large source production:

Postirradiation examination 

of test capsules and fuel 

pins

Co-60, Cs-137, Ir-192, and PoBe,

irradiated, uranium and plutonium

and other TRU elements and FPs

1961–1980

Building B–Metals Plant First floor: small-scale 

metals production, fuel 

pellet production, materials 

testing

First floor: DU, UO2, U3O8, UF4, 

fully clad U-233, U-235, and Pu-239

1963–1974

Second floor: Pu-238 

pacemakers

Second floor: Pu-238-powered 

heart pacemakers

1963–1964

Building B–Machine Shop Occasional machining of 

clad or unclad uranium, 

and clad plutonium and U233; fabrication and repair 

of new and contaminated 

equipment from Parks 

Township and Apollo; 

machining of DU

Fully clad U-233 and Pu-239 and 

clad or unclad U-235 (any

enrichment), primarily DU 

contamination; could include HEU, 

plutonium, thorium, and mixed FPs

1964–1980

Building C, Type II Facility 

or T-2 Plant

HEU processing to form 

sintered product

HEU (1973–1978), soluble 

chloride/oxide complexes, SNM 

oxides (UO3, UO2 and U3O8)

1969–1980

Outdoor Scrap Storage 

Area

Storage UF6 cylinders 1971–1980

a. Sources: Author unknown (2004); Author unknown (undated a); NUMEC (1963)
Document No. ORAUT-TKBS-0041 Revision No. 02 Effective Date: 11/26/2012 Page 24 of 89

Figure 2-3. Parks Township site layout. Plutonium Plant (Building A), Metals Complex 

(Building B), and T-2 Facility (Building C) (Austin 1979). 

2.2.2 Facilities

Building A

This building was originally a plutonium processing facility and at various times it was known as the 

Plutonium Laboratory, the Plutonium Building, and the NUMEC Advanced Material Center. The 

original portion of Building A was designed as a plutonium laboratory to perform R&D that led to 

plutonium-based products. 

Building B

This building was a uranium processing facility with the primary radioactive material being DU, 

although smaller quantities of NU, thorium, and plutonium were also processed in the building. The 

main facilities in Building B were the Hafnium Facility, the Metals Facility, and the Machine Shop.

Building C

This facility was built east of Building A in the 1969 to 1972 timeframe. It was used to fabricate HEU 

fuel, called Type II fuel, from 1973 to 1978. At various times, the building was known as the Type II or 

T-2 Plant (Author unknown 2004).

Plutonium Plant Storage Area (Trailer) 

This was a locked storage area for large quantities of plutonium and uranium. 

Scrap Storage Area 

This was a 150- by 80-ft outdoor area that was enclosed by cyclone fence. It was guarded 24 hr/d. 

Scrap was received in criticality-safe shipping containers known as birdcages and stored as received. 

Specific lots were moved to the Apollo site on NUMEC trucks and logged into the Apollo process 

storage area on the second floor of the Apollo site.
Document No. ORAUT-TKBS-0041 Revision No. 02 Effective Date: 11/26/2012 Page 25 of 89

2.2.3 Process Descriptions

The information in this section is from Author unknown (undated b) unless otherwise noted.

2.2.3.1 Building A – Plutonium Plant

The original portion of Building A was designed as a plutonium laboratory to perform R&D of

plutonium-based products for emerging nuclear businesses. Initial operations were authorized in 

1960. Many experimental fuel forms and compositions were produced in the 1960s, including oxides, 

carbides, and metal alloys in the form of plates, powder, pellets, and special shapes. Work with 

nuclear materials in Building A was conducted in fabrication areas Fab 1 through 7 and Fab 9, and in 

several small laboratories adjacent to the fabrication areas. Fab 8 was used only for storage of 

nuclear materials. All significant work on nuclear fuel materials was done in containment systems 

such as radiochemical hoods and gloveboxes. 

The two largest production runs of fuel were ZPPR fuel plates in the late 1960s and FFTF fuel rods in 

the 1970s and early 1980. More than 12,200 ZPPR fuel elements were produced under contract to 

Argonne National Laboratory–East (ANL-E) using plutonium-uranium-molybdenum ternary alloy 

plates encapsulated in stainless steel. The plutonium content was primarily 239Pu, with 11,500 of the 

plates having 11.5% 240Pu and the balance of the plates having 27% 240Pu. The uranium was 

depleted. FFTF fuel was the largest order of fuel processed in Building A. More than 50,000 FFTF 

pins were made; portions of Cores 1 and 2, and the entire Cores 3 and 4. The FFTF fuel was a mixed 

oxide based on a 20:80 ratio by mass of PuO2 and UO2. The plutonium isotopic content was about

86 wt % 239Pu, 12 wt % 240Pu, and 2 wt % 241Pu with trace amounts of 242Pu and 238Pu. The uranium 

was either natural or depleted, depending on the customer’s specifications.

Plutonium scrap recovery was an integral part of Building A operations. Scrap recovery operations 

were conducted in Fab 5 until August 1, 1967. They were moved in 1968 to a much larger and 

improved operation in Fab 6, which operated throughout the 1970s and into 1980. The scrap was 

dissolved in concentrated nitric acid to which a small amount of hydrofluoric acid had been added. 

The valence of the plutonium ion in the impure nitrate solution was adjusted by the addition of small 

quantities of ferrous sulfamate and sodium nitrite. The plutonium nitrate solution was purified by 

passing through ion exchange columns. The purified solution was then concentrated by evaporation, 

put into bottles, and stored for shipment to the customer or for conversion to plutonium oxide.

Conversion of plutonium nitrate to plutonium oxide was performed at the north end of Fab 1 in HEPAfiltered gloveboxes. 

Alpha, neutron, and thermal sources were produced in Building A. The two most common neutron 

sources were double-encapsulated PuBe metallic sources and compacted mixtures of americium 

oxide and beryllium metal powders. These neutron sources were made in the Fab 4 area. A 

standard alpha source consisted of a plutonium oxide film that was deposited on one or both sides of

a flat metal backing plate. Limited quantities of other neutron, beta, and gamma sources were made 

to customer specifications. The materials that were used to manufacture these specialty sources 

included polonium, plutonium, americium, iridium, cesium, cobalt, and beryllium. Source 

manufacturing always took place in HEPA-filtered gloveboxes, with the exception of high-activity 

sources that were fabricated in the Building A Hot Cell. 

The north end of Building A was divided into two large rooms. The Hot Cell and the Cell Control Area 

occupied the east room, and the Hot Handling Facilities occupied the west room. The Hot Cell was a 

reinforced high-density concrete structure that was designed to shield personnel from gamma 

radiation. The Cell Control Area contained a fume hood for mixing chemicals before inserting them 

into the cell, and a second fume hood over the fission gas analysis equipment. A metallographic cell 

was abutted to the west side of the Hot Cell, just north of the sliding doors. Two small steel-walled hot 
Document No. ORAUT-TKBS-0041 Revision No. 02 Effective Date: 11/26/2012 Page 26 of 89

cells were in the Hot Handling Facilities room. One cell was used as a dissolving cell and the other 

for storage of radioactive specimens.

The interior of the Hot Cell was at a lower pressure than the exterior to prevent radioactive materials 

from reaching workers or the environment. Air from the Hot Cell passed through a HEPA filter before 

it reached the stack. 

Gamma sources of 192Ir and 60Co, which required extensive shielding (i.e., a hot cell), and high-yield 

neutron sources of 210Po/Be were fabricated in the Hot Cell, but the primary work in the Hot Cell was 

destructive postirradiation examination of test capsules and fuel pins that had been irradiated in 

research reactors. 

Fuel processing and source manufacturing in Building A required support from other systems such as 

water heaters; heating, ventilation, and air conditioning; natural gas-fired boiler; air compressor; 

emergency generator; and a cooling tower. Building A housed a repair shop for uncontaminated 

equipment, a shipping and receiving area, administrative offices, and lunchroom areas. SNM was 

neither processed nor stored in these support areas. Routine repair and maintenance of 

contaminated equipment was performed in the glovebox or radiochemical fume hood where the 

equipment was. More extensive repairs were performed in the Warm Maintenance Area, which 

contained a series of ventilated HEPA-filtered gloveboxes that contained a lathe, drill press, and other 

required equipment. 

All plutonium gloveboxes and fume hoods were removed from Building A during a 1981 to 1983 

deactivation program, during which most of the effluent streams that existed during the years of 

plutonium fuel production were eliminated. The workload in Building A shifted to repair and 

refurbishment of contaminated equipment that had been used at reactor sites, building 

decontamination, and low-level radioactive waste (LLRW) volume reduction services for commercial 

customers.

Although these operations involved much smaller quantities of radioactive isotopes, they still 

generated radioactive contamination, so the building exhaust air continued to require HEPA filtration 

before exiting through roof stacks. This exhaust was monitored to ensure compliance with existing 

regulations. As commercial work slowed in the mid-1990s, the pace of building decontamination 

increased. 

2.2.3.2 Building B – Multipurpose Fabrication Building

Building B was constructed in three stages beginning in 1961 when the Hafnium Facility was built to 

produce crystal-bar hafnium. The second stage of construction occurred in 1963 when the Metals 

Plant was built to the east of the Hafnium Facility. The third stage occurred in 1964 when the space 

between the Hafnium Facility and the Metals Plant was closed in to create the Machine Shop. Later 

in its life, the combined facility became known as the Metals Building and then as Building B.

DU was the primary radioactive material processed in Building B, but smaller quantities of NU, 

thorium, and 238Pu were also processed. The DU was primarily in the form of metal or metal alloy, 

and the processing consisted mostly of forming (rolling, etc.) and machining operations that did not 

generate significant airborne emissions. A limited amount of powder products was produced at the 

northeast end of Building B. Plutonium-238 was processed in a room in the northwest corner. All 

238Pu work was performed in interconnected gloveboxes. Receiving and shipping operations were 

conducted in a chemical fume hood. In addition, nonradioactive metals and alloys were processed in 

significant quantities in Building B. Most of the work was production of crystal-bar zirconium and 

hafnium and zirconium-beryllium alloys.
Document No. ORAUT-TKBS-0041 Revision No. 02 Effective Date: 11/26/2012 Page 27 of 89

2.2.3.2.1 The Hafnium Facility

The original product from the Hafnium Facility was crystal-bar hafnium. Crystal-bar zirconium was 

produced in the Hafnium Facility using a similar process. A specialty zirconium alloy product was 

produced consisting of zirconium-beryllium-titanium alloy powder. Metal powders of other alloys were 

produced in the Metals Plant with an identical process. 

Under contract with AEC, 238Pu nitrate was converted to an oxide product in a room in the northwest 

corner of the Hafnium Facility known as the Plutonium Annex. The conversion process was similar to 

the process for converting 239Pu nitrate to fuel products in Building A, but only oxalate precipitation 

was used. The conversion was performed in eight HEPA-filtered gloveboxes. 

2.2.3.2.2 The Machine Shop

The Machine Shop between the Hafnium Facility and the Metals Plant was used to fabricate 

equipment and machine metals in support of the production lines at the Apollo and Parks Township 

sites. The equipment in the Machine Shop included drill presses, lathes, shears, formers, grinders, 

polishers, welders, and sandblasting, degreasing, and other metalworking machinery. Machining of 

DU was performed in the Machine Shop. In addition, repair and refurbishment of equipment from the 

Apollo and Parks Township sites was performed. Some of this equipment contained levels of 

radioactivity that exceeded the criteria at that time for release for unrestricted use.

The machining operations took place on the ground floor. The second floor contained primarily offices 

and a training room, although a small environmental laboratory was at the south end of the second 

floor until 1991. After 1991, most of the second floor was used as office space until decontamination 

operations started in Building B in late 1996.

2.2.3.2.3 The Metals Plant

The Metals Plant was built in 1962 and was operational in 1963. The original layout of the first floor of 

the Metals Plant included equipment to process various metals including zirconium-beryllium, 

tantalum, magnesium, copper, nickel, cadmium, and uranium. The uranium operations included 

electroplating, melting, grinding, and powder handling involving DU.

Metals production from the Metals Plant was small scale and intermittent. Most of the processing 

equipment was removed for resale or disposal in 1973 and 1974. 

The second floor of the Metals Plant initially contained only one office. However, over the years 

offices were added along with two physical and mechanical testing laboratories for quality control 

testing, and the Energy Conversion Laboratory (also called the R&D Laboratory) where R&D projects 

were performed such as the development of 238Pu-fueled heart pacemakers, under an AEC sealedsource license.

2.2.3.3 Building C – Highly Enriched Uranium Processing Facility

Combined with the general expansion of Building A in 1969 and 1970, a new building was erected to 

the east of Building A and called the Incinerator Building. In 1972, the building was modified to 

include facilities for processing HEU. The building sat unused until 1973 when the company received 

a contract to fabricate an HEU product, and processing of SNM in the building was authorized by the 

AEC as an amendment to SNM-414 (Author unknown undated b).

The manufacturing operations involved dissolving HEU in a solution of hydrochloric acid (HCl) and 

hydrogen peroxide, then diluting the solution with demineralized water. The diluted uranium solution 

was fed through dialysis columns and an electrolysis cell. The solution then passed through forming 
Document No. ORAUT-TKBS-0041 Revision No. 02 Effective Date: 11/26/2012 Page 28 of 89

columns to create a solid sintered form (Reitler 1973a). The solid material was rinsed, dried, and 

placed in a furnace. The material was placed into containers and stored before being shipped to a

licensed site for finishing operations. Most of the processing operations were conducted in 

gloveboxes, radiochemical fume hoods, or other ventilated HEPA-filtered enclosures. In addition, the 

room air from the building was exhausted through HEPA filters. 

Materials processing produced several types of liquid wastes: process, laboratory, hexanol, utilities 

and blowdown, and sanitary. Uranium-rich liquid process waste was concentrated in a boildown unit 

and transported, along with solid waste that contained recoverable amounts of uranium, to the Apollo 

site for recovery.

2.2.4 Source Term 

There are three main sources that describe the amounts and types of radioactive material that were 

handled at the Parks Township site: (1) Federal and State licenses for the possession and use of 

radioactive materials; (2) descriptions and reviews of proposed experiments or jobs handling 

radioactive material in the form of safety and process feasibility reports, which contain information 

about radionuclides, quantities, and recommended safety precautions for the described activity; and 

(3) inventory and material handling (accountability) records. 

Some possession limits at different periods are listed for the Parks Township facilities in Tables 2-4

through 2-7.

The use of SNM was governed by AEC regulations and licenses under License SNM-414 issued by 

the AEC in 1961 (Docket No. 70-364). 

The Parks Township site radiological source term included uranium, thorium, plutonium, and fission 

and activation products. No definitive information is available to relate measurement of one 

component of the source term (e.g., plutonium) to another unmonitored component (e.g., americium) 

for a given area or process. Much of the work was R&D, so unique source terms could be 

encountered in a particular job. 

Uranium. Uranium in the form of metals, oxides, and carbides was used for Parks Township fuel 

fabrication (Building A), uranium fuel product (Building C), and reactor fuel research studies in the hot 

cells and laboratories. The typical amounts of uranium in use in any area ranged from milligrams to 

hundreds of kilograms. Work with chemical forms of uranium such as UF6 or uranyl nitrate was 

occasionally conducted. Uranium forms included DU, NU, and EU (up to 93.5%), as well as 232U, 233U 

and 236U. Uranium from recycling operations would have included relatively small activities of 

nonuranium isotopes such as 99Tc, 237Np, 230Th, and 239Pu. 

Thorium. Thorium dioxide was used at the Parks Township site in preparation of special reactor fuel. 

The total mass of thorium that was used on site was probably less than that of uranium, but the 

thorium activity in use in an area at a given time could have been greater or less than uranium activity. 

Thorium dioxide was obtained from virgin thorium sources. 

Plutonium. Chemical forms included metals, nitrates, and oxides. The heat source and heart 

pacemaker programs used 238Pu. The heart pacemaker program used 238Pu nitrate as a starting 

material. The reactor fuel projects used a 239/240Pu-dominated source term. There could be 241Am 

associated with the plutonium source term; in 5 years, the 241Am ingrowth would account for about 1% 

of the total radioactivity in a reactor source term. The 241Pu activity in a heat source is initially less 

than 5% of the total radioactivity (ORAUT 2004). 
Document No. ORAUT-TKBS-0041 Revision No. 02 Effective Date: 11/26/2012 Page 29 of 89

Table 2-4. Parks Township Building A source and SNM possession limits.

Source/chemical or physical form Period Maximum possession Reference

Plutonium and EU 1961–1969 Any combination of plutonium

and U-235 up to 400 kg

Nussbaumer 1965

1969–1979 Any combination of plutonium

and U-235 up to 1,000 kg

AEC ca. 1965–1969

Plutonium (in nonpyrophoric form, 

containing at least 3 wt % Pu-240)

1979–1991 Up to 1,000 kg fissile Rouse 1979; Austin 

1981

1991–end Less than 200 g Haughney 1991

EU of less than 5 wt % U-235 1969–1991 25,000 kg AEC ca. 1965–1969;

Rouse 1979; Austin 

1981

EU above 5 wt % U-235 1979–1981 Possession: < 5 kg 235U

Use: 1 kg effective

Rouse 1979

1981–1991 50 kg fissile Austin 1981

EU above 5 wt % U-235 in storage 1979–1991 50 kg fissile Rouse 1979

Uranium, any enrichment U-235 1991–end Less than 250 g U Haughney 1991

NU or DU and thorium 1969–1979 No limits AEC ca. 1965–1969

NU or DU any form 1979–1991 100,000 kg U Rouse 1979;

Haughney 1991

Plutonium with greater than 5 wt % 

Pu-238

1969–1979 300 g AEC ca. 1965–1969

Pu-238 as oxide or metal 1979–1991 60 g Rouse 1979; Austin 

1981

Pu-238 as sealed source 1981–1991 60 g Austin 1981

Pu-239 as electroplated calibration 

or reference sources

1981–1991 10 g Austin 1981

1991–end 20 g Haughney 1991

Pu-239 as evaporated calibration or 

reference sources

1981–end 5 g Austin 1981

Pu-239 as encapsulated calibration 

or reference sources

1981–1984 230 g Austin 1981

1984–1991 50 g Austin 1984

1991–end 285 g Haughney 1991

U-235 as evaporated calibration or 

reference sources

1981–end 5 g Austin 1981;

Haughney 1991;

Austin 1984

U-235 as encapsulated calibration or 

reference sources

1984–end 5 g Haughney 1991;

Austin 1984

U-235 as electroplated calibration or 

references sources

1991–end 5 g Haughney 1991

U-233 1961–1979 4 kg Nussbaumer 1965;

AEC ca. 1965–1969

U-233 as evaporated calibration or 

reference sources

1981–1991 1 g Austin 1981, 1984

1991–end 2 g Haughney 1991

Any fissile radioactive material 

encapsulated to meet 49 CFR 

173.398 requirements for special 

form material

1969–1972 300 g AEC ca. 1965–1969

Byproduct material encapsulated 1979–end 10 Ci per source of each 

isotope

Rouse 1979; Austin 

1984; Haughney 

1991

Byproduct material any form 1979–end 20 mCi of any isotope Rouse 1979; Austin 

1984; Haughney 

1991

Byproduct material any form, 

contaminated waste

1984–end 1,000 Ci of any isotope Austin 1984;

Haughney 1991

Byproduct material any form, 

contaminated waste on/in equipment

1991–end 5 Ci Haughney 1991
Document No. ORAUT-TKBS-0041 Revision No. 02 Effective Date: 11/26/2012 Page 30 of 89

Source/chemical or physical form Period Maximum possession Reference

Byproduct material any form, 

contaminated waste on/in equipment 

and metallic materials from other 

licensees

1984–end 5,000 Ci Austin 1984;

Haughney 1991

Byproduct material any form, 

contamination in volume reduction 

services waste

1984–end 500 Ci Austin 1984;

Haughney 1991

Table 2-5. Parks Township Building B source and SNM possession limits.

Source/chemical or physical form Period Maximum possession Reference

EU of 5 wt % U-235 1961–1979 5,000 kg Puechl 1965; AEC 

ca. 1965–1969

1991–end Possession: <250 g U-235 Haughney 1991

EU above 5 wt % U-235 1961–1979 500 kg Puechl 1965; AEC 

ca. 1965–1969;

Rouse 1979

1981–1991 Possession: <5 kg

Use: 1 kg effective

Austin 1981

1991–end Possession: <700 g U-235 Haughney 1991

U-233 1961–1979 4 kg Puechl 1965 

Pu-239 with at least 3 wt % Pu-240 1961–1969 250 kg Puechl 1965

Plutonium as fully clad, 

encapsulated, or otherwise 

contained material in operating areas 

or in any form in the storage vault

1969–1979 500 kg AEC ca. 1965–1969

NU or DU and Th 1969–1979 No limits AEC ca. 1965–1969

Plutonium nonpyrophoric form 1991–end <200 g Haughney 1991

Pu-238 encapsulated 1979–1981 60 g Rouse 1979

Byproduct material any form 1979–end 20 mCi of any isotope Rouse 1979;

Haughney 1991

Byproduct material encapsulated 1979–1984 10 Ci per source of each 

isotope

Rouse 1979, Austin 

1984

1984–1991 5 Ci per source of each 

isotope

Austin 1984; 

Haughney 1991

1991–end 10 Ci per source of each 

isotope

Haughney 1991

Byproduct material any form, 

contaminated waste on/in equipment 

1991–end 5 Ci Haughney 1991

Byproduct material any form, 

contaminated waste on/in 

equipment, and metallic materials 

from other licensees

1991–end 5,000 Ci Haughney 1991

NU or DU, any covered or authorized 

activities 

1991–end 100,000 kg U Haughney 1991

Pu-239 as electroplated calibration 

or reference source

1981–1991 5 g Austin 1981

1991–end 20 g Haughney 1991

Pu-239 as encapsulated calibration 

or reference sources

1981–1991 50 g Austin 1981

1991–end 285 g Haughney 1991

Pu-241 as electroplated calibration 

or reference source

1981–1991 5 g Austin 1981

Pu-241 as encapsulated calibration 

or reference source 

1991–end 5 g Haughney 1991

U-233 as evaporated calibration or 

reference sources

1981–1991 1 g Austin 1981

1991–end 2 g Haughney 1991

U-235 as evaporated calibration or 

reference sources

1991–end 5 g Haughney 1991
Document No. ORAUT-TKBS-0041 Revision No. 02 Effective Date: 11/26/2012 Page 31 of 89

Source/chemical or physical form Period Maximum possession Reference

U-235 as encapsulated calibration or 

reference source

1991–end 5 g Haughney 1991

U-235 as electroplated calibration or 

reference source

1981–end 5 g Austin 1981, 

Haughney 1991

Table 2-6. Parks Township Building C source and SNM possession limits.

Source/chemical or physical form Period Maximum possession Reference

EU above 5 wt % U-235 1973–1978 HEUa

NU or DU and thorium 1969–1979 No limits AEC ca. 1965–1969

EU above 5 wt % U-235 1979–1991 Possession: <5 kg U-235

Use: 1 kg effective

Rouse 1979, Austin 

1981

1991–end Possession: <700 g U-235 Haughney 1991

U-235 as encapsulated or 

electroplated calibration or reference 

sources

1981–1991 5 g Austin 1981

1981–1991 Possession: <5 kg

Use: 1 kg effective

Austin 1981

1991–end Possession: <700 g U-235 Haughney 1991

Pu-239 as encapsulated calibration 

or reference sources

1981–1991 5 g Austin 1981

1991–end 285 g Haughney 1991

Pu-239 as electroplated calibration 

or reference sources

1981– 5 g Austin 1981

1991–end 20 g Haughney 1991

Any isotope encapsulated in one or 

more sealed sources

1984–1991 10 Ci Austin 1984

Plutonium in nonpyrophoric form 1991–end <200 g Haughney 1991

a. Possession limits could not be found from available information, but are probably similar to the possession limits for HEU work at the 

Apollo facility (75,000 kg HEU).

Table 2-7. Parks Township other facility source and SNM possession limits.

Place of use

Source/chemical or physical 

form Period

Maximum 

possession Reference

LLRW storage 

areas

Radioactive fissile material in 

approved storage containers

1969–1984 100 g/container AEC ca. 

1965–1969

Plutonium plant 

outdoor storage 

area

EU up to 5 wt % U-235 in UF6

cylinders

1972–1973 75,000 kg UF6 Browne 1978

1973–1984 200,000 kg UF6 Browne 1978

1984–1991 100,000 kg UF6 Austin 1984

Plutonium plant 

storage trailer

Plutonium in nonpyrophoric form 

with at least 3 wt % Pu-241

1979–1991 Up to 1,000 kg fissile Rouse 1979

1991–end <200 g Haughney 

1991

Plutonium and U-235 in approved 

shipping containers with valid 

certificates of compliance

1984–1991 Any quantity Haughney 

1991

EU up to 5 wt % U-235, any 

physical or chemical form covered 

by authorized activities

1979–1991 25,000 kg U Rouse 1979

NU or DU, any physical or 

chemical form covered by 

authorized activities

1979–end 100,000 kg U Rouse 1979

Uranium, any enrichment U-235 1991–end Possession: <250 g 

U

Haughney 

1991

Byproduct material any form, 

contaminated waste

1991–end 1,000 Ci of any 

isotope

Haughney 

1991

Byproduct material, any form, 

contaminated waste on/in 

equipment

1991–end 5 Ci Haughney 

1991
Document No. ORAUT-TKBS-0041 Revision No. 02 Effective Date: 11/26/2012 Page 32 of 89

Place of use

Source/chemical or physical 

form Period

Maximum 

possession Reference

Byproduct material, any form, 

contaminated waste on/in 

equipment, and metallic materials 

from other licensees

1991–end 5,000 Ci Haughney 

1991

Byproduct material, any form, 

contamination in volume 

reduction services

1984–end 500 Ci Austin 1984,

Haughney 

1991

Storage areas EU to >5 wt % U-235 1991–end Possession: <700 g 

U-235

Haughney 

1991

Outside storage 

areas

EU of any enrichment in U-235 1991–end Possession: <350 g 

U

Haughney 

1991

Byproduct material, any form, 

contaminated waste

1991–end 1,000 Ci of any 

isotope

Haughney 

1991

Byproduct material, any form, 

contaminated waste on/in 

equipment, and metallic materials 

from other licensees

1991–end 5,000 Ci Haughney 

1991

Any state except 

Agreement States

Neutron irradiator source 1969–

unknown

Up to 96 g plutonium

as PuBe neutron 

source

AEC ca. 

1965–1969

Byproduct material, any form, 

contaminated waste on/in 

equipment

1991–end 5 Ci Haughney 

1991

Byproduct material, any form, 

contaminated waste on/in 

equipment, and metallic materials 

from other licensees

1991–end 5,000 Ci Haughney 

1991

The two largest production runs of fuel were ZPPR fuel plates in the late 1960s and FFTF fuel rods in 

the 1970s and early 1980. More than 12,200 ZPPR fuel elements were produced under contract to 

ANL-E using plutonium-uranium-molybdenum ternary alloy plates encapsulated in stainless steel. 

The plutonium content was primarily 239Pu, with 11,500 of the plates having 11.5% 240Pu and the 

balance of the plates having 27% 240Pu. The uranium was depleted. FFTF fuel was the largest order 

of fuel processed in Building A. More than 50,000 FFTF pins were made – portions of Cores 1 and 2 

and all of Cores 3 and 4. FFTF fuel was a mixed oxide based on a 20:80 ratio by mass of PuO2 and 

UO2. The plutonium isotopic content was about 86 wt % 239Pu, 12 wt % 240Pu, and 2 wt % 241Pu with 

trace amounts of 242Pu and 238Pu. The uranium was natural or depleted depending on the customer’s 

specifications (Author unknown undated b).

Other Sources. Various sources were manufactured including AmBe, PuBe, and PoBe neutron 

sources; 192Ir, 137Cs, 7

Be, and 60Co beta/gamma sources; 210Po, 241Am, and 238/239Pu alpha sources; 

and 238Pu heat and pacemaker sealed sources.

2.2.5 Remediation, Decontamination, and Decommissioning

The Parks Township site ceased work for DOE operations in 1980. Decontamination and 

decommissioning of the facilities began in 1978 at Building C, and continued through several phases 

for all facilities. Starting in 1994, B&W began final decontamination and decommissioning at the site 

to the extent permitted under the terms of its license. In January 1996, B&W submitted a site-wide 

decontamination and decommissioning plan and subsequent plan revisions in 1997 and 1998. In 

October 1998, NRC approved Revision 3.1 of the plan. Demolition and removal of all facilities was 

started at that time. All decommissioning activities had been completed by January 2002. All waste 

had been shipped to a licensed waste disposal facility, and the final status survey had been 
Document No. ORAUT-TKBS-0041 Revision No. 02 Effective Date: 11/26/2012 Page 33 of 89

performed. After B&W completed 2 years of groundwater monitoring that showed site groundwater 

was within established limits, the NRC terminated the license and released the site for unrestricted 

use on August 24, 2004 (PDEP 2008).

Building A

In 1980, B&W began dismantling the fuel fabrication lines to allow Building A use for other operations. 

Process and analytical equipment, gloveboxes, and hoods were decontaminated and removed. After 

the removal of this equipment, B&W used the area for commercial decontamination. In 1982, B&W 

used areas of the building for nuclear power site support operations. These activities continued into 

1990 and involved the maintenance, testing, and refurbishment of equipment and materials that were 

contaminated with mixed fission and activation products. In the mid-1980s, a facility for LLRW volume 

reduction was under preparation, but the project was terminated in 1988 before operations were 

started (Author unknown undated a).

Building B

Decommissioning of the Hafnium Facility and Metals Facility started in 1976 with the removal of 

process equipment, which was sent for burial or offered for sale (Author unknown undated a). A 

radiation survey of the Metals Facility was performed during September 1980. From 1983 to 1986, 

the Metals Facility was used for storage of nuclear power plant spare parts (Author unknown 

undated a). As of 1991, the facilities were used for nondestructive assay and for calibration and 

testing in relation to decontamination, maintenance, and storage of nuclear industry equipment 

(Haughney 1991). Final decommission of the facility was included in the final site remediation, which 

started in 1998.

Building C

In 1978, B&W ended all HEU operations and began decommissioning efforts at the Parks Township 

Type II facility (Building C). Decommissioning included removal of all process and related equipment 

and the disposal of the contents of the discard ponds associated with Building C. Pond remediation 

included (1) removing the liquids and sludge and solidifying them in 55-gal drums for shipment for

burial, (2) breaking up the asphalt liner for packaging in wooden boxes and steel drums for shipment 

for burial, and (3) packaging soil higher than background into boxes or drums for shipment for burial. 

Decommissioning of the process equipment started with each piece of dismantled equipment being 

assayed for the quantity of SNM. All possible SNM was removed. Equipment was packaged for 

burial. After the equipment was removed, B&W initiated a cleanup of the walls, floors, and ceilings to 

remove loose surface contamination. Surface areas known to contain high-level fixed contamination 

were chipped away and packaged for shipment for burial. The residual activity was determined to be 

fixed and inaccessible to diversion. Access to the building was restricted to authorized personnel, and 

the building’s entrances were secured (Martin 1982). Initial decommissioning was completed in May 

1979. During 1979, drums of U3O8 were temporarily stored at the facility and were shipped in 1980 

(Author unknown undated a). Additional decommissioning was performed during September 1981. 

Final decommissioning was included in the final site remediation that started in 1998.

3.0 OCCUPATIONAL MEDICAL DOSE

The information in this section applies to the Apollo and Parks Township sites. NUMEC apparently 

did not have its own medical X-ray department during AEC operational years and the medical X-rays 

for NUMEC employees appear to have been performed at a local clinic or hospital. Therefore, in

compliance with ORAUT-OTIB-0079, Guidance on Assigning Occupational X-Ray Dose Under 

EEOICPA for X-Rays Administered Off Site (ORAUT 2011a), no occupational medical dose should be 

assigned.
Document No. ORAUT-TKBS-0041 Revision No. 02 Effective Date: 11/26/2012 Page 34 of 89

4.0 ENVIRONMENTAL OCCUPATIONAL DOSE

The Apollo site petition evaluation report for petitions SEC-00047 and SEC-00080 (NIOSH 2007a) 

determined that it is not feasible to reconstruct ambient environmental dose from 1957 through 1965 

for the Apollo site based on limitations associated with stack monitoring data. Reliable information for 

the period after 1965 could not be found to bound the internal and external ambient dose, as 

described below. 

The Parks Township site petition evaluation report for petition SEC-00108 (NIOSH 2008a) did not 

address ambient environmental dose, and reliable information could not be found to bound the 

internal and external ambient dose for the site.

4.1 ENVIRONMENTAL INTERNAL DOSE

Adequate information on environmental air concentrations near the NUMEC Apollo and Parks 

Township sites was not found. Therefore, no estimates of internal ambient dose can be made for 

workers for any period. 

4.2 ENVIRONMENTAL EXTERNAL DOSE 

Information on ambient external dose levels at the NUMEC Apollo and Parks Township sites were not 

found. Therefore, no estimates of external ambient dose can be made for workers for any period. 

5.0 OCCUPATIONAL INTERNAL DOSE

Occupational internal dose is the dose received by an individual from an intake of radioactive material 

while performing tasks in buildings and structures at the Apollo and Parks Township sites or from 

activities outside the buildings, such as handling materials in storage yards. This section describes 

NUMEC internal dosimetry systems and practices and provides supporting data to evaluate internal 

doses that can reasonably be associated with worker radiation exposures covered by EEOICPA. The 

health and safety coverage for both sites were administered by one department. This section covers 

exposure at both facilities because it is difficult to distinguish bioassay results between the sites.

5.1 INTERNAL EXPOSURE SOURCES

The primary sources of internal radiation exposure at the Apollo site were uranium, with some 

potential for exposure to plutonium or thorium dust from the manipulation and chemical processing of 

those materials during uranium scrap recovery and fuel fabrication processes. Uranium enrichment 

levels included depleted, natural, low (3.5%), and high (93%). Exposure to mixed fission and 

activation products was possible at some locations (Laundry Building). 

The sources of internal radiation exposure at the Park Township site were uranium, plutonium, and 

thorium and chemical processing of those materials during plutonium scrap recovery and fuel 

fabrication processes. Uranium enrichment levels at the site included depleted, natural, low (3.5%), 

and high (93%). Exposure to other radionuclides was possible for workers who were involved in 

source fabrication (7

Be, 60Co, 137Cs, 192Ir, 210Po, 241Am, and 238Pu.)

Table 5-1 lists the enrichments and chemical forms of processed radionuclides for the Apollo and 

Parks Township sites. 

ICRP (1994) lists UF6, UO2F2, and UO2(NO3)2 (uranyl nitrate) as type F; UF4 and UO3 as type M; and 

U3O8 and UO2 as type S. The chemical forms and the enrichments varied over time at the NUMEC 

sites. The manufacture of uranium products occurred in most of the buildings at Apollo and Parks 
Document No. ORAUT-TKBS-0041 Revision No. 02 Effective Date: 11/26/2012 Page 35 of 89

Table 5-1. Fuel types, chemical form, isotope, and enrichment of NUMEC process material.

a

Radionuclide or fuel

Chemical form and 

solubility type(s)b

Isotope 

(% in mass, where listed) Enrichment

Uranium UF6, UO2F2, & UO2(NO3)2 (F)

UO3 & UF4 (M)

U3O8 & UO2 (S)

U-234

U-235

U-238

DU, NU, 

LEU (3.5%), HEU

Thoriumc ThO2 (M, S) Th-228,

Th-232

Natural thorium

Plutoniumd PuO2 (M, S, SS) Pu-238 0.64%, Pu-239 2.06%, 

Pu-240 1.07%, Pu-241 95.4%, 

Am-241 0.86%--Activity 

Fuel grade

Aged 10 years

Technetium or other 

TRU elements

Same as the thorium, 

uranium, or plutonium matrix

Tc-99, Np-237 Not applicable

MOXe PuO2 (M, S, SS)

UO2 (M, S)

20% PuO2 and 80% UO2

[7% plutonium – fuel grade/5% 

plutonium – weapons grade]

About 4.5% 235U

Fission and 

activation products

Unknown Be-7, Co-60, Sr-90, Ru/Rh106, Cs-137, Tc-99 (from 

ruthenium), Ir-192

Not applicable

a. Sources: Author unknown (2004), NUMEC (1963).

b. SS refers to highly insoluble plutonium (type Super S). 

c. All thorium work was with unirradiated thorium material.

d. Only small amounts of plutonium were licensed for the Apollo site.

e. Mixed oxide (MOX) work was probably limited to the Parks Township site.

Township. See Tables 2-1 and 2-4 for more information. The dose reconstructor should use the 

solubility type that results in the highest dose. 

Table 5-2 lists NUMEC-specific uranium source term information for various enrichments. For a given 

uranium process, the mass of (long-lived) uranium released to air does not change because of 

enrichment.

Table 5-2. Uranium source term information.

Uranium 

source term Reference

Specific activity

(pCi/μg)

Activity fractions

U-234 U-235 U-236 U-238

NU IMBAa 0.683 0.489 0.023 - 0.489

93.% IMBAa 68.1 0.968 0.030 0.002 0.0003

3.5% IMBAa 2.20 0.818 0.034 - 0.147

2% HPSb 1.20 0.648 0.041 0.0009 0.311

Typical DU IMBAa 0.402 0.155 0.011 0.0005 0.834

Uranium 

source term Reference

Specific activity

(pCi/μg)

Specific constituent activity in mixture 

(μCi/g, nCi/mg, or pCi/μg)

NU IMBAa 0.683 0.334 0.016 - 0.334

93.% IMBAa 68.1 65.9 2.04 0.136 0.020

3.5% IMBAa 2.20 1.80 0.075 - 0.323

2% HPSb 1.20 0.778 0.049 0.001 0.373

Typical DU IMBAa 0.402 0.062 0.004 0.0002 0.335

a. IMBA = Integrated Modules for Bioassay Analysis software.

b. American National Standards Institute N13.22 (HPS 1995).

Many forms of plutonium were possible over the years including metal and oxides. Because the 

feasibility reports for the recovery or manufacture of plutonium have not been located, the exact 

amount processed of each chemical form is not known.

In general, plutonium oxides, carbides, and hydroxides are absorption type S; nitrates and other 

compounds are type M (ICRP 1995, p. 299). Older materials, even when starting out as soluble, can 

have a tendency to oxidize when left in contact with air. Oxides, metals, and old contamination should 
Document No. ORAUT-TKBS-0041 Revision No. 02 Effective Date: 11/26/2012 Page 36 of 89

be treated as type S. If nothing is known about the chemical form of plutonium, either type M or S can 

be used to maximize the dose to the organ of concern. In addition, because highly insoluble forms of 

plutonium might have been present, guidance in ORAUT-OTIB-0049, Estimating Doses for Plutonium 

Strongly Retained in the Lung (ORAUT 2010a), should be followed for the evaluation of highly 

insoluble (Super S) plutonium. Americium-241 is a component of plutonium contamination and should 

be modeled in the lung the same as the plutonium matrix in which it has grown. In other words, the 

americium should be treated as absorption type S if the plutonium is type S (ORAUT 2010b). If the 

plutonium is type Super S, follow guidance in ORAUT (2010a) for assignment of the 241Am solubility 

type.

There are essentially three types of plutonium-based material: reactor grade, weapons grade, and 

fuel grade, which falls between reactor and weapons grades. For this section, lacking specific 

information on the actual composition of the processed plutonium, the composition of Hanford 

plutonium can be used because this was the source of plutonium for FFTF fuel fabrication (Author 

unknown 2004). The activity composition for Hanford reference fuel-grade plutonium (12%) is listed in 

Table 5-3 for fuel aged up to 20 years (ORAUT 2010b). The age of plutonium to assume for a given 

analysis depends on the radionuclide measured in the bioassay analysis. When 238Pu or 239Pu is 

measured, the dose is maximized by assuming longer decay times (20 years). When 241Am is 

measured and the intake is estimated using ingrowth of 241Am from decay of 241Pu, the dose is 

maximized by assuming a short (5-year) decay time. A best estimate of intake can be made by 

assuming a 10-year decay time because this is midway between the possible low and high ages of 

plutonium from the Hanford Site. If the actual age of the fuel is known (such as from an incident 

investigation report), that age can be used in the intake and dose analysis.

Table 5-3. Activity composition of Hanford reference fuel-grade plutonium mixture (12%).a

Mixture 

designation Specific activity (Ci/g)

Years of agingb 0 5 10 15 20

Pu-238 1.71E-02 1.64E-02 1.58E-02 1.52E-02 1.46E-02

Pu-239 5.26E-02 5.26E-02 5.26E-02 5.26E-02 5.26E-02

Pu-240 2.72E-02 2.72E-02 2.72E-02 2.72E-02 2.72E-02

Pu-241 3.09E+00 2.43E+00 1.91E+00 1.50E+00 1.18E+00

Pu-242 3.93E-06 3.93E-06 3.93E-06 3.93E-06 3.93E-06

Am-241 0.00E+00 2.19E-02 3.89E-02 5.22E-02 6.24E-02

a. Source: ORAUT (2010b).

b. Time since separation of Am-241 from the plutonium mix.

5.2 IN VITRO BIOASSAY

The bioassay program for NUMEC workers primarily involved urine and fecal sampling for isotopes of 

uranium, plutonium, and 241Am. Occasional analyses were performed for fission products (FPs) and 

232Th. The reported bioassay data generally include a measurement error that indicates the detection 

level. The bioassay analyses are described in the following sections.

Several bioassay vendors were used to evaluate in vitro bioassay samples for the NUMEC sites. The 

SEC evaluations for the Apollo site (NIOSH 2007a) and the Parks Township site (NIOSH 2008a) 

indicated that Controls for Environmental Pollution has been implicated in the falsification of data and 

that its bioassay analyses provided to NUMEC cannot be considered reliable. Bioassay data from 

Controls for Environmental Pollution should be used only to indicate the potential for exposure to a 

particular radionuclide on a particular date. The data cannot be used in a dose reconstruction to 

evaluate intakes or assign internal dose.

The in vitro bioassay records for individuals nearly always include an indication of the detection level 

for the measurement. Dose reconstructors should use the listed detection level information in 

evaluation of intakes for specific radionuclides when available.
Document No. ORAUT-TKBS-0041 Revision No. 02 Effective Date: 11/26/2012 Page 37 of 89

5.2.1 Plutonium Urine and Feces Bioassay

Plutonium might have been present at the NUMEC sites in several forms that include type M, type S, 

and possibly type Super S material solubility categories. The intake analysis based on bioassay 

monitoring results should evaluate intakes based on all three types and use the type that provides the 

highest dose estimate. 

5.2.1.1 Plutonium Urine Minimum Detectable Concentrations and Frequencies

Plutonium-239 was analyzed in urine from about 1962 to 1999, and 241Am was analyzed starting in 

about 1966. The minimum detectable concentrations (MDCs) are listed in Table 5-4 for NUMEC 

facilities. If an MDC value is needed before the dates in the table, the values for the earliest date 

should be used. No bioassay monitoring results were found between 1985 and 1999. In addition, 

because one health physics department was responsible for the bioassay program at both sites, it is 

difficult to determine from the reported results if the employee worked at the Apollo or Parks Township 

site. It is likely that much of the plutonium bioassay results were for work at the Parks Township site. 

5.2.1.2 Plutonium Urine Analytical Procedure 

Information on the specific procedure used to analyze for plutonium in urine is not known. Based on 

bid specifications (Author unknown undated c), the early analytical procedure probably consisted of 

drying 500 mL of urine to dryness with nitric acid (HNO3). The residue was reevaporated successively 

with HNO3 and then 30% hydrogen peroxide (H2O2) and washed again with HNO3. The ash was 

dissolved in 2N HNO3 and transferred to a lusteroid centrifuge cone. Hydroxylamine hydrochloride, 

lanthanum carrier, and hafnium were added, and the plutonium was coprecipitated with LaF3. After 

centrifuging, the precipitate was dissolved in aluminum nitrate solution and the plutonium oxidized to 

plutonium (IV) with sodium nitrite (NaNO2). Plutonium was extracted into 2-thenoyltrifluoroacetone 

and back-extracted into 8N HNO3. The aqueous phase was evaporated on a planchet and flamed to 

remove organic residue. The planchet was counted in a Nuclear Measurement Corporation gas flow 

proportional counter for 4 hours. The minimum sample volume was 500 mL. Because nearly 

weightless samples were obtained in the procedure, no absorption corrections were made. The 

sensitivity for this procedure was expected to be about 0.44 ±0.20 dpm/L in 1964 (Author unknown 

undated c).

Procedures used to analyze for plutonium in urine in later years are not known. 

From a review of the worker dosimetry records, once per quarter seemed to be the average 

frequency. Special bioassays were ordered for workers who exceeded 40 maximum permissible 

concentration-hours (MPC-hr) of exposure or nose wipes exceeding 25 dpm.

5.2.1.3 Plutonium Fecal Minimum Detectable Concentrations and Frequencies 

The analytical procedure for plutonium fecal analysis has not been located. The estimated MDCs are 

listed in Table 5-5 for NUMEC facilities. 

Fecal sampling was initiated in January 1966 at the NUMEC facilities. Three goals of the program 

were (1) the early detection of acute inhalation exposures, (2) estimation of detected lung burdens, 

and (3) screening for potential chronic exposures (Caldwell 1966). The fecal analyses continued until

about 1985 as indicated in worker dosimetry records. The results, reported as dpm/sample, should 

be considered equivalent to the daily excretion rate (dpm/d).
Document No. ORAUT-TKBS-0041 Revision No. 02 Effective Date: 11/26/2012 Page 38 of 89

Table 5-4. Plutonium and americium urine bioassay MDC, frequency, and period.

a,b

Date Laboratory Radionuclide Frequencyb MDCc,d Errore

10/1961–

12/1965

Controls for 

Radiation

Plutonium Quarterly/as needed 0.28 dpm/L 0.01–0.48 dpm/L

1/1966–

12/1968

Eberline Pu-238 Quarterly/as needed 0.06 dpm/sample 0.03 dpm/sample

1/1966–

12/1975

Eberline Pu-239 Quarterly/as needed 0.06 dpm/sample 0.03 dpm/sample

1/1966–

12/75

Eberline Am-241 Quarterly/as needed 0.06 dpm/sample 0.03 dpm/sample

1/1976–

4/1980

Controls for 

Environmental 

Pollution

Pu-238 As needed (f) (f)

1/1976–

4/1980

Controls for 

Environmental 

Pollution

Pu-239 Quarterly/as needed (f) (f)

1/1976–

4/1980

Controls for 

Environmental 

Pollution

Am-241 Quarterly/as needed (f) (f)

5/1980–

9/1985

Controls for 

Environmental 

Pollution

Pu-238 As needed (f) (f)

5/1980–

9/1985

Controls for 

Environmental 

Pollution

Pu-239 Quarterly/as needed (f) (f)

5/1980–

9/1985

Controls for 

Environmental 

Pollution

Am-241 Quarterly/as needed (f) (f)

1999 Quanterra Pu-238 Unknown 0.0025–0.044 

pCi/L

-

1999 Quanterra Pu-239/240 Unknown 0.0025–0.045 

pCi/L

-

11/1974–

8/1975

Eberline Gross alpha 

(plutonium + 

americium)

Quarterly/as needed <10.0 dpm/sample -

a. Based on review of worker dosimetry reports in Boyd (2006a,b,c,d,e,f).

b. Records indicate quarterly monitoring for plutonium workers, unless an intake was suspected, initiating more frequent 

special sample analyses.

c. Assumes the MDC is twice the reported error

d. The MDC for Controls for Radiation plutonium measurements is twice the 95th percentile of the reported error values for 

zero result measurements (LaBone 2010). 

e. Error values are the error reported (as plus-or-minus values) for zero measurement values.

f. Bioassay data analyzed by Controls for Environmental Pollution are not to be used in internal dose assessments. 

NUMEC health physicist Roger Caldwell believed that fecal sampling was the only satisfactory 

method for estimating lung burdens for insoluble actinide alpha emitters classified as Y in the 

contemporary lung model. The most important alpha emitters included 239PuO2, 241AmO2, 234UO2, 

and 232ThO2 (Caldwell 1966). Caldwell calculated that easily detectable plutonium quantities were 

excreted in feces: 49 dpm/d PuO2 is eliminated from a 16-nCi lung burden, 32 dpm/d by way of the 

feces. One-tenth of this value or 3.2 dpm/d was believed to be a suitable reference level. Workers 

excreting safely below this level were assumed to have nonhazardous lung burdens (Caldwell 1966). 
Document No. ORAUT-TKBS-0041 Revision No. 02 Effective Date: 11/26/2012 Page 39 of 89

Table 5-5. Plutonium fecal bioassay MDC, frequency, and period.

a,b

Date Laboratory Radionuclide Frequencyb MDCc Error

1/1966–1/1977 Eberline Pu-239, Pu-238 

or Am-241

Quarterly/as 

needed

0.1 dpm/sample 0.05 dpm/sample

5/1975–9/1975 Eberline Gross alpha 

(plutonium + 

americium)

As needed 0.1 dpm/sample 0.05 dpm/sample

2/1977–

10/1985

Controls for 

Environmental 

Pollution

Pu-239, Pu-238 

or Am-241

Quarterly/as 

needed

(d) (d)

a. Based on review of worker dosimetry reports in Boyd (2006a,b,c,d,e,f).

b. Records indicate quarterly monitoring for plutonium workers, unless an intake was suspected, initiating more frequent 

special sample analyses.

c. Assumes the MDC is twice the sensitivity or error. 

d. Bioassay data analyzed by Controls for Environmental Pollution are not to be used in internal dose assessments. 

Caldwell noted that fecal sampling should be performed after a person had been away from exposure 

(e.g., plutonium nitrate) for at least 2 days and that individuals would have to be removed from any 

possible UO2 exposure for at least 7 days before fecal data could be used to estimate long-term lung 

burdens (Caldwell 1966; Caldwell, Potter, and Schnell 1967). 

Caldwell analyzed the correlation between lapel breathing-zone air (BZA) sampling and early fecal 

clearance of plutonium and uranium. There was good agreement between the proposed International 

Commission on Radiological Protection (ICRP) lung model (Bates et al. 1966) and lapel sampler data 

(Caldwell, Potter, and Schnell 1967). 

NUMEC health physicists used BZA and general air (GA) sample results to screen for possible 

exposures. If an exposure occurred (based on BZA sample or incident), bioassays of usually both 

fecal and urine samples were collected and correlated with BZA samples. The suspected exposed 

worker was removed from radiation work and fecal and urine samples were collected. This was the 

method employed by the mid-1960s because it was noticed by the NUMEC health physics group that 

fecal sampling was well correlated to the contemporary lung model and lapel or BZA results (see 

Figure 5-1) (Caldwell, Potter, and Schnell 1967). 

The basic fecal sample procedure was that employees were given a quart plastic refrigerator carton, a 

small roll of tape, paper bag, and a written set of instructions. Employees took the bioassay kit home 

to prepare the sample. After depositing the sample in the carton, employees replaced the lid and 

sealed it with tape. The carton was placed in the paper bag and brought back to the laboratory to ship 

to the bioassay vendor. NUMEC added formaldehyde as requested by the vendor (Caldwell 1966). A 

number of different bioassay vendors performed plutonium urine and fecal analyses as observed from 

worker dosimetry records. 

5.2.2 Uranium Urine and Feces Bioassay

Uranium was processed at both the Apollo and Parks Township sites. Enrichment levels varied with 

time and included DU, NU, LEU (3.5%), and HEU (93%).

5.2.2.1 Uranium Urine Analytical Procedure

Information in HASL-82 (AEC 1960a) indicates that before 1960 urine samples were obtained at the 

Apollo site on a monthly and bimonthly basis, with the commercial laboratory Nuclear Engineering and 

Sciences Corporation performing the urine analysis. The results frequently included high values that 

ranged from 50 to 150 μg/L of uranium in urine. However, the available dosimetry records do not 

contain urine bioassay results before late 1959.
Document No. ORAUT-TKBS-0041 Revision No. 02 Effective Date: 11/26/2012 Page 40 of 89

 Figure 5-1. Correlation of fecal bioassay with air sampling 

 (Caldwell, Potter, and Schnell 1967). 

Available uranium bioassay data indicate uranium was analyzed in urine from about late 1959 through 

1988 and in 1999 based on available bioassay reports. Information on the specific procedure used to 

analyze for uranium in urine is not known. Based on a bid specification (Author unknown undated c),

the early analytical procedure probably consisted of taking 0.5 g of sodium carbonate (NaHCO3) and 

adding 125 mL of urine and adjusting the pH with ammonia hydroxide (NH4OH). After 2 hours the 

sample was centrifuged and the precipitated proteins, with the calcium and magnesium salts, were 

discarded. The supernatant was evaporated to dryness with HCl and HNO3, then with hydrogen 

peroxide (H2O2), and finally with HNO3, to ensure destruction of all organic matter (Author unknown 

undated c). The residue was taken up in 0.1N HNO3 and added to a plating cell. A buffer solution 

containing ammonium oxalate, sodium phosphate, and ferrous ammonium sulfate was added and the 

pH adjusted to 5. The uranium was plated on a nickel disk anode in an electrodeposition unit of AEC 

laboratory design at a temperature of 95°

F and 2 amps of current for 1 hour. The nickel disk was then 

dried and counted in a Nuclear Measurements Corporation gas flow proportional counter of the PC 

series (Author unknown undated c). 

With a sample volume of 125 mL in a minimum counting time of 1 hour, the sensitivity was expected 

to be 12 ±3.2 dpm/L at a 90% confidence level. The recovery was expected to average 88% and an 

accuracy of 100 ±15%. The sample counted was to all intents and purposes weightless, so no 

absorption correction was necessary (Author unknown undated c).

The fluorimetric analysis would require a 5-mL sample volume and would have a sensitivity of 1 μg/L 

with a precision of ±10% (Author unknown undated c). However, the reported detection limit was 

0.1 μg/L for results reported by Controls for Radiation, and 5 μg/L for results reported by Eberline in 

available bioassay reports. 

No information on sample analysis methods for other periods is available.
Document No. ORAUT-TKBS-0041 Revision No. 02 Effective Date: 11/26/2012 Page 41 of 89

5.2.2.2 Urine MDCs and Frequencies

The MDCs and approximate frequencies for the uranium urine bioassays are listed in Table 5-6 for 

NUMEC facilities. These values are based on review of bioassay monitoring results (Boyd

2006a,b,c,d,e,f). If an MDC value is needed before the dates listed in the table, the values for the 

earliest date should be used. The measurements based on activity (gross alpha, gross alpha 

uranium, and EU) should be evaluated as total uranium activity.

Table 5-6. Uranium urine bioassay MDC, frequency, and period.

a,b

Date Laboratory Radionuclide Frequencyb MDCc Error

3/1961–

2/1966

Controls for Radiation Total uranium Quarterly/as 

needed

<1 μg/Ld -

9/1972–

12/1976

Eberline Total uranium Quarterly/as 

needed

<5 μg/sample -

1/1977–

11/1987

Controls for 

Environmental Pollution

Total uranium Unknown (e) (e)

1999 Quanterra Total uranium Quarterly/as 

needed

<0.006 μg/L -

4/1962–

1/1967

Controls for Radiation Gross alpha Quarterly/as 

needed

26 dpm/L 13 dpm/L

2/1967–

8/1972

Tracerlab Gross alpha 

uranium

Quarterly/as 

needed

0.2 dpm/sample 0.1 dpm/sample

9/1972–

1/1974

Eberline Gross alpha 

uranium

Quarterly/as 

needed

<50.0 dpm/sample 0.05 dpm/mL

2/1974–

4/1974

Eberline Gross alpha 

uranium

Quarterly/as 

needed

<10 dpm/sample -

2/1974–

12/1976

Eberline Gross alpha 

uranium

Quarterly/as 

needed

2 dpm/sample 1 dpm/sample

3/1964–

6/1967

Controls for Radiation EU Quarterly/as 

needed

4 dpm/L 2 dpm/L

7/1967–

8/1972

Tracerlab EU Quarterly/as 

needed

0.2 dpm/sample 0.1 dpm/sample

1/1977–

2/1987

Controls for 

Environmental Pollution

EU Quarterly/as 

needed

(e) (e)

a. Based on review of worker dosimetry reports in Boyd (2006a,b,c,d,e,f).

b. Records indicate quarterly monitoring for uranium workers, unless an intake was suspected, initiating more frequent 

special sample analyses.

c. When an MDC is not available in the records, assumes the MDC is twice the error. 

d. The MDC for Controls for Radiation for total uranium (1961 to 1966) is based on the reported value (Author unknown

undated c) and should be used as a minimum value in place of the reported values in the individual bioassay records.

e. Bioassay data analyzed by Controls for Environmental Pollution are not to be used in internal dose assessments. 

The MDC value for the Controls for Radiation is set to 1 μg/L even though the vendor reports often 

indicate a value of 0.1 μg/L. Therefore, the value in Table 5-6 has been set to 1 μg/L based on the 

reported value for the analytical method by Author unknown (undated c). This value should be used 

as a minimum value in place of the values in the bioassay records; larger reported values can be used 

in the intake assessment. 

Care should be taken in evaluating the Controls for Radiation reported urine bioassay results because

the records could have errors in the reported units. The reported values are in units of micrograms 

per milliliter although sometimes are listed as micrograms per liter. 

Urine samples were typically 24-hour samples. The frequency for uranium urine samples was in 

general:

• Wet analytical chemistry personnel every 3 months,
Document No. ORAUT-TKBS-0041 Revision No. 02 Effective Date: 11/26/2012 Page 42 of 89

• Nonradiation workers annually,

• Production workers every 4 to 6 weeks maximum (NUMEC 1963),

• Maintenance personnel every quarter (NUMEC 1963),

• All other (radiation) personnel every 6 months (NUMEC 1963), and

• At the discretion of health and safety in the event of an incident such as a uranium 

hexafluoride release (NUMEC 1963).

Although the above information indicates nonradiation workers were monitored annually, many worker 

files contain no record of bioassay monitoring. From a cursory review of the worker records, once per 

month seemed to be the highest frequency, although an average frequency was closer to once per 

quarter for uranium workers. Special bioassays were ordered for those workers who exceeded

40 MPC-hr of exposure or nose wipes exceeding 25 dpm.

There were as many as about 100 urine bioassay analyses each month. In the early years (to about 

1964), urine samples were normally analyzed on a weight basis and then a radiometric analysis was 

performed if the level approached 50 μg/L. As stated above, the urinary control levels were 50 μg/L 

and/or 500 dpm/L for HEU (93%). According to the 1963 program review, the records for the few 

years before 1963 indicated that there had been no restrictions as a result of the personnel monitoring 

program (Hervin and Pryor 1963). However, during a hazard evaluation conducted by the AEC in 

1959, a number of personnel had a urine concentration result between 50 and 150 μg/L (AEC 1960a). 

In later years urine was analyzed using one or both methods (weight basis and radiometric basis.)

The maximum allowable concentration in urine was 500 dpm/L for 93%-enriched 235U (NUMEC 1963). 

At some time in 1963 this was decreased to 300 dpm/L, and by October 1964 this was decreased 

further to 150 dpm/L (Thornton and Johnson 1964). The NU urine control limit was 50 μg/L weight 

basis or 75 dpm/L activity basis (Hervin and Pryor 1963). 

By the mid-1960s, both fecal and urine bioassay samples were being collected by NUMEC to 

determine the appropriate clearance model. The permissible NU urine level of 75 dpm/d was used 

(Caldwell, Potter, and Schnell 1967). 

5.2.2.3 Uranium Fecal Minimum Detectable Concentrations and Frequencies

The analytical procedure for uranium fecal analysis has not been located. The MDCs and 

approximate frequencies for the uranium urine bioassays are listed in Table 5-7 for the Apollo and 

Parks Township sites. The fecal analysis results reported as dpm/sample should be considered 

equivalent to the daily excretion rate (dpm/d). When results are provided as dpm/g along with the 

sample weight, the daily excretion value is based on the total sample activity evaluated as the product 

of the sample weight and the reported activity concentration. 

Fecal sampling (in addition to urine sampling) began on a large scale at the Apollo uranium plant in 

June 1966 (Caldwell 1966). The fecal analyses continued until about 1985 as indicated in worker 

dosimetry records. Caldwell observed that some UO2 exposures were poorly detected in urine 

(Caldwell 1966). According to Caldwell, literature available at the time indicated that whole-body 

(WB) counting was effective for EU lung burdens greater than 7 nCi, but fecal sampling was 

necessary for smaller fractions of the permissible lung burden. 
Document No. ORAUT-TKBS-0041 Revision No. 02 Effective Date: 11/26/2012 Page 43 of 89

Table 5-7. Uranium fecal bioassay MDC, frequency, and period.a

Date Laboratory Radionuclide Frequencyb MDCc

6/1967–6/1972 Tracerlab Radiometric uranium Quarterly/as needed 2 dpm/sample

7/1972–1/1976 Eberline Total uranium Quarterly/as needed <5 μg/sample

2/1976–

10/1985

Controls for Environmental 

Pollution

Total uranium Quarterly/as needed (d)

a. Based on review of worker dosimetry reports in Boyd (2006a,b,c,d,e,f).

b. Records indicate quarterly monitoring for uranium workers, unless an intake was suspected, initiating more frequent 

special sample analyses.

c. The MDC for radiometric uranium (Tracerlab) is based on a reported error value of about 1 dpm/sample, multiplied by 2.

d. Bioassay data analyzed by Controls for Environmental Pollution are not to be used in internal dose assessments. 

Caldwell used a permissible fecal excretion rate of 50 dpm/d for uranium assuming the ICRP 

recommended 380-day half-time for chronic UO2 exposures (Caldwell, Potter, and Schnell 1967).

By 1972 or later, Caldwell believed that fecal sampling for all radionuclides was a valuable tool for 

early assessment of inhalation exposures but that information on the urine-to-fecal-excretion ratio was 

necessary for the complete interpretation of urine data. Caldwell found that the most important use of 

fecal-sampling data was for estimating the magnitude of single inhalations of uranium from accidental 

exposures. For uranium plant operations, Caldwell believed that lung burdens should be based on 

urine sampling or in vivo counting (Caldwell ca. 1972). 

5.2.3 Thorium Exposures

There is not sufficient air-sampling or urinalysis information available for the NUMEC sites to conduct 

a thorium intake analysis for workers in general. If the case files include thorium measurement 

results, an intake and dose assessment can be performed. Thorium was processed at the Apollo site

for a few years starting in 1963 and at the Parks Township site in the early 1960s. Limited information 

on thorium bioassay analyses was found in worker dosimetry records. In 1971, the error was reported 

as 0.1 dpm/sample for 232Th (Tracerlab analysis) for a 100-mL urine sample, which provides a 

minimum detectable activity (MDA) value of 0.2 dpm/sample. The fecal analysis error for the same 

workers was reported as 0.1 dpm/sample, which provides an MDA value of 0.2 dpm/sample.

The thorium oxides, carbides, and hydroxides are absorption type S; nitrates and other compounds 

are type M. The dose reconstructor can assume either type M or S (ICRP 1994) to maximize the 

internal dose. The internal dose is evaluated for intake as 232Th. Because these bioassay analyses 

were specific for 232Th, consideration should be given to the 228Th that would be present from decay of 

232Th. Based on ORAUT (2012a), after preparation of the thorium dioxide from ore, the amount of 

228Th initially decreases and later builds in from continued decay of the 232Th. The recommended 

assumption of an 80% ratio of 228Th to 232Th is appropriate for cases in which the time from initial 

purification is unknown. The internal dose is evaluated based on the estimated intake of 232Th plus an 

equilibrium activity of 80% as 228Th.

5.2.4 Mixed Fission Products

The records indicate urine bioassay analyses were performed occasionally for mixed FPs from 1962 

through 1968. The MDC for these analyses was about 5 dpm/sample throughout the period, with 

sample analysis provided by Controls for Radiation. If bioassay records are found in case files with 

results provided by Controls for Environmental Pollution, the results should not be used to estimate 

intake of mixed fission products.

Exposure to FPs at the Apollo site was most likely to have occurred in the Laundry Building as part of 

the commercial decontamination of clothing by laundering. Exposure to FPs at the Parks Township 

site would most likely be related to source fabrication (60Co and 137Cs). The radionuclides 

representing mixed FPs could have included fission and activation products representative of reactor 
Document No. ORAUT-TKBS-0041 Revision No. 02 Effective Date: 11/26/2012 Page 44 of 89

operations. Possible radionuclides include 60Co, 90Sr, 99Tc [a recycled uranium (RU) contaminant], 137Cs, 106Ru/Rh, and possibly others. No information is available on the methods used to analyze for 

mixed FPs in urine. Urine bioassay data for mixed FPs should be used, if included in the case files, to 

estimate intakes of FPs. ORAUT-OTIB-0054, Fission and Activation Product Assignment for Internal 

Dose-Related Gross Beta and Gamma Analyses (ORAUT 2007a) can be used to determine the 

radionuclide appropriate for the dose calculation. This technical information bulletin requires 

specification of the decay time for the fission product mixture. Because little is known about the FP

material likely to be present at the Apollo site, a decay time of 1 year should be assumed to provide 

an assessment of dose that is favorable to claimants (ORAUT 2007a). 

The urine bioassay results do not indicate if the measurements are based on beta or gamma analysis, 

so the intake should be based on both methods according to guidance in ORAUT-OTIB-0054

(ORAUT 2007a). 

5.2.5 Unmonitored Radionuclides from Recycled Uranium

The uranium processed at the Apollo and Parks Township sites might have included RU. This 

material would contain contamination radionuclides formed during fission and activation processes 

when the material was irradiated in production or test reactors. The spent fuel elements were 

reprocessed to recover the uranium, which was returned to the DOE inventories along with trace 

contaminants that included 99Tc, 237Np, and 239Pu. The intake of RU contaminant radionuclides can 

be estimated using the contaminant fraction values in Table 5-12 of the occupational internal dose 

technical basis document for the Fernald Environmental Management Project (ORAUT 2004). The 

estimated intake of each radionuclide is obtained by multiplying the estimated uranium intake by the 

appropriate contaminant fraction. The assignment of material solubility type is based on the guidance 

in Table 5-2 of ORAUT-OTIB-0060 (ORAUT 2007b). Because plutonium is a minor contaminant in 

the recycled uranium matrix, the consideration of type Super S plutonium is not necessary for 

evaluation of internal dose from this source of plutonium.

5.3 IN VIVO COUNTING

In vivo or lung counting for 239Pu, 241Am, uranium, and some FPs was started in about 1966 for 

incident evaluation (Caldwell 1966, 1968b). The counting in 1966 was provided by the University of 

Pittsburgh Low Level Radioactivity Monitoring Facility at the Presbyterian-University Hospital, using a 

thin NaI crystal system (Caldwell and Judd 1966). The bioassay records indicate this is the facility 

where the majority of routine WB counts were performed for NUMEC workers starting in 1969. 

In 1968 and 1971, Helgeson performed WB counts on individuals for fission products, 235U, 241Am, 

with 239Pu estimated from the 241Am results based on expected activity ratios for 239Pu/241Am (Caldwell 

1968b). The MDA for 235U was listed as 0.08 mg for this system. The MDA for 241Am ranged from 

0.13 to 0.38 nCi for individual measurements at the 2-sigma level. The 239Pu activity was estimated 

using an activity ratio ranging from 9 (ZPPR fuel) to 19. 

The procedure for lung counting used by the University of Pittsburgh Low Level Radioactivity 

Monitoring Facility included a standard stretcher technique that was used with two 5- by 3-in. NaI(Tl) 

dual crystal low energy detectors positioned above the stretcher near the anterior chest region of the 

subject (Boyd 2006g, pp. 92–97). The calibration was for 0.5 keV per channel and the count time was 

40 minutes for gross counts and background. Background correction was made using spectra 

obtained from unexposed individuals. Minor differences in the potassium and cesium body burdens 

were corrected by normalizing the spectra at an energy region from 90 to 125 keV. Activity 

calibrations were obtained from data published by Los Alamos National Laboratory using a detector 

configuration identical to the one used by the Laboratory. The calibration factor was adjusted for 

attenuation due to variation in the subject’s chest wall thickness as measured with an 

encephaloscope. The evaluation of 239Pu activity was based on the assumption that only 239Pu was 
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present and all 17-keV X-rays were from 239Pu. The difficulty in measuring the low-energy X-rays 

results in MDA values that represent significant lung burdens (ORAUT 2007a).

Lung counts were performed from about 1966 to 1992, and possibly later. Uranium lung counting 

started regularly in December 1971. Plutonium and americium counting started in 1966 (Caldwell 

1966) and on a regular basis in 1968. FPs were counted intermittently. Lung counts are in general 

not as reliable as urinalysis (or fecal analysis – Caldwell 1966) for routine monitoring. However, this 

monitoring was routine and was used to assess routine exposures to transuranic (TRU) elements and 

FPs and to further analyze results from accidental acute and routine chronic intakes. Table 5-8 lists 

uranium lung-counting MDAs for common enrichments that might have been processed at NUMEC. 

Actual MDAs from worker records should be used if available because the MDA for a measurement is 

dependent on the chest wall thickness, which varies by individual. The MDA for 235U was about 

63 μg, as indicated from the cursory review of worker dosimetry records in 1971 and later years, 

which is a reasonable default MDA value.

Table 5-8. Lung-counting MDAs of uranium based on enrichment in picocuries.

a

Uranium source term Total uranium MDA (μg) Total uranium MDA (pCi)

NU 8.75E+03 5.98E+03

93.00% 6.77E+01 4.61E+03

3.50% 1.80E+03 3.96E+03

2% 3.15E+03 5.09E+03

Typical DU 3.17E+04 1.27E+04

RU (1% 235U) 6.30E+03 5.73E+03

a. Based on U-235 MDA of 63 μg.

Table 5-9 is a summary of in vivo MDAs for 239Pu and 241Am based on a review of claimant files. The 

results are generally reported as WB counts in the dosimetry records. Data after 1985 are sparse in 

the bioassay records.

Table 5-9. In vivo MDAs for 239Pu, 241Am.a

Year

Pu-239 MDA (nCi) Am-241 MDA (nCi)

Minimum Maximum Counts Minimum Maximum Counts

1968 NRb NRb NRb 0.13 0.38 17

1969 NRb NRb NRb NRb NRb NRb

1970 NRb NRb NRb NRb NRb NRb

1971 NRb NRb NRb NRb NRb NRb

1972 9.0 11.5 3 0.13 0.13 1

1973 5.6 15.6 46 0.11 0.21 28

1974 5.44 21.3 122 0.09 0.22 96

1975 4.8 19.9 133 0.11 0.21 104

1976 5.0 20.3 109 0.11 0.19 91

1977 4.4 19.6 113 0.09 0.19 88

1978 4.7 19.0 132 0.10 0.19 100

1979 5.16 24.3 168 0.08 0.26 132

1980 5.03 28.2 132 0.09 0.21 94

1981 7.21 27.8 55 0.12 0.20 31

1982 7.12 34.3 77 0.12 0.21 44

1983 9.41 15.6 6 0.12 0.16 4

1984 8.67 22.32 9 0.12 0.15 5

1985 8.84 31.07 31 0.11 0.22 29

a. From a review of worker dosimetry records (Boyd 2006a,c,d,g,h,i). Values for 1968 

through 1971 are based on the Helgeson system, with remaining values for the University 

of Pittsburgh system.

b. NR = none reported.
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The in vivo bioassay records for individuals nearly always include an indication of the detection level 

for the measurements where the radionuclide was not detected. The detection levels are reported as 

“less than” values. Dose reconstructors should use the listed in vivo detection level information in 

evaluation of intakes for specific radionuclides.

5.4 APOLLO PROCESS URANIUM AIR SAMPLING STUDIES

This discussion focuses on documented air sampling data from five HASL reports for the Apollo site. 

The information in the reports can be useful in providing an estimate of the likely intake for a worker 

when job description and location are known. A general description of processes for various 

enrichments of uranium at Apollo is outlined in HASL Survey Reports 82, 92, 103, 106, and 114 

(Occupational Exposure to Radioactive Dusts reports), which cover the period from December 1959 

to January 1961 and the Procedure for Recovery of Scrap Uranium from 1962 (AEC 1960a,b,c, 

1961a,b; NUMEC 1962). HASL survey reports contain results for loose and total alpha samples, GA 

samples, fixed-station and weighted BZA samples, and some personnel protective equipment, 

ventilation description, and general observations of activities. Attachment A lists the results of all BZA 

surveys for HASL Survey Reports 82, 92, 103, 106, and 114. A summary of the HASL reports is 

listed in Table 5-10 based on information from Attachment A.

Table 5-10. HASL report summary. 

Report Description of report Date

HASL-82 

(AEC 1960a)

Production of UO2 from UF6, UO2 pellet formation, uranium-graphite pellet 

production, recovery of U3O8 HEU from uranium-zirconium scrap, R&D for coating 

uranium particles. NU was used in coating studies. Fully enriched (93%) HEU

was used in the other processes.

12/1959

HASL-92 

(AEC 1960b)

Production of UO2 from UF6, UO2 pellet formation, uranium-graphite pellet 

production, recovery of U3O8 HEU from uranium-zirconium scrap, R&D for coating 

uranium particles. NU was used in coating studies. Fully enriched (93%) HEU

was used in the other processes.

6/1960

HASL-103 

(AEC 1960c)

Production of UO2 from UF6, UO2 pellet formation, uranium-graphite pellet 

production, recovery of U3O8 HEU from uranium-zirconium scrap, R&D for coating 

uranium particles. NU was used in coating studies. Uranium (1.8% to 93%) was 

used in the other processes.

10/1960

HASL-106 

(AEC 1961a)

Processing (93%) UO3 from UF6 and 93% U3O8 from uranium-zirconium, powder 

handling of HEU (93%) in the Ceramics Laboratory. Pressing EU (3.5%), 

centerless grinding and sintering EU and HEU (3.5% and 93%) were performed 

in the Ceramics Fabrication Area. Chemical reprocessing of EU (1.8%) and 

coating of uranium particles HEU (93%) were in operation.

12/1960

HASL-114 

(AEC 1961b)

CP-2 chemical processing (3.4%) UO3 from UF6, CRP-2 chemical reprocessing 

(5.7%) U3O8 from uranium-zirconium, Ceramics Fabrication (5.7%) EU. Uranium

particle coating involved NU. No powder handling activities were evaluated.

5/1961

The air samples consisted of radioactive particulates collected on filters from breathing zones and 

general areas during processing. The alpha activity measured on the filter was used to determine the 

airborne alpha activity concentrations. When multiple samples at a location were collected, the AEC 

used the mean air concentration in subsequent calculations. The AEC matched air concentration 

determinations with information about worker categories, locations, tasks, and time at each location or 

task.

When estimating the intake for a specific worker, the dose reconstructor should look for all available 

information related to intakes of uranium. The information can include:

• Bioassay monitoring results,
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• Workplace breathing zone sampling results,

• General area monitoring results, 

• Work location and job classification by period,

• Air concentration information from HASL reports (summarized in this section and 

Attachment A),

• Reports to the AEC/NRC of overexposures to airborne activity, and

• Reports to the worker of overexposures and work restrictions.

The bioassay monitoring results and workplace breathing zone results provide the best information 

because the data relate to the exposed individual. The reports to the AEC/NRC and work restriction 

letters also relate directly to the individual. The information related to work location and job 

classification is useful to establish potential for intakes, and for correlation to the HASL air sampling 

results. The HASL results have been analyzed to determine statistical information on air 

concentrations, as presented in Attachment A. The analysis of the reported doses for all individuals 

from all HASL reports was based on the assumption of the data being represented by a lognormal 

distribution. The highest value (6,300 dpm/m3

) is assumed to represent the 95% value and the lowest 

value (7 dpm/m3

) is assumed to represent the 5% value. The resulting distribution has a median 

value of 210 dpm/m3 and a geometric standard deviation (GSD) of 7.91. This representation provides 

an estimate of air concentration that is more favorable to claimants than a strict numerical evaluation 

of the data. Several reported very high values are not well captured in a standard statistical analysis. 

Case files might contain information on breathing zone air monitoring for individuals. This information 

can be used to establish potential for exposure and estimates of intakes for specific work tasks. In 

general, bioassay monitoring data should be used to estimate intakes if available. 

The results in the HASL reports represent estimates of daily exposures to airborne uranium. These 

include specific work tasks that might have involved much higher air concentrations, but for a short 

period. Although it is not possible to identify in specific detail an individual worker’s occupancy versus 

uranium airborne exposures to estimate uranium intakes during the entire operational period for the

Apollo site (1957 to 1983), it is believed that the results of the HASL BZA results can give a 

reasonable upper bound intake value. The HASL studies resulted in improvements to working 

conditions, reducing the overall exposure of workers to airborne uranium.

The median value from the HASL reported average daily exposures (applied as a lognormal 

distribution with a GSD of 7.91) would provide a reasonable estimate of the distribution of uranium air 

concentrations during the 1960-through-1983 period for individuals whose specific work locations are 

not known. This intake should be limited to periods when the individual was likely to be involved with 

uranium work. It should not be applied to periods when the worker was on work restriction because of 

previous high exposures. This intake would represent an intake that is very favorable to claimants for 

individuals who did not work routinely in the Apollo uranium facility processes, such as health physics 

personnel, inspectors, and maintenance workers.

The inhalation intake of uranium can be estimated from the air concentration by multiplying by the 

breathing rate of 1.2 m3

/hour and the annual period of exposure (hours).

In the case where inhalation intakes are calculated from air concentrations, ingestion intakes are also 

to be considered. NIOSH (2004) states that the daily ingestion rate in picocuries can be estimated by 

multiplying the daily concentration in picocuries per cubic meter by a factor of 0.2 for an 8-hour

workday. 
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5.5 REPORTS OF OVEREXPOSURES AND INCIDENTS

NUMEC Health Physics reported to the AEC any time a radiation worker exceeded 40 MPC-hr in a 

workday or a workweek. Individual dosimetry records should indicate if a worker exceeded the 

40-MPC-hr limit. The records should also indicate if the overexposed individual was placed on work 

restrictions to limit internal and external radiation dose. This information would be useful in evaluation 

of bioassay data to indicate periods during which intakes might have occurred, and when intakes were 

unlikely. Individual dosimetry records should be used to reconstruct intakes on an individual basis 

whenever possible. 

Overexposures were required to be reported to the AEC/NRC. Overexposures were measured in 

terms of MPC-hours. If calculated MPC-hours exceeded 40 for a week, it was considered an 

overexposure. MPC-hours were related to inhalation of uranium or plutonium suspended in the air.

To protect the workers, half-face and full-face respirators were available and used during certain 

operations. NUMEC used routine nasal smears and bioassay samples as proof of protection. A 

nasal smear exceeding 100 dpm acted as a flag to indicate possible inadequate protection or potential 

misuse of a respirator; it was assumed that no protection was afforded by the respirator and a 

bioassay was conducted. If a high nasal smear coincided with an impermissible air sample, it was 

assumed that an overexposure had occurred and NUMEC reported it in compliance with 10 CFR 

Part 20 requirements. If a high nasal smear could not be corroborated by a high air sample, NUMEC 

reported only if the bioassay data indicated an overexposure (Shapiro 1969).

5.5.1 Apollo Site Incidents

The following incidents occurred at the NUMEC site in Apollo, Pennsylvania, and might have 

contributed to employee exposures, but no exposure information was provided in the incident report 

or was reported as permissible. Information about individual involvement in incidents is likely to be in 

the workers’ dosimetry records, and should be consulted for evaluation of intakes of radionuclides 

during such events.

In February 1963, a fire occurred when a polyethylene bottle containing recoverable powdered scrap 

uranium carbide stored under aqueous aluminum nitrate solution exploded from overpressure and the 

contents spontaneously ignited. Five bottles were damaged, containing about 8.8 kg of HEU. 

NUMEC estimated about 0.5 kg of HEU might have been lost. No information is provided about

worker exposures due to this incident (George 1963).

During an investigation of a ventilation problem in the CRP-1 process area, it was discovered that the 

CP-1/CRP-1 ammonia fume scrubber exhaust duct had become plugged with 18 in. of material. The 

material was found to be about 400 kg of dry 10% uranium by weight and 3.3% enriched in 235U; 

therefore, about 1.32 kg of 235U were present in the duct. The material was removed from the duct. 

Routine inspections of the ductwork were put in place and a HEPA filter installed (Reitler 1973b).

On April 20, 1974, a maximum of 6 kg of low-enriched UF6 was released to the in-plant atmosphere. 

A pipe and valve on the suction side of a hydrolysis column recirculating pump failed and blew out 

from the penton pipe, releasing the water from the bottom of the hydrolysis column, thereby releasing 

the UF6. Nasal smears were taken from all personnel involved, and all were within permissible limits 

(Fink 1974). 

5.5.2 Parks Township Site Incidents

The following incidents occurred at the NUMEC site in Parks Township and might have contributed to 

employee exposures. Information about individual involvement in incidents is likely to be in worker 
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dosimetry records and should be consulted for evaluation of intakes of radionuclides during such 

events. The following discussion is not intended to be a complete list of all incidents at the site.

In January 1966, a glovebox exploded when a worker struck a sparker to light a propane torch. The 

explosion blew out the glovebox and knocked the worker to the floor. He then ran from the room, with 

the box gloves still on his arms, spreading contamination. A nasal smear was taken, and he was 

decontaminated until the remaining activity was unremovable. Emergency personnel in protective 

clothing entered the room within minutes and extinguished the fire that had started in the damaged 

glovebox. All 150 personnel in the building at the time of the explosion were given nasal smears. 

Only the worker involved in the incident was found to have been seriously exposed (Caldwell, Potter, 

and Schnell ca. 1969).

In November 1966, an explosion in a glovebox resulted from a planned decomposition of hydrogen 

peroxide during a procedure test. The explosion shattered a glovebox window, sprayed the operator 

with nitric acid and plutonium, and caused physical injuries to a hand and an eye. After showering, 

the operator was transported to a local hospital for medical attention. The total uptake of plutonium 

was estimated to be less than 10 times the maximum permissible body burden (Crocker and 

Cleveland 1966; Caldwell, Potter, and Schnell ca. 1969). 

In January 1967, a technician cut open an iridium source in a glovebox that did not have the proper 

negative pressure gradient. The cutting operation resulted in airborne release of more than 75 Ci of 

192Ir as fine particles; the two technicians in the area received significant inhalation exposure. The 

incident was discovered about an hour after the release when the technicians went through a 

radiation monitor on their way to lunch. After showering, they were checked with a beta and gamma 

counter, and high levels were discovered in the chest area that could not be removed by scrubbing. 

Lung counts were taken, but the workers had to stand outside the lung counter to avoid swamping the 

detectors. Estimated lung doses were 14 and 45 rem for the two technicians (Caldwell, Potter, and 

Schnell ca. 1969).

In December 1967, a technician amputated his right hand while operating a milling machine in a 

plutonium glovebox. The accident occurred when the technician’s box glove was caught in the 4-in. 

cutter tool of a clausing milling machine. Contamination was limited to the severed wrist, and there 

was no significant release to the Plutonium Plant at large (Caldwell, Potter, and Schnell ca. 1969).

5.6 URANIUM AND PLUTONIUM AIR SAMPLING PROGRAMS

NUMEC uranium and plutonium workers wore lapel samplers starting in 1965. The primary purpose 

of air sampling was determination of personal exposure (Caldwell, Potter, and Schnell 1967). Sample 

duration using lapel samplers was one 8-hour shift at about 2 to 4 L/min. BZA sampling was 

performed during the HASL surveys from December 1959 to January 1961, and BZA sampling was 

observed in the 1963 health protection program review conducted by the AEC Oak Ridge Office R&D 

Division (Hervin and Prior 1963). NUMEC used a Rochester Imaging Detector Laboratory gas flow 

proportional counter.

Before 1965, the BZA samples were probably fixed-station BZA samplers; later NUMEC studies 

conducted in the 1966 to 1967 timeframe indicated that there was little difference between fixedstation BZA and GA samplers. The correspondence between lapel sampler data and early fecal 

clearance for plutonium showed very good agreement, but fixed-station BZA samplers and general 

area air sampling usually underestimated airborne concentrations. Fifty percent of the lapel air 

sample results at the Apollo site showed concentrations seven times greater than stationary air 

samples. The median of the ratio of lapel BZA to GA concentrations results was found to be ~7 at the 

Apollo and Parks Township sites (Caldwell, Potter, and Schnell 1967). 
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According to the 1963 NUMEC Health and Safety Manual, average or weighted airborne exposure 

studies were performed on every new operation and repeat studies were made on old operations on a 

frequent basis (NUMEC 1963). According to a health protection program review conducted in 1963, 

75 short-term breathing-zone air samples were obtained in Apollo process buildings every week but 

no routine general area air samples were taken that would indicate an average air concentration over 

an 8- to 24-hour period (Hervin and Pryor 1963).

Not all employees were assigned lapel samplers. Lapel samplers were used as a “diagnostic tool”

and provided to personnel whose work activities were likely to result in a local "micro-climate" of radioaerosol. Localized airborne exposure conditions existed during such activities as moving a 

contaminated beaker from one hood to another or working in a glovebox that had a pinhole leak in a 

glove (Caldwell, Potter, and Schnell 1967). During the NUMEC respirator effectiveness study (1966

to 1967), whenever a BZA sample indicated an exposure, the worker was removed from radiation 

work and fecal and urine samples were collected (Caldwell and Schnell 1968). 

The MPCs in the NUMEC Health and Safety Manual (NUMEC 1963) were 1 × 10-10 μCi/mL or 

220 dpm/m3 for in-plant airborne uranium.

6.0 OCCUPATIONAL EXTERNAL DOSE

This section describes NUMEC external dosimetry monitoring practices and provides supporting 

technical data to evaluate external occupational doses based on available dosimetry information. 

Although, DHHS has determined that there is insufficient information to either: (1) estimate the 

maximum radiation dose, for every type of cancer for which radiation dose are reconstructed, that 

could have been incurred under plausible circumstances by any member of the class; or (2) estimate 

the radiation doses to members of the class more precisely that a maximum dose estimate, at the 

Apollo and Parks Township sites during the operational period (Leavitt 2007, 2008), partial dose 

reconstructions can be completed using any external monitoring data in an individual’s file (and that 

can be interpreted using NIOSH dose reconstruction processes or procedures) for Parks Township

workers.

The NUMEC Health and Safety organization provided general radiological safety, criticality safety, 

instrumentation, and personnel dosimetry support to the Apollo nuclear fuel and the Parks Township 

plutonium facilities and associated operations, as presented in correspondence about resolution of 

safety issues at the sites (Caldwell 1967a, 1968c).

6.1 EXTERNAL EXPOSURE SOURCES

The Apollo Nuclear Fuel Facility started operations in 1957 with the small-scale production of HEU 

and LEU fuel. Between 1958 and 1983, the Apollo site produced LEU dioxide fuel for use in 

commercial nuclear reactors. The process consisted of conversion of UF6 to UO2. In 1963, an 

additional production line was added to produce HEU fuel for U.S. naval propulsion reactors. From 

1958 through the 1960s, NUMEC processed unirradiated EU scrap under license from the AEC 

(NIOSH 2007a). Smaller operations consisted of analytical laboratories, UO2 pellet production, and 

R&D into coating techniques for uranium particles (B&WNES 1997). HEU operations at the Apollo 

site were discontinued in 1978, and LEU and all other processing operations that involved radioactive 

materials had ended by the end of 1983. In the mid-1960s, NUMEC was involved in production of 

thorium oxide (ThO2) pellets for use in nuclear fuel. 

Parks Township site operations began in about 1959; DOE operations ended in 1980. The initial 

function of the Parks Township facilities was fabrication of plutonium fuel, preparation of HEU fuel, 

and production of zirconium-hafnium bars under AEC/NRC License SNM-414, received March 1961, 

which allowed the handling of plutonium already on the site. The Parks Township site made fuel for 

the FFTF that consisted of a mixture of PuO2 and depleted UO2. It also made fuel plates for the 
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ZPPR in the late 1960s and ZPPR-III fuel wafers. Activities included plutonium scrap recovery, DU 

fabrication, HEU fuel manufacturing, source manufacturing (primarily 60Co, 192Ir, PuBe, and AmBe), 

irradiated fuel sample examination, laboratory operations, and supporting nuclear power site 

operations. In 1980, B&W began dismantling the fuel fabrication lines to enable the area to be used 

for commercial decontamination and possibly LLRW volume reduction operations until the early 

1990s. In 1982, B&W used areas of the building for nuclear power site support operations. 

6.2 WORKPLACE RADIATION FIELDS

Occupational exposures were primarily associated with NUMEC activities with plutonium, thorium, 

and HEU to produce reactor fuel. Fissile material arrived in approved shipping and storage cylinders 

and was present in various forms (liquid, powder, or metal) to be converted for use in nuclear fuel. 

Available information indicates PuBe neutron source production was performed at the Parks 

Township site. There was some fission and activation product exposure (Caldwell and Judd 1966).

The primary sources of external radiation exposure from operations at NUMEC are summarized in 

Table 6-1.

Table 6-1. Workplace potential exposures.

Source Exposure potential

Plutonium fuel fabrication: 1959–1980 Gamma, X-ray, and neutron radiation primarily

HEU production: 1957–1978 Beta radiation primarily, possibly photon dose from 

uranium progeny such as radium, etc.

Source manufacturing Gamma and neutron radiation depending on source

LEU production: 1957–1984 As above

Mixed plutonium and EU fuel fabrication As above

HEU and LEU scrap recovery As above

UO2 pellet production started in 1961 As above

R&D for coating uranium particles started in 1961 As above

Thorium operations and pellet production started in 

1963

Beta radiation and more significant photon radiation

Laundry operations Uranium and thorium residues

6.2.1 Beta Radiation

Beta radiation associated with plutonium and thorium fuel operations is expected to be comparatively 

minimal. The beta dose rate for uranium operations such as on the surface of yellowcake (an NU 

compound) just after separation is negligible, but rises steadily thereafter due to the buildup of the 

238U decay products 234Pa and 234Th. A few months after chemical separation, when equilibrium is 

reached, the beta dose rate from yellowcake is about 150 mrad/hr. There would typically be mixed 

beta and photon radiation associated with fission and activation products.

6.2.2 Photon Radiation

Photon radiation, typically of lower energy, is characteristic of plutonium operations. Thorium emits 

significant higher energy photon radiation. Uranium has comparatively less significant photon 

radiation with dose rates of about 1.2 mrad/hr in contact with fresh yellowcake. However, during the 

buildup of the 234Th and 234Pa progeny in fresh yellowcake, the radiation levels increased somewhat 

for several months after yellowcake production. Photon exposure rates were estimated to be about

1 mrad/hr at 30 cm from a drum of aged yellowcake (NIOSH 2010b, Table 7).

6.2.3 Neutron Exposures

Neutron exposures might have occurred from both spontaneous fission in isotopes of uranium or 

plutonium and from alpha-neutron reactions with low atomic number materials such as oxides and 

impurities. Neutron exposures from plutonium occur and levels are generally described in the Guide 
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of Good Practices for Occupational Radiological Protection in Plutonium Facilities (DOE 1998). 

Neutron exposures from thorium and uranium such as yellowcake are considerably lower than the 

photon exposures and are, therefore, not generally considered significant based on analyses in 

ORAUT-OTIB-0024, Estimation of Neutron Dose Rates from Alpha-Neutron Reactions in Uranium 

and Thorium Compounds (ORAUT 2005a). That document describes the expected neutron dose 

rates from the various forms of uranium compounds. For a large cylinder of uranium hexafluoride, the 

dose rate at 3 ft is about 0.003 mrem/hr for NU, 0.016 mrem/hr for 5% EU, and 0.45 mrem/hr for 

+97% EU.

6.3 DOSIMETER TECHNOLOGY 

NUMEC historically used beta/photon and neutron dosimeters to measure potential WB beta/photon, 

WB neutron, and extremity beta/photon exposures to personnel. A summary of the NUMEC 

dosimetry systems and periods of use is presented in Table 6-2.

6.3.1 Beta/Gamma Dosimeters

Nuclear Science & Engineering, Controls for Radiation, Eberline, or Landauer provided film dosimeter 

services to NUMEC from 1959 until about 1968, when thermoluminescent dosimeters (TLDs) were 

implemented. There is evidence of NUMEC concern about film dosimetry over-response to the lowenergy photons from plutonium (Caldwell and Judd 1966). Landauer began providing dosimeter 

service to NUMEC in 1964. Eberline provided dosimetry service beginning in 1966, and NUMEC 

apparently ran an in-house TLD program beginning in about 1968. The dosimetry service was again 

provided by Landauer beginning in 1976. External results were found for July to December 1991 with 

dosimetry provided by Teledyne Isotopes (BWXT 2006).

6.3.2 Neutron Dosimeters

Workers were monitored for neutron exposures with nuclear track emulsion, Type A (NTA) film from 

commercial venders until about 1968, and with TLDs thereafter. In addition, criticality dosimetry 

monitoring was done with an array of area critical assemblies that fed into a central system. This 

system existed from at least 1963; in September 1963, each visitor and employee was issued an 

indium foil criticality dosimeter as part of each security badge (NUMEC 1963).

6.3.3 Limits of Detection

External dosimetry technology minimum detectable levels (MDLs) are expected to be similar to 

contemporary commercial vendor capabilities. Examination of dose reports for individual dosimeter 

exchange periods and workers shows recorded doses as low as 2 mrem for photons (Boyd 2006a, 

p. 6), which is certainly less than a statistically based MDL. However, other documentation indicates 

that film dosimeter MDLs in the workplace are higher. The film badge dosimetry at NUMEC was likely 

similar to dosimeters used at Hanford during the period 1957 to 1968 (ORAUT 2010c). Therefore, the 

recommended MDLs for estimation of missed dose are 30 mrem for gamma and beta radiation

through 1968, and 50 mrem for neutron radiation for periods to 1975 (ORAUT 2010c,d). For neutron 

radiation after 1975, the MDL is reduced to 20 mrem based on studies at the Hanford Site (Fix et al. 

1981) and Savannah River Site (SRS) (Taylor et al. 1995) that indicated the MDL was closer to 10 

mrem for neutron exposures to fast neutrons from 252Cf. The 20-mrem value is consistent with the 

SRS site profile (ORAUT 2005b) value for this period because work with plutonium was similar to that 

at the NUMEC Plutonium Facility. For estimating the potential annual missed dose in accordance 

with OCAS-IG-001 (NIOSH 2007b) for monitored workers, Table 6-3 summarizes the annual potential 

missed dose to be assigned in relation to the dosimetry service providers, periods of use, dosimeter 

exchange frequencies, and estimated MDLs.
Document No. ORAUT-TKBS-0041 Revision No. 02 Effective Date: 11/26/2012 Page 53 of 89

Table 6-2. Dosimetry for external whole-body, wrist, and extremity exposures.

Period Monitoring technique Dosimeter description

Beta/photon dosimeters – whole body

1957–5/1968 Photographic film badge Film badges contained single film packet. Three filters (front 

and back) were incorporated into film badge for energy 

dependence: cadmium, aluminum, and lead. Beta and 

photon radiation capabilities are similar to other dosimetry 

systems at that time as presented in the 1954 AEC 

dosimeter performance study (AEC 1955).

6/1968–1975 NUMEC or Eberline TLD Comprised of 2 TLD-700 chips, 2 TLD-600 chips, and 1 

CaF2 for monitoring beta, X-ray, and gamma exposure.

1976–present Landauer or equivalent TLD

(Z1 dosimeter -1990)

Comprised of 3 TLD-700 chips for monitoring beta, X-ray, 

and gamma exposure. Insensitive to neutron radiation. 

Beta/photon dosimeters – wrist and ring

July 1963–about 

5/1968

Landauer (Type M – wrist 

beta-gamma) film badges or 

equivalent.

Film dosimeter known as Type M responsive to beta and 

gamma radiation.

About 6/1968– 1983 TLD wrist badge Comprised of 3 TLD-100 chips.

7/1991 – 12/1991 Teledyne Isotopes TLD 

Badge

TLD badge for monitoring beta and gamma exposure 

(BWXT 2006). Details of the dosimeter are not available, 

other than detection limits.

Neutron dosimeters – whole body

1957–5/1968 NTA film badge Film badges using NTA films: Fast neutrons undergoing 

elastic collision with content of emulsion or cellulose acetate

base material produce recoil protons, which are recorded as 

photographic tracks in emulsion. Track density is a linear 

function of dose. Developed image exhibits tracks caused 

by neutrons, which can be viewed using appropriate 

imaging method (i.e., oil immersion) and 1000-power 

microscope or projection capability.

6/1968–1995 Landauer Neutrak Extended 

Range dosimeter (types I8, 

I1, or RI)

Combined TLD albedo neutron monitor with track recoil 

device [CR-39 (allyl diglycol carbonate)] that responds to 

neutron radiation through proton recoil events. The 

dosimeter is responsive to a neutron energy range of about

0.0001 to 10 MeV. Dosimeter response to thermal neutron 

radiation was subtracted to yield fast neutron dose. The 

Neutrak ER has an albedo element with above-described 

elements. Qualitative relationship was derived to determine 

ratios of neutrons of various energies. The RI badge was 

capable of monitoring beta, X-ray, gamma, and neutrons. 

7/1991 – 12/1991 Teledyne Isotopes TLD 

badge

Combined gamma, beta, and neutron TLD (BWXT 2006). 

Details of the dosimeter are not available, other than 

detection limits.

Source: ORAUT (2011b).

Performance of dosimetry technology at many commercial and AEC laboratory service providers was 

tested in 1954 by the AEC (AEC 1955). Characteristics of dosimetry systems at the NUMEC sites are 

listed in Table 6-2 for beta, gamma, and neutron radiation monitoring.
Document No. ORAUT-TKBS-0041 Revision No. 02 Effective Date: 11/26/2012 Page 54 of 89

Table 6-3. MDLs and potential missed photon, beta, or neutron dose.

Vendor or processor/area monitored Period of use MDLa (rem)

Annual missed doseb (rem)

(frequency in parenthesis)

Nuclear Science & Engineering or 

Controls for Radiation film and NTA film

whole body

1957–1963 0.03c photons 0.18 beta photons (monthly)

0.03c beta

0.05 neutrons 0.30 neutrons [fast] (monthly)

Landauer film and NTA film 

whole body

1964–1965 0.03c photons 0.18 beta photons (monthly)

0.03c beta

0.05 neutrons 0.30 neutrons [fast] (monthly)

Eberline film and NTA film 

whole body

1966–6/1968 0.03c photons 0.18 beta photons (monthly)

0.03c beta

0.05 neutrons 0.30 neutrons [fast] (monthly)

NUMEC or Eberline (Film-Apollo) until 

about 1970 and TLD for neutron

whole body

7/1968–1975 0.02 photonsd 0.12 beta photons (monthly)

0.02 betad

0.05 neutrons 0.30 neutrons (monthly)

Landauer TLD and Teledyne Isotopes

whole body

1976–1995 0.01 photonse 0.06 beta photons (monthly)

0.01 betae

0.02 neutrons 0.12 neutrons (monthly)

a. Estimated MDLs for each dosimetry technology. Dose levels were recorded at values less than the MDLs.

b. Annual missed dose calculated based on the MDL/2 method from NIOSH (2007b).

c. MDLs for photons and beta during these periods are based on Hanford dosimeter values (ORAUT 2010c).

d. MDL during this period is probably twice the recording level of 0.010 rem.

e. Landauer MDL values from Boyd (2006j) and Teledyne Isotopes MDL values from BWXT (2006).

6.3.4 Radiological Records

A single dosimetry program was conducted at NUMEC. Records of radiation doses to individual 

workers from personnel dosimeters worn by the worker and coworkers are available for NUMEC 

operations beginning in 1957 for Apollo and 1959 for Parks Township as observed from a review of 

claimant records. Doses that were received by these dosimeters were recorded at the time of 

measurement and routinely reviewed by the NUMEC operations and radiation safety staff for 

compliance with radiation control limits. OCAS-IG-001 (NIOSH 2007b) indicates that these represent 

the highest quality records for retrospective dose assessments. Not all workers were assigned 

radiation dosimeters. Workers who received less than 25% of the quarterly dose limits in 10 CFR 

Part 20 were not required to be monitored (Boyd 2006a, p. 6). However, even though claimant 

records show that not all personnel were assigned dosimeters at all times, the records show that work 

areas were monitored. 

Substantial worker-specific dose data have been received from NUMEC. Shallow, deep, neutron, and 

extremity doses are typically available. A computerized records system was implemented in October 

1975 (Boyd 2006a, p. 7), and records for previous years are in hard-copy form. In addition, NUMEC 

was required to submit routine dose reports of personnel exposure information to the AEC and NRC 

for terminating employees (Boyd 2006f) as well as annual statistical data, such as those listed in 

Table 6-4 for 1976 and 1977 (Breuer 1977, 1978).

Table 6-4. Annual occupational radiation exposures at the Apollo site (Breuer 1977, 1978).

Year

Total 

number 

monitored

No. with 

measured 

dose

Number of individuals with WB doses in the ranges (rem)

<0.1 0.1–0.25 0.25–0.5 0.5–0.75 0.74–1.0 1.0–12 >12.0

1976 42 42 27 14 1 0 0 0 0

1977 39 39 15 16 6 0 2 0 0
Document No. ORAUT-TKBS-0041 Revision No. 02 Effective Date: 11/26/2012 Page 55 of 89

6.4 LIMITATIONS IN MEASURED DOSE

Potential limitations in measured dose with NUMEC dosimetry capabilities include low-energy photons 

and neutron radiation. 

6.4.1 Low-Energy Photons

Experience at NUMEC since the mid-1960s indicates there are potential limitations of the film 

dosimeter to measure accurately low-energy photon radiation such as that in NUMEC plutonium 

facilities. Caldwell and Judd (1966) indicated that photon radiation from plutonium could be 

considered to be in three effective energy groups: 

• 17-keV X-rays that had a low penetrating ability,

• Effective energy of 60 keV from plutonium and its progeny including 241Am, and

• Effective energy of 400 keV. 

A spectroscopy survey at the Parks Township site evaluated photon fields from plutonium work. 

Surveys were conducted of the plutonium chemical processing line and ceramics line. The 60-keV 

peak from 241Am was found to predominate. The 17-keV X-rays did not produce a peak and must 

have been substantially absorbed by the glovebox walls (Caldwell and Judd 1966, p. 4). Higher 

energy peaks at 208, 267, and 333 keV were produced by 237U. The photon energy spectrum is 

shown in Figure 6-1.

A survey of the ceramics line and plutonium-uranium molybdenum alloy melt box line indicates a 

predominance of 60-keV 241Am gamma radiation. The gamma energy spectrum is shown in 

Figure 6-2. However, the relative amount of 60-keV radiation is a factor of 6 higher for the ceramics 

glovebox in comparison with the melt box. This is attributed to the plutonium in the chemical 

processing line having aged an additional 2 years, whereas the plutonium in the melt box had just 

been received (Caldwell and Judd 1966).

An analysis of the Eberline film dosimeter response for open window (OW) versus aluminum (AL),

plastic (PL), and cadmium (CD) filters was made as shown in Figure 6-3.
Document No. ORAUT-TKBS-0041 Revision No. 02 Effective Date: 11/26/2012 Page 56 of 89

Figure 6-1. Gamma spectrum external to 

plutonium chemical processing line and plutonium 

ceramics line (Caldwell and Judd 1966, Figure 3).

Figure 6-2. Gamma spectrum external to 

plutonium ceramics line and plutonium melt 

glovebox (Caldwell and Judd 1966, Figure 4).

 Figure 6-3. Dosimeter filter ratios (Caldwell and Judd 1966).

The following interpretation was made by Caldwell and Judd (1966):

1. Hot cell workers are exposed to 60Co and FPs. The energy response of the film badge is 

constant above 200 keV. The dose was taken directly from a 60Co calibration curve. If the 

OW:CD ratio was close to 1.0, the reported dose was accepted.

2. Plutonium workers were exposed to a wide range of gamma energies. The upper end of the 

spectrum would produce OW:CD ratios close to 1.0. The lower gamma energies would 

produce an OW:CD ratio of 3.0 or greater. One plutonium worker might be exposed to an 

entirely different effective energy than another due to shielding, working distance, and other 
Document No. ORAUT-TKBS-0041 Revision No. 02 Effective Date: 11/26/2012 Page 57 of 89

geometry factors. If the OW:CD ratio was less than 2.0, the reported dose was accepted. If 

the OW:CD ratio was greater than 2, NUMEC would use a plutonium spectrum calibration 

curve that represented a typical plutonium gamma spectrum.

3. The OW:PL (plastic) ratio was about 1 and was within a 6% standard deviation. This meant 

that the large OW:CD ratio is not due to beta radiation. 

4. The OW:AL ratio was sensitive to X-ray exposures, but NUMEC did not incorporate this in its

analysis.

Caldwell and Judd (1966) presented an assessment of the energy dose fraction for personnel 

exposure due to typical plutonium fuel fabrication from 1,000 MWd/t of plutonium. Sixty-five percent 

of the dose was from the 241Am 60-keV gamma. Less than 7% was from the highest energy groups 

(Caldwell and Judd 1966). The summary of the energy dose fraction is shown in Figure 6-4. The

gonadal dose was 50% of the WB or trunk dose due to the effect of the steel bottom of the plutonium 

gloveboxes (Caldwell and Judd 1966). Table 6-5 summarizes the gamma energy distribution for 

NUMEC plutonium in comparison with Hanford plutonium. Beta energies are included as well as 233U 

and 241Am, which have similar overall photon and beta properties.

6.4.2 Neutron Radiation

NTA film has a characteristic decreasing response to neutron radiation at energies below about 500 to 

800 keV, depending on the extent of photon fogging and the overall process to develop and read the 

tracks (ORAUT 2006). However, at this time, the neutron dosimeter readings should be used without 

correction for this effect. 

 Figure 6-4. Relative contribution by energy group to

 personnel exposure during plutonium fuel fabrication from 

 1,000 MWd/t plutonium (Caldwell and Judd 1966, p. 18, 

 Figure 8).

Table 6-5. Plutonium photon (and beta) energy factors.

Energy–photon NUMEC plutonium, (241Am & 233U) Hanford plutonium

<30 keV 15 25

30 – 250 keV 82 75

>250 keV 3 0

Energy–beta NUMEC plutonium, (241Am & 233U) Hanford plutonium

>15 keV 100 100

0

20

40

60

80

Fraction of 

Dose

20 kev60 kev 115

kev

195

kev

335

kev

Energy in kev

 

 

 
Document No. ORAUT-TKBS-0041 Revision No. 02 Effective Date: 11/26/2012 Page 58 of 89

6.5 DOSE RECONSTRUCTION RECOMMENDATIONS

6.5.1 Recorded Dose Practices

Recorded and reported dose practices are summarized in Tables 6-6 and 6-7. 

Table 6-6. Recorded dose practices.

Period Dosimeter measured quantities Compliance dose quantities

Photon/electron film dosimeter + NTA neutron dosimeter

1957–1971 Gamma (G) 

Neutron (N) 

Beta (B)

WB or total = gamma (photon) + neutron

Beta separate

Extremity = gamma (+ neutron)

Photon/electron film dosimeter + TLD neutron dosimeter

1972–1983 Deep = gamma and neutron (DBG)

Shallow beta gamma (SBG)

WB = gamma + neutron

Skin = beta 

Extremity = gamma + neutron

Photon/electron/neutron–Panasonic TLD + CR-39 neutron dosimeter

1983–present Deep

Shallow

Skin = beta + soft gamma and neutron 

WB = photon + neutron 

Extremity = gamma + neutron

6.5.2 Adjustments to Recorded Dose

6.5.2.1 Beta Dose Adjustments

Beta and nonpenetrating dose was usually reported before 1975. In general, nonpenetrating radiation 

doses should be assigned as <30-keV photons if the employee worked with or around plutonium; 

otherwise, >15-keV electrons (beta) should be assigned (ORAUT 2005c). 

Table 6-7. Interpretation of reported data.

Period

Reported 

quantity Description

Interpretation of 

zeroes

Interpretation

of blanks (no 

data)

Rollup of 

individual and 

annual data

Monitored/ 

unmonitored

1957–

1971

R or rem Reported WB 

doses 

include 

gamma and 

neutron 

doses

Zeroes were 

generally not 

reported. Reported 

zeroes should be 

interpreted as 

meaning less than 

MDL.

The absence of 

data should be 

interpreted as 

individual was 

monitored with 

zero result.

Photon WB 

dose, neutron 

WB dose, 

shallow skin 

dose, total 

deep WB dose

All employees 

with significant 

exposure 

potential were 

monitored

1972–

1982

rem Reported WB 

doses 

qualified as 

either photon 

or neutron

Zeroes were 

generally not 

reported. Reported 

zeroes should be 

interpreted as 

meaning less than 

MDL.

The absence of 

data should be 

interpreted as 

individual was 

monitored with 

zero result.

Photon WB 

dose, neutron 

WB dose, 

shallow skin 

dose, total 

deep WB dose

All employees 

with significant 

exposure 

potential were 

monitored

1983–

present

rem Photon deep, 

neutron 

deep, and 

skin dose 

reported.

Zeroes were typically 

reported. Reported 

zeroes should be 

interpreted as 

meaning less than 

MDL.

No data or 

blanks should 

be interpreted 

as individual 

was monitored 

with zero result.

Photon WB 

dose, neutron 

WB dose, 

shallow skin 

dose, total 

deep WB dose

All employees 

with significant 

exposure 

potential were 

monitored

The guidance from ORAUT (2005c) is as follows:
Document No. ORAUT-TKBS-0041 Revision No. 02 Effective Date: 11/26/2012 Page 59 of 89

If the nature of the nonpenetrating dose is unknown, consider the following guidance:

1. For a likely noncompensable case, it is acceptable to assume the nonpenetrating dose is 

associated with <30-keV photons, because this maximizes the probability of causation (POC).

2. For a likely compensable case, it is acceptable to assume the nonpenetrating dose is 

associated with >15-keV electrons, because this minimizes the POC.

3. If the compensability decision might hinge on this issue, and if the partitioning of the 

nonpenetrating dose cannot be decided based on the available information, additional 

research might be required.

6.5.2.2 Photon Dose Adjustments

No adjustment is recommended for NUMEC recorded shallow and deep doses and photon radiation. 

The existing recorded doses provide a realistic estimate of the actual doses. Exposure dose 

conversion factors are applied to reported and missed photon doses to estimate dose to specific 

internal organs.

6.5.2.3 Neutron Weighting Factor Adjustments

Recorded NUMEC neutron doses are assumed to have been based on quality factors in National 

Council on Radiation Protection and Measurements Report 38 (NCRP 1971). The quality factors in 

Report 38 were compared with the neutron weighting factors in ICRP Publication 60 (ICRP 1991) to 

arrive at factors to convert the recorded dose to equivalent ICRP Publication 60 neutron doses as 

required by OCAS-IG-001 (NIOSH 2007b). A dose multiplier of 1.91 should be used for the 0.1- to 

2-MeV energy range (ORAUT 2006). This range includes 100% for HEU, EU, NU, and plutonium 

work locations. The neutron doses from fuel work recorded for personnel at NUMEC using NTA 

should, therefore, be multiplied by a factor of 1.91 (for ICRP Publication 60 correction). 

6.5.3 Missed and Unmonitored Dose

The potential for missed dose exists when workers are exposed to radiation at levels below the 

detection limit of their personnel dosimeters.

6.5.3.1 Shallow Dose and Deep Dose

The assignment of missed dose based on dosimetry records is performed using guidance in OCAS, 

IG-001, External Dose Reconstruction Implementation Guideline (NIOSH 2007b). Using this 

guidance, a dose equal to the limit of detection (LOD) divided by 2 is assigned for each dosimetry 

measurement that is recorded as less than the LOD/2 including zero values. The LOD values for 

NUMEC dosimeters are listed in Table 6-3. 

For cases involving the skin as the target organ, guidance in ORAUT-OTIB-0017, Interpretation of 

Dosimetry Data for Assignment of Shallow Dose (ORAUT 2005c) should be followed for assignment 

of missed shallow and deep doses. 

6.5.3.2 Neutron Dose

The potential missed neutron dose can be estimated from LOD values in Table 6-2 for monitored 

workers using the same approach. 

If monitoring records do not include neutron dosimetry information, then neutron dose should not be 

applied. For workers who were likely to be exposed to neutrons, and for whom no neutron dosimetry 
Document No. ORAUT-TKBS-0041 Revision No. 02 Effective Date: 11/26/2012 Page 60 of 89

is available, a partial dose reconstruction would result. Exposure to uranium hexafluoride cylinders is 

a possible source of neutron exposure at the Apollo site. 

6.5.4 Uncertainty

Dose reconstructors can incorporate consideration of uncertainty in the dose calculation for measured 

and missed doses as follows:

• The technology used to measure worker dose at NUMEC is similar to the technology that was 

used by commercial and AEC/NRC laboratory facilities. The errors in the penetrating dose are 

anticipated to be about ±30% and normally distributed. For noncompensable cases, the dose 

reconstructor can assume that errors are all positive (i.e., use only +30%) and multiply the 

measured dose by a factor of 1.3 (i.e., increase of 30%) to be used for Interactive 

RadioEpidemiological Program (IREP) Parameter 1 and to set Parameter 2 to zero (NIOSH

2007b). A constant distribution is applied.

• For missed dose, a lognormal distribution is assumed. Dose reconstructors should calculate 

the unmonitored dose or missed dose to arrive at Parameter 1 input and to set Parameter 2 

equal to 1.52 (NIOSH 2007b). A lognormal distribution is applied.

6.5.5 Radiation Dose Fraction

Uranium represents the primary exposure hazard to NUMEC workers. Naturally occurring uranium is 

primarily a beta radiation hazard with an accepted surface dose rate of about 233 mrad/hr. The IREP 

input category for beta radiation is >15 keV. There is a small photon dose component of <10 mrem/hr 

(DOE 2001). As naturally occurring uranium is enriched, the photon dose is lowered but the spectra 

become correspondingly more energetic. The average energy of the spectra can increase from solid 

or liquid uranium sources because these can provide substantial shielding resulting in proportionally 

greater attenuation of lower energy photons. Exposure to thin layers of uranium on a surface will 

have a higher proportion of lower energy photons. The recommendation is to assign the photon dose 

as 100% to the 30-to-250-keV category to result in a higher calculated organ dose under most 

situations. 

Apollo site workers had limited potential for exposure to radioactive sources in addition to uranium. 

These include thorium, plutonium, and photon sources such as radium, 192Ir, 137Cs, and 60Co. Mixed 

FP exposure could have occurred at the Apollo Laundry Building that provided commercial laundering 

of contaminated clothing. Generally recommended categories for IREP input for the measured and 

assigned components of radiation dose are listed in Table 6-8 unless there is claim-specific 

information on the source of radiation exposure.

Table 6-8. Beta, photon, and neutron radiation energies and percentages for IREP input. 

Description

Dates Radiation 

type

Energy 

Begin End selection Percent

Uranium facilities 1/1/1957 12/31/1983 Beta >15 keV 100

1/1/1957 12/31/1983 Photon 30–250 keV 100

Plutonium facilities 1/1/1959 12/31/1980 Photon 30–250 keV 100

1/1/1959 12/31/1980 Neutron 0.1–2 MeV 100

Thorium handling 1/1/1957 12/31/1983 Beta >15 keV 100

1/1/1957 12/31/1983 Photon 30–250 keV 25

1/1/1957 12/31/1983 >250 keV 75

Photon and neutron sources 1/1/1957 12/31/1983 Beta >15 keV 100

1/1/1957 12/31/1983 Photon 30–250 keV 50

1/1/1957 12/31/1983 >250 keV 50

1/1/1957 12/31/1983 Neutron 0.1–2 MeV 100
Document No. ORAUT-TKBS-0041 Revision No. 02 Effective Date: 11/26/2012 Page 61 of 89

7.0 ESTIMATION OF EXPOSURE TO RESIDUAL ACTIVITY

The Apollo site stopped manufacturing nuclear fuel in 1983. Final decommissioning of the facilities 

was completed in 1995. For the period of residual contamination, employees of subsequent owners 

and operators of this facility are covered under EEOICPA. The residual period for the Apollo site

covers the period from 1984 through 1995, and the residual period for the Parks Township site covers 

the period from 1981 through 2004.

The uranium work at Parks Township was with HEU in the Type II facility (Building C). The equipment 

in this building was removed in 1978, and by May 1979 the remaining surface contamination was 

fixed and inaccessible to diversion. The effluent reports after decontamination indicate contaminated 

liquid effluents from residual material in drains but no airborne emissions. This would suggest there is 

a potential for external exposure from residual DOE material, but inhalation exposure would be 

minimal. 

The plutonium facility at Parks Township (Building A) was decontaminated and the equipment 

removed in 1980. However, the building continued to be used for non-DOE activities. Residual 

activity (from DOE operations) could have remained and caused exposure to workers. There was 

probably not much use of the buildings after the late 1980s, but NUMEC was licensed to have nuclear 

material on the site until final decommissioning was approved in about 1998. The license probably

was kept in place to cover residual activity.

The following sections provide guidance for assignment of dose for the residual period.

7.1 EXTERNAL DOSE FROM RESIDUAL ACTIVITY IN THE WORKPLACE

7.1.1 Apollo Site

Because all operations at the Apollo site had stopped by 1983, the only exposures would be from 

residual activity. The potential external dose during the residual period has been estimated based on 

the mean surface concentration at the end of the operational period as described in Section 7.4, 

derived in support of internal dose estimates during the residual period. The analysis resulted in a 

mean surface concentration of 4.97 × 106 dpm/m2 (GSD = 7.91). This value can be used to estimate 

the annual external dose to workers exposed to the residual activity. The annual dose is evaluated as 

follows, using a dose conversion factor (DCF) for exposure to uniform activity on a ground plane.

Dose (rem/yr) = residual level (dpm/m2

) × DCF (rem/dpm/m2/hr) × exposure time (hr/yr) (7-1)

The external dose conversion factor for exposure to isotopes of uranium and short-lived progeny is 

provided in Federal Guidance Report 12 (Eckerman and Ryman 1993). The median annual external 

dose from exposure to residual surface contamination is listed in Table 7-1 for organs considered in 

the Federal Guidance Report. These values are based on an exposure time of 2,000 hours per year. 
Document No. ORAUT-TKBS-0041 Revision No. 02 Effective Date: 11/26/2012 Page 62 of 89

Table 7-1. External annual dose from residual 

surface contamination at the Apollo site.

Organ Annual dosea rem

Adrenals 0.001

Bladder wall 0.001

Bone surface 0.002

Brain 0.001

Breast 0.001

Esophagus 0.001

Stomach wall 0.001

Small intestine wall 0.001

Upper large intestine wall 0.001

Lower large intestine wall 0.001

Kidneys 0.001

Liver 0.001

Lungs 0.001

Muscle 0.001

Ovaries 0.001

Pancreas 0.001

Red marrow 0.001

Skin 0.276

Spleen 0.001

Testes 0.001

Thymus 0.001

Thyroid 0.001

Uterus 0.001

a. Values are input into IREP as a lognormal distribution;

value listed is the geometric mean; the GSD is 7.91.

The dose values are for exposure to NU because this provides a higher external dose than other 

enrichments (except DU). This provides a dose estimate that is favorable to claimants because most 

uranium at the Apollo uranium facility was NU or EU. NU provides a higher external dose, per unit 

activity, because significant contributions come from the short-lived progeny of 238U (234Th and 

234mPa).

The skin dose value represents the dose at 1 meter above the ground. This provides an overestimate 

of dose to the skin for cancers above the waist and an underestimate of dose for cancers below the 

waist. 

The dose should be entered into the IREP input as a lognormal distribution with a GSD of 7.91 (IREP 

Parameter 2) as photons of energy from 30 to 250 keV as indicated in Table 6-10. This provides a 

favorable estimate of POC for all organs, even though some of the photon energy is likely to be of 

higher energy. 

7.1.2 Parks Township Site

Because the residual period for the Parks Township site includes the period of 1981 – 1983, when the 

Apollo site was still in operation, it is possible that Parks Township workers were exposed to DOE 

work if they visited the Apollo site. If dosimeter readings are available for these years of the residual 

period, the dose should be based on the recorded and missed dose (Section 6.0) unless it is known 

that the worker did not visit the Apollo site. External dosimetry data might exist during later years of 

the residual period; however, operations were not part of the nuclear weapons-related program, and 

are not covered for dose reconstruction under EEOICPA (Leiton 2011). 
Document No. ORAUT-TKBS-0041 Revision No. 02 Effective Date: 11/26/2012 Page 63 of 89

Work at the Parks Township site included work with plutonium in Building A and work with HEU in 

Building C. Estimates of external dose from these two materials are described below. If it is not 

known where the energy employee worked during the residual period, the dose based on uranium 

exposures can be assigned as a favorable to claimant assumption. 

An estimate of the external dose from residual plutonium activity at the Parks Township site has been 

made using the same approach used for the Apollo site uranium external dose. The mean surface 

concentration at the end of the operational period as described in Section 7.4, derived in support of 

internal dose estimates during the residual period, was 2.74 x 105 dpm/m2 (GSD = 4.97). This value 

can be used to estimate the annual external dose to workers exposed to the residual activity. The 

annual dose is evaluated using Equation 7-1, using DCFs for exposure to uniform activity on a ground 

plane. The annual dose was evaluated using the plutonium isotopic compositions given in Table 5-3 

as a function of material age. The highest dose was obtained for the 20-year aged material for all 

organs. However, all doses were less than 0.001 rem. Therefore, no external dose need be assigned 

during the residual period for Parks Township workers based on potential exposure to plutonium for 

Building A.

The work at the Parks Township site included work with HEU in Building C. No information has been 

found related to air concentrations in Building C during the operating period. Because the Apollo 

uranium operations covered a longer period than the Building C operations, and because high levels 

of airborne uranium were present during the Apollo operations (AEC 1960a,b,c, 1961 a,b), the 

residual activity levels for the Apollo site provide a favorable to claimant estimate of external dose 

from residual activity at the Parks Township HEU operations in Building C. The annual external dose 

values for the Parks Township residual period are the same as those given in Table 7-1.

7.2 EXTERNAL AMBIENT DOSE FROM RESIDUAL ACTIVITY 

All unmonitored workers are assigned external dose as described in Section 7.1. The assigned 

external dose would cover any additional ambient external dose, and the assignment of ambient dose 

is not necessary.

7.3 OCCUPATIONAL MEDICAL DOSE

During the residual period, medical X-ray doses are not to be included in the dose reconstruction 

because the work is not directly related to DOE employment.

7.4 INTERNAL DOSE FROM RESIDUAL ACTIVITY IN THE WORKPLACE

7.4.1 Apollo Site

The following information provides a method for estimating exposures during the residual radiation 

period due to uranium contamination. Valid bioassay data are unlikely to be available during the 

residual period. However, monitoring data can be used instead of the default intake assumptions 

given below to limit dose, as appropriate (NIOSH 2008b).

Guidance in Dose Reconstruction During Residual Radioactivity Periods at Atomic Weapons 

Employer Facilities (ORAUT 2012b) describes methods to estimate intake of radionuclides during 

residual periods. The method relates residual surface contamination to workplace air concentration 

during periods after completion of DOE work.

The majority of the DOE-related work at the Apollo site was involved with scrap recovery and fuel 

fabrication operations in the Apollo uranium facility (East Bay of the Main Building). The HASL reports 

documented the average workplace air concentration to which workers were exposed during 1960 

and 1961 (AEC 1960a,b,c, 1961a,b). The results of these studies are described in Section 5. The 
Document No. ORAUT-TKBS-0041 Revision No. 02 Effective Date: 11/26/2012 Page 64 of 89

resulting mean average daily air concentration was estimated to be 210 dpm/m3 with a GSD of 7.91 

(assuming a lognormal distribution). 

The mean average daily air concentration can be used to estimate the residual surface concentration, 

using guidance from ORAUT (2012b). The annual deposition amount is estimated using a deposition 

velocity of 0.00075 m/s, with deposition assumed to occur for 1 year. Using this approach, a surface 

concentration of uranium is estimated as follows.

210 dpm/m3 × 31,536,000 s/yr × 0.00075 m/s = 4.97 × 106 dpm/m2

This mean surface concentration is described as a level at the end of the operating period that is 

favorable to claimants. The deposited material is assumed to be resuspended and inhaled during the 

residual period. The amount of resuspension is assumed to be reduced with time due to fixing of the 

material on surfaces and to depletion (ORAUT 2012b). The depletion factors applied to each year are 

listed in Table 4-2 of ORAUT (2012b). The depletion factors indicated for the residual concentration 

at the end of the operational period should be used for the first year; the remaining years should be 

reduced by factors listed in Table 4-2 of ORAUT (2012b).

The air concentration for each year is estimated using a resuspension factor (ORAUT 2012b) of 

1 × 10-6

/m. Application of this resuspension factor and the above-described depletion factors to the 

residual contamination level of 4.97 × 106 dpm/m2

, results in the air concentration and annual intakes 

in Table 7-2. The intake evaluation is based on exposure for 2,000 hr/yr and an inhalation rate of 

1.2 m3

/hr.

Table 7-2. Uranium air concentration and annual intake in 

the Apollo site residual period.

Year

Air concentration 

(dpm/m3)

Inhalation

Intake 

(dpm/yr)

Ingestion 

Intake 

(dpm/yr)a

1984 4.97 1.19E+04 10,500

1985 3.89 9.33E+03 8,200

1986 3.04 7.31E+03 6,400

1987 2.38 5.72E+03 5,000

1988 1.87 4.48E+03 3,900

1989 1.46 3.50E+03 3,100

1990 1.15 2.75E+03 2,400

1991 0.90 2.16E+03 1,900

1992 0.70 1.68E+03 1,500

1993 0.55 1.32E+03 1,200

1994 0.43 1.03E+03 910

1995 0.34 8.09E+02 710

a. Doses are assigned as a lognormal distribution with a GSD of 7.91.

The intakes in Table 7-2 can be used to estimate the internal dose to the target organ for the years of 

employment for the worker. The estimated internal doses are assigned as a lognormal distribution 

with a GSD of 7.91, corresponding to the distribution of the average daily air concentrations used to 

estimate the annual uranium intake. The uranium intake is represented as 234U in the dose estimate. 

The dose should be evaluated for the three uranium material solubility types of F, M, and S, with the 

dose from the highest type used in the IREP input. 

Although uranium-aluminum alloy was present at the Apollo scrap recovery facility, the form of the 

material was not likely to be an inhalation hazard and modeling an intake of uranium aluminide is not 

necessary for the Apollo site. The internal dose analysis should include the potential inadvertent 
Document No. ORAUT-TKBS-0041 Revision No. 02 Effective Date: 11/26/2012 Page 65 of 89

ingestion of uranium activity, based on guidance in OCAS-TIB-009, Estimation of Ingestion Intakes

(NIOSH 2004).

The daily intake rate (dpm/d) is estimated to be 0.2 times the average daily air concentration 

expressed in units of dpm/m3

. Using the air concentration at the end of the operating period of 210 

dpm/m3

, an intake rate of 10,500 dpm/yr is obtained for 250 workdays per year for the first year of the 

residual period (1984). This intake rate is reduced for subsequent years by the depletion factors listed 

in Table 4-2 of ORAUT (2012b). This provides an assessment of ingestion intake that is favorable to 

claimants.

7.4.2 Parks Township Site

The following information provides a method for estimating exposures during the residual radiation 

period due to plutonium and uranium contamination. Valid bioassay data for plutonium are unlikely to 

be available during the residual period. However, monitoring data can be used instead of the default 

intake assumptions given below to limit dose, as appropriate (NIOSH 2008b). For the first three years 

of the Parks Township residual period (1981 – 1983), bioassay monitoring data for uranium may be 

available. If the worker may have worked at the Apollo site during this period, the data may be related 

to DOE work and would be valid for use in assignment of internal dose. For workers who were known 

to have only worked at the Parks Township site during this period, the bioassay data should only be 

used to limit internal dose.

Guidance in ORAUT-OTIB-0070, Dose Reconstruction During Residual Radioactivity Periods at 

Atomic Weapons Employer Facilities (ORAUT 2012b) describes methods to estimate intake of 

radionuclides during residual periods. The method relates residual surface contamination to 

workplace air concentration during periods after completion of DOE work.

The majority of the DOE-related work at the Parks Township site involved plutonium fuel fabrication. 

While no formal air monitoring studies are available for these activities, an assessment of bounding 

general area air activity in the facilities can be made through a review of available air sampling data. 

Starting in October 1967, NUMEC reported personnel exposures above regulatory thresholds to the 

AEC (Caldwell 1967b). A bounding representation of air activity at the Parks Township site was 

determined based on a review of these reports. The data set includes 105 reported values between 

1966 and 1982. The reported values were given as MPC-hours (for plutonium exposure), which were 

converted to dpm/m3 by multiplying by the MPC and dividing by the number of hours given for the 

reported value. Values reported at the MPC were set to the MPC air concentration. The resultant 

median air concentration was estimated to be 11.6 dpm/m3 with a GSD of 4.97 (assuming a lognormal 

distribution). The air concentration values and the lognormal fit of the data are shown in Figure 7-1.

The median plutonium air concentration from the above-described study was used to estimate the 

residual surface concentration along with the guidance from ORAUT (2012b). The annual deposition 

amount was estimated using a deposition velocity of 0.00075 m/s, with deposition assumed to occur 

for 1 year. Using this approach, a surface concentration of plutonium was estimated as follows.

11.6 dpm/m3 × 31,536,000 s/yr × 0.00075 m/s = 2.74 × 105 dpm/m2

This surface concentration is favorable to claimants at the end of the operating period. The deposited 

material was assumed to be resuspended and inhaled during the residual period. The amount of 

resuspension was assumed to reduce with time due to fixing of the material on surfaces and to 

depletion (ORAUT 2012b). The depletion factors that were applied to each year are described in 

Table 4-2 of ORAUT (2012b). The depletion factors indicated for the residual concentration at the 

end of the operational period should be used for the first year; the remaining years should be reduced 

by factors listed in Table 4-2 of ORAUT (2012b).
Document No. ORAUT-TKBS-0041 Revision No. 02 Effective Date: 11/26/2012 Page 66 of 89

The air concentration for each year was estimated using a resuspension factor of 1 × 10-6/m (ORAUT 

2012b). Application of this resuspension factor and the above-described depletion factors to the 

residual contamination level of 2.74 × 105 dpm/m2 resulted in the air concentration and annual intakes

in Table 7-3. The intake evaluation was based on exposure for 2,000 hr/yr and an inhalation rate of 

1.2 m3

/hr.

Because the plutonium is based on gross alpha general air monitoring results, the activity represents 

the total alpha activity. The dose from the residual activity should be evaluated using the fractional 

isotopic activity of each radionuclide given in Table 5-3. The values in this table must be converted to 

fractional alpha activity for evaluation of dose. The age of the material must also be considered 

because the maximum dose (per unit alpha activity) is obtained for the shorter aged material. The 

beginning of plutonium processing at the NUMEC plutonium facility was in 1962. Therefore, the age 

of the residual material could have ranged from zero to 20 years or more if aged material was used in 

1962. The minimum age would be zero years if it was deposited at the end of the operating period in 

1980. A favorable to claimant approach would be to use the zero aged material activity fractions for 

the period from 1981 – 1984, the 5-year activity fractions from 1985 – 1989, the 10-year activity 

fractions from 1990 – 1994, the 15-year activity fractions from 1995 to 1999, and the 20-year activity 

fractions for 2000 – 2004. A minimizing approach would be to use the 20-year activity fractions for all 

years.

 Figure 7-1. Reported plutonium air concentrations.

The intakes in Table 7-3 can be used to estimate the internal dose to the target organ for the years of 

employment for the worker. The estimated internal doses are assigned as a lognormal distribution 

with a GSD of 4.97, which corresponds to the distribution of the average daily air concentrations that 

were used to estimate the annual plutonium intake. The dose should be evaluated for the two 

plutonium material solubility types of M and S, and the dose from the highest type should be used in 

0.1

1

10

100

1000

-3 -2 -1 0 1 2 3

Standard Normal Deviate, z

Airborne Pu Concentration (dpm/m3

)

Sorted Data

ln(Data) Pred. GM= 11.6 GSD= 4.97

x
Document No. ORAUT-TKBS-0041 Revision No. 02 Effective Date: 11/26/2012 Page 67 of 89

the IREP input. Because the residual plutonium is in the form of aged material, type Super S 

plutonium should be considered and adjustments made based on guidance in ORAUT (2010a).

The internal dose analysis should include the potential inadvertent ingestion of plutonium activity, 

based on guidance in NIOSH (2004). The daily intake rate (dpm/d) is estimated to be 0.2 times the 

average daily air concentration in units of dpm/m3

. Using the air concentration at the end of 

operations of 11.6 dpm/m3

, an intake rate of 580 dpm/yr is obtained for 239Pu for 250 workdays per 

year for the first year of the residual period (1981). This intake rate is reduced for subsequent years 

by the depletion factors listed in Table 4-2 of ORAUT (2012b). This provides an assessment of 

ingestion intake that is favorable to claimants.

 Table 7-3. Plutonium air concentration and annual intake in the Parks 

Township site residual period.

Year

Air concentration 

(dpm/m3)

Inhalation Intake 

(dpm/yr)a

Ingestion 

Intake 

(dpm/yr)a

239Pu 239Pu

1981 0.27 658 580

1982 0.21 516 450

1983 0.17 404 360

1984 0.13 316 280

1985 0.10 248 220

1986 0.081 194 170

1987 0.063 152 130

1988 0.050 119 100

1989 0.039 93 82

1990 0.030 73 64

1991 0.024 57 50

1992 0.019 45 39

1993 0.015 35 31

1994 0.011 27 24

1995 0.0089 21 19

1996 0.0070 16.8 15

1997 0.0055 13.2 12

1998 0.0043 10.3 9.0

1999 0.0034 8.1 7.1

2000 0.0026 6.3 5.6

2001 0.0021 4.9 4.4

2002 0.0016 3.9 3.4

2003 0.0013 3.0 2.7

2004 0.0010 2.4 2.1

a. Doses are assigned as a lognormal distribution with a GSD of 4.97.

Table 7-4. Activity fraction for plutonium material at the Parks Township plutonium facility.

Mixture 

designation Fraction of alpha activity

Years of aging 0 5 10 15 20

Pu-238 0.18 0.14 0.12 0.10 0.093

Pu-239/240 0.82 0.68 0.59 0.54 0.51

Pu-241 31.9 20.6 14.2 10.2 7.5

Am-241 0.000 0.19 0.29 0.35 0.40
Document No. ORAUT-TKBS-0041 Revision No. 02 Effective Date: 11/26/2012 Page 68 of 89

The intakes in Table 7-3 relate to DOE work conducted in Building A of the Parks Township site. 

Work with uranium in Building C could also contribute to internal dose for workers during the residual 

period. However, no information was found to allow determination of the residual uranium activity 

from work performed in Building C. For workers who are known to be exposed from Building C 

residual activity, the internal dose can be evaluated based on intakes given in Table 7-2, with intakes 

shifted back three years to account for the difference in residual period between the Apollo and Parks 

Township sites. These intakes are given in Table 7-5 for all years of the Parks Township residual 

period. If the work location is not known for a Parks Township worker, then the higher of the doses 

based on plutonium (Table 7-3) and uranium (Table 7-5) can be assigned as a favorable to claimant 

approach.

Table 7-5. Uranium air concentration and annual intake in 

the Parks Township residual period.

Year

Air concentration 

(dpm/m3)

Inhalation

Intake 

(dpm/yr)a

Ingestion 

Intake 

(dpm/yr)a

1981 4.97 1.19E+04 10500

1982 3.89 9.33E+03 8200

1983 3.04 7.31E+03 6400

1984 2.38 5.72E+03 5000

1985 1.87 4.48E+03 3900

1986 1.46 3.50E+03 3100

1987 1.15 2.75E+03 2400

1988 0.90 2.16E+03 1900

1989 0.70 1.68E+03 1500

1990 0.55 1.32E+03 1200

1991 0.43 1.03E+03 910

1992 0.34 8.09E+02 710

1993 0.26 6.34E+02 560

1994 0.21 4.96E+02 440

1995 0.16 3.89E+02 340

1996 0.13 3.04E+02 270

1997 0.10 2.38E+02 210

1998 0.077 1.86E+02 160

1999 0.061 1.47E+02 130

2000 0.048 1.14E+02 100

2001 0.037 8.95E+01 79

2002 0.029 7.01E+01 62

2003 0.023 5.50E+01 48

2004 0.018 4.30E+01 38

a. Doses are assigned as a lognormal distribution with a GSD of 7.91.

8.0 ATTRIBUTIONS AND ANNOTATIONS

All information requiring identification was addressed via references integrated into the reference 

section of this document.
Document No. ORAUT-TKBS-0041 Revision No. 02 Effective Date: 11/26/2012 Page 69 of 89

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Uranium, ANSI Standard N13.22-1995, American National Standards Institute, McLean, 

Virginia, October.

ICRP (International Commission on Radiological Protection), 1991, 1990 Recommendations of the 

International Commission on Radiological Protection, Publication 60, Pergamon Press, Oxford, 

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Publication 71, Pergamon Press, Oxford, England. [SRDB Ref ID: 22729]

Katine, E., 1960, “NUMEC Feasibility Report for General Atomic HTGR Critical Assembly Fuel 

Elements,” memorandum to H. V. Werner, U.S. Atomic Energy Commission, New York 

Operations Office, New York, New York, March 25. [SRDB Ref ID: 11937, p. 3]

Katine, E., 1961, “Visit to NUMEC on June 7, 1961,” memorandum to J. E. McLaughlin, U.S. Atomic 

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LaBone, T. R. 2010, Spreadsheet Created by Tom LaBone for NUMEC Pu Urine Measurement Errors 

from 10/1961 - 12/1965, Oak Ridge Associated Universities Team, Oak Ridge, Tennessee, 

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Leavitt, M. O., 2007, HHS Designation of Additional Members of the Special Exposure Cohort under 

the Energy Employees Occupational Illness Compensation Program Act of 2000, Designating 

a Class of Employees from Nuclear Materials and Equipment Corporation (NUMEC), Apollo, 

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Leavitt, M. O., 2008, HHS Designation of Additional Members of the Special Exposure Cohort under 

the Energy Employees Occupational Illness Compensation Program Act of 2000, Designating 

a Class of Employees from Nuclear Materials and Equipment Corporation (NUMEC), Parks 

Township, Pennsylvania, U.S. Department of Health and Human Services, Office of the 

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13.

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Report, Petition SEC-00080, Rev. 0, Office of Compensation Analysis and Support, Cincinnati, 

Ohio, August 31.

NIOSH (National Institute for Occupational Safety and Health), 2007b, External Dose Reconstruction 

Implementation Guideline, OCAS-IG-001, Rev. 3, Office of Compensation Analysis and 

Support, Cincinnati, Ohio, November 21.

NIOSH (National Institute for Occupational Safety and Health), 2008a, Petition Evaluation Report,

Petition SEC-00108, Rev. 0, Office of Compensation Analysis and Support, Cincinnati, Ohio, 

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NIOSH (National Institute for Occupational Safety and Health), 2008b, The Use of Data from Other 

Facilities in the Completion of Dose Reconstructions Under the Energy Employees 

Occupational Illness Compensation Program Act, OCAS-IG-004, Rev 0, Office of 

Compensation Analysis and Support, Cincinnati, Ohio, August.

NIOSH (National Institute for Occupational Safety and Health), 2010a, Radiation Exposures Covered 

for Dose Reconstructions Under Part B of the Energy Employees Occupational Illness 

Compensation Program Act, DCAS-IG-003, Rev. 2, Division of Compensation Analysis and 

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Corporation Procedures for Recovery of Uranium Scrap, Nuclear Materials and Equipment 

Corporation, Apollo, Pennsylvania, May. [SRDB Ref ID: 11959]

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Environmental Management Project (FEMP) – Occupational Internal Dose, ORAUT-TKBS0017-5, Rev. 00, Oak Ridge, Tennessee, May 28.

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Alpha-Neutron Reactions in Uranium and Thorium Compounds, ORAUT-OTIB-0024, Rev. 00, 

Oak Ridge, Tennessee, April 7.
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GLOSSARY

absorption

In external dosimetry, process in which radiation energy is imparted to material. In internal 

dosimetry, movement of material to blood regardless of mechanism.

absorption type

Categories for materials according to their rate of absorption from the respiratory tract to the 

blood, which replaced the earlier inhalation clearance classes. Defined by the International 

Commission on Radiological Protection, the absorption types are F: deposited materials that 

are readily absorbed into blood from the respiratory tract (fast solubilization), M: deposited 

materials that have intermediate rates of absorption into blood from the respiratory tract 

(moderate rate of solubilization), and S: deposited materials that are relatively in the 

respiratory tract (slow solubilization). Also called solubility type. See inhalation class.

accuracy

The characteristics of an analysis or determination that ensures that both the bias and 

precision of the resultant quantity will remain within the specified limits.

activity

Amount of radioactivity. The International System unit of activity is the becquerel 

(1 disintegration per second); the traditional unit is the curie [37 billion (3.7 × 1010) becquerels].

activity fraction

Proportion of the total activity due to a particular radionuclide.

acute exposure

Radiation exposure to the body delivered in a short period. See chronic exposure.

air sampling

Collection of samples of the ambient atmosphere to detect or measure the presence of 

radioactive material in the air.

albedo dosimeter

Thermoluminescent dosimeter that measures the thermal, intermediate, and fast neutrons 

scattered and moderated by the body or a phantom from an incident fast neutron flux.

alpha radiation

Positively charged particle emitted from the nuclei of some radioactive elements. An alpha 

particle consists of two neutrons and two protons (a helium nucleus) and has an electrostatic 

charge of +2.

ambient atmosphere

Depending on context, the air external to buildings, in the outside environment, or the air that 

surrounds an individual.

americium–beryllium (AmBe)

Common neutron source created by homogeneously mixing 241Am and beryllium powders. 

Neutrons are produced when 241Am alpha particles interact with beryllium nuclei.

atomic weapons employer (AWE) [42 U.S.C. § 7384l(5)]

Entity other than the United States, that—(A) processed or produced, for use by the United 

States, material that emitted radiation and was used in the production of an atomic weapon, 

excluding uranium mining and milling, and (B) is designated by the Secretary of Energy as an 
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atomic weapons employer for purposes of the [Energy Employees Occupational Illness] 

compensation program.

attenuation

Process by which absorption and scattering reduces the number of particles or photons 

passing through a body of matter.

beta dose

Designation (i.e., beta) on some records for external dose from beta and less-energetic X-ray 

and gamma radiation, often for shallow dose or dose to the lens of the eye.

beta radiation

Charged particle emitted from some radioactive elements with a mass equal to 1/1,837 that of 

a proton. A negatively charged beta particle is identical to an electron. A positively charged 

beta particle is a positron.

bioassay

Measurement of amount or concentration of radionuclide material in the body (in vivo 

measurement) or in biological material excreted or removed from the body (in vitro 

measurement) and analyzed for purposes of estimating the quantity of radioactive material in 

the body.

body burden

Amount of radioactive material in an individual's body at a particular point in time.

byproduct

Material left over from a nuclear, physical, or chemical process designed to produce a 

particular substance.

calcine

(1) Dry solid (grainy or granular) product of a chemical process that removes liquids from a 

solution. (2) Process for creating the chemical reaction that removes liquids from a solution.

calibration

Adjustment or determination of the response or reading of an instrument relative to a standard 

or a series of conventionally true values.

chronic exposure

Radiation dose to the body delivered in small amounts over a long period (e.g., days or years). 

See acute exposure.

cohort

Group of individuals selected for inclusion in a study. See Special Exposure Cohort.

confidence level

The interval about an estimate of a stated quantity within which the value of the quantity is 

expected to be with a specified probability. See uncertainty.

contamination

Radioactive material in an undesired location including air, soil, buildings, animals, and 

persons.

control rod

Neutron-absorbing device in a reactor used to slow or speed the reaction.
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decay

(1) Disintegration of atomic nuclei from spontaneous radioactivity including alpha, beta, and 

neutron radiation, often accompanied by gamma radiation. (2) Decrease in the amount of 

radioactive material over time due. See half-life.

decay products

See progeny.

decommissioning

Removal of a facility from service, usually involving decontamination of radioactivity to 

specified levels and often involving demolition of the facility.

decontamination

Reduction or removal of radioactive material from a structure, area, object, or person. 

Decontamination can occur through (1) treating the surface to remove or decrease the 

contamination or (2) allowing natural radioactive decay to occur over a period of time.

deep dose

See personal dose equivalent.

depleted uranium (DU)

Uranium with a percentage of 235U lower than the 0.7% found in natural uranium.

dose

In general, the specific amount of energy from ionizing radiation that is absorbed per unit of 

mass. Effective and equivalent doses are in units of rem or sievert; other types of dose are in 

units of roentgens, rads, reps, or grays.

dose equivalent (H, DE)

In units of rem or sievert, product of absorbed dose in tissue multiplied by a weighting factor 

and sometimes by other modifying factors to account for the potential for a biological effect 

from the absorbed dose. See dose.

dosimeter

Device that measures the quantity of received radiation, usually a holder with radiationabsorbing filters and radiation-sensitive inserts packaged to provide a record of absorbed dose 

received by an individual. See albedo dosimeter, film dosimeter, and thermoluminescent 

dosimeter, and track-etch dosimeter.

dosimetry

Measurement and calculation of internal and external radiation doses.

dosimetry system

System for assessment of received radiation dose. This includes the fabrication, assignment, 

and processing of external dosimeters, and/or the collection and analysis of bioassay samples, 

and the interpretation and documentation of the results.

electron

Basic atomic particle with negative charge and a mass 1/1,837 that of a proton. Electrons 

surround the positively charged nucleus of the atom. See element.
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enriched uranium (EU)

Uranium in which processing has increased the proportion of 235U to 238U to above the natural 

level of 0.7% by mass. Reactor-grade uranium is usually about 3.5% 235U; weapons-grade 

uranium contains greater than 90% 235U.

enrichment

Isotopic separation process that increases the percentage of a radionuclide in a given amount 

of material above natural levels. For uranium, enrichment increases the amount of 235U in 

relation to 238U. Along with the enriched uranium, this process results in uranium depleted in 

235U. See depleted uranium and enriched uranium.

environmental occupational dose

Dose received from radiation site-related activities (i.e., above normal background levels) 

while on a site, which is often recorded by monitoring stations in specific areas or along the 

boundaries of facilities (e.g., plant stack emissions).

error

Difference between the correct, true, or conventionally accepted value and the measured or 

estimated value. Sometimes used to mean estimated uncertainty. See accuracy and 

uncertainty.

exchange period (frequency)

Period (weekly, biweekly, monthly, quarterly, etc.) for routine exchange of dosimeters.

exposure

(1) In general, the act of being exposed to ionizing radiation. See acute exposure and chronic 

exposure. (2) Measure of the ionization produced by X- and gamma-ray photons in air in units 

of roentgens.

external dose

Dose received from radiation emitted by sources outside the body.

film dosimeter

Package of film for measurement of ionizing radiation exposure for personnel monitoring 

purposes. A film dosimeter can contain two or three films of different sensitivities, and it can 

contain one or more filters that shield parts of the film from certain types of radiation. When 

developed, the film has an image caused by radiation measurable with an optical 

densitometer. Also called film badge.

fission product (FP)

Radionuclides produced by fission or by the subsequent radioactive decay of radionuclides.

gamma radiation

Electromagnetic radiation (photons) of short wavelength and high energy (10 kiloelectron-volts 

to 9 megaelectron-volts) that originates in atomic nuclei and accompanies many nuclear 

reactions (e.g., fission, radioactive decay, and neutron capture). Gamma photons are identical 

to X-ray photons of high energy; the difference is that X-rays do not originate in the nucleus.

geometric standard deviation (GSD)

In probability theory and statistics, the geometric standard deviation describes the spread of a 

set of numbers whose preferred average is the geometric mean.
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glovebox

Enclosure with special rubber gloves through which an operator can handle radioactive or 

toxic material without risk of injury or contamination normally operated at a slightly reduced 

pressure so that air leakage, if any, is inward.

high-efficiency particulate air (HEPA) filter

Dense filter that removes contaminants from air flows before return to the working environment 

or discharge to the outside air (exhaust).

high-temperature gas-cooled reactor (HTGR)

Nuclear reactor cooled with helium.

highly enriched uranium (HEU)

Uranium enriched to at least 20% 235U for use as fissile material in nuclear weapons 

components and some reactor fuels.

hot cell

Shielded laboratory for handling of radioactive materials with the aid of remotely operated 

manipulators. The walls and windows are made of materials that protect workers from 

radiation.

ingestion

Process of taking a substance into the body through the mouth.

insoluble

Having very low solubility. No material is absolutely insoluble. See absorption type and 

soluble.

intake

Radioactive material taken into the body by inhalation, absorption through the skin, injection, 

ingestion, or through wounds.

Integrated Modules for Bioassay Analysis (IMBA)

Computer program that uses bioassay results and other information to calculate intakes of 

radionuclides and subsequent doses.

Interactive RadioEpidemiological Program (IREP)

Computer program that uses a person’s calculated annual organ doses and other information 

(e.g., gender, age at diagnosis, and age at exposure) to calculate the probability of causation 

of a specific cancer for a given pattern and level of radiation exposure.

internal dose

Dose received from radioactive material in the body.

internal dose assessment

Estimation of an intake of radioactive material and the consequent radiation dose based on 

bioassay or other measurements in the work environment.

in vitro bioassay

Measurements to determine the presence of or to estimate the amount of radioactive material 

in the excreta or in other biological materials removed from the body.
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in vivo bioassay

The measurements of radioactive material in the human body utilizing instrumentation that 

detects radiation emitted from the radioactive material in the body.

ionizing radiation

Radiation of high enough energy to remove an electron from a struck atom and leave behind a 

positively charged ion. High enough doses of ionizing radiation can cause cellular damage. 

Ionizing particles include alpha particles, beta particles, gamma rays, X-rays, neutrons, highspeed electrons, high-speed protons, photoelectrons, Compton electrons, positron/negatron 

pairs from photon radiation, and scattered nuclei from fast neutrons. See alpha radiation, beta 

radiation, gamma radiation, neutron radiation, photon radiation, and X-ray radiation.

limit of detection (LOD)

Minimum level at which a particular device can detect and quantify exposure or radiation. Also 

called lower limit of detection and detection limit or level. See minimum detectable level.

low-level radioactive waste (LLRW)

Unwanted radioactive materials that do not require shielding during normal handling or 

transport because of their low activity. Mildly radioactive material is usually disposed of by 

incineration and burial.

minimum detectable activity or amount (MDA)

Smallest amount (activity or mass) of an analyte in a sample that can be detected with a 

probability β of nondetection (Type II error) while accepting a probability α of erroneously 

deciding that a positive (nonzero) quantity of analyte is present in an appropriate blank sample 

(Type I error). See action level, decision level, and minimum reporting level.

minimum detectable concentration (MDC)

Minimum detectable activity (or amount) in units of concentration. See minimum detectable 

activity.

minimum detectable level (MDL)

See minimum detectable activity.

missed dose

(1) In relation to external dose, dose to monitored workers that was not measured or recorded 

due to such factors as a missing or damaged dosimeter or a result below the detection limits of 

the dosimeter. Missed dose is especially important in the early years of radiation monitoring, 

when relatively high detection limits were combined with short exchange periods. (2) In 

relation to internal dose, potential dose that could have been received by a bioassay program 

participant but, because of limitations in the monitoring system, was undetected.

mixed oxide (MOX)

Nuclear fuel that contains both plutonium oxide and uranium oxide.

monitoring

Periodic or continuous determination of the presence or amount of ionizing radiation or 

radioactive contamination in air, surface water, groundwater, soil, sediment, equipment 

surfaces, or personnel (for example, bioassay or alpha scans). In relation to personnel, 

monitoring includes internal and external dosimetry including interpretation of the 

measurements.
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natural uranium (NU)

Uranium as found in nature, approximately 99.27% 238U, 0.72% 235U, and 0.0054% 234U by 

mass. The specific activity of this mixture is 2.6 × 107 becquerel per kilogram (0.7 microcurie

per gram).

neutron film dosimeter

Film dosimeter with a nuclear track emulsion, type A, film packet.

neutron radiation

Radiation that consists of free neutrons unattached to other subatomic particles emitted from a 

decaying radionuclide. Neutron radiation can cause further fission in fissionable material such 

as the chain reactions in nuclear reactors, and nonradioactive nuclides can become 

radioactive by absorbing free neutrons.

nonpenetrating dose

Dose from beta and lower energy photon (X-ray and gamma) radiation that does not penetrate 

the skin. It is often determined from the open window dose minus the shielded window dose. 

See dose.

nuclear track emulsion, type A (NTA)

Film sensitive to fast neutrons made by the Eastman Kodak. The developed image has tracks 

caused by neutrons that become visible under oil immersion with about 1,000-power 

magnification.

nuclide

Stable or unstable isotope of any element. Nuclide relates to the atomic mass, which is the 

sum of the number of protons and neutrons in the nucleus of an atom. A radionuclide is an 

unstable nuclide.

occupational dose

Internal and external ionizing radiation dose from exposure during employment. Occupational 

dose does not include that from background radiation or medical diagnostics, research, or 

treatment, but does include dose from occupationally required radiographic examinations that 

were part of medical screening.

occupational environmental dose

Dose received while on the grounds of a site but not inside a building or other facility.

occupational medical dose

Dose from X-ray procedures performed for medical screening of workers as part of an 

occupational health program. Doses from X-rays used to diagnose diseases or injuries, even 

if incurred on the job, are not considered occupational and are therefore not eligible to be 

included in dose reconstruction under EEOICPA.

open window (OW)

Area of a film dosimeter that has little to no radiation shielding (e.g., only a holder and visible 

light protection). See film dosimeter.

penetrating dose

Dose from moderate to higher energy photons and neutrons that penetrates the outer layers of 

the skin. See dose.
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personal dose equivalent [Hp(d)]

Dose equivalent in units of rem or sievert in soft tissue below a specified point on the body at 

an appropriate depth d. The depths selected for personal dosimetry are 0.07 millimeter 

(7 milligrams per square centimeter) and 10 millimeters (1,000 milligrams per square 

centimeter), respectively, for the skin (shallow) and whole-body (deep) doses. The 

International Commission on Radiological Measurement and Units recommended Hp(d) in 

1993 as dose quantity for radiological protection.

photon radiation

Electromagnetic radiation that consists of quanta of energy (photons) from radiofrequency 

waves to gamma rays.

probability of causation (POC)

For purposes of dose reconstruction for the Energy Employees Occupational Illness 

Compensation Act, the percent likelihood, at the 99th percentile, that a worker incurred a 

particular cancer from occupational exposure to radiation.

progeny

Nuclides that result from decay of other nuclides. Also called decay products.

radiation

Subatomic particles and electromagnetic rays (photons) with kinetic energy that interact with 

matter through various mechanisms that involve energy transfer. See ionizing radiation.

radiation source

Any object or substance that emits radiation.

radiation worker

Employee who works on, with, or in the proximity of radiation-producing machines or 

radioactive materials.

radioactive

Of, caused by, or exhibiting radioactivity.

radioactive waste

Radioactive solid, liquid, and gaseous materials for which there is no further use. Wastes are 

generally classified as high-level (with radioactivity as high as hundreds of thousands of curies 

per gallon or cubic foot), low-level (in the range of 1 microcurie per gallon or cubic foot), 

intermediate level (between these extremes), mixed (also contains hazardous waste), and 

transuranic.

radionuclide

Radioactive nuclide. See radioactive and nuclide.

recycled uranium (RU)

Uranium first irradiated in a reactor then recovered through chemical separation and 

purification. RU contains minor amounts of transuranic material (e.g., plutonium and 

neptunium) and fission products (e.g., technetium) or uranium products (e.g., 236U) after 

purification.

resuspension

Process by which the wind lifts small particulates (generally from soil) into the air, which for 

radionuclides can result in an exposure hazard. Radionuclides released to the air undergo a 

cycle of suspension in the air, deposition on the ground, resuspension, and redeposition. 
Document No. ORAUT-TKBS-0041 Revision No. 02 Effective Date: 11/26/2012 Page 85 of 89

However, the initial suspension process is generally included in resuspension if the initial 

contaminating event did not result from deposition of airborne material.

shallow dose

See personal dose equivalent.

shielding

Material or obstruction that absorbs ionizing radiation and tends to protect personnel or 

materials from its effects.

solubility type

See absorption type.

soluble

In relation to health physics, refers to the speed with which radionuclides naturally dissolve in 

lung fluids. See absorption type and insoluble.

source term

Description of the types and quantities of radioactive materials. The source term is usually 

specified as a rate of exposure or an amount of radioactivity (i.e., becquerels or curies) 

sometimes by specific radionuclide. Often includes distinctions in chemical and physical forms 

and history of the material.

Special Exposure Cohort (SEC) [42 U.S.C. § 7384l(14)]

... “member of the Special Exposure Cohort" means a Department of Energy employee, 

Department of Energy contractor employee, or atomic weapons employee who meets any of 

the following requirements:

(C) (i) Subject to clause (ii), the employee is an individual designated as a member of the 

Special Exposure Cohort by the President for purposes of the compensation 

program under section 7384q of this title.

(ii) A designation under clause (i) shall, unless Congress otherwise provides, take 

effect on the date that is 180 days after the date on which the President submits to 

Congress a report identifying the individuals covered by the designation and 

describing the criteria used in designating those individuals.

Special Nuclear Material (SNM)

Plutonium or uranium enriched to a higher-than-natural assay including 239Pu, 233U, uranium 

containing more than the natural abundance of 235U, or any material artificially enriched in one 

of these isotopes.

standard deviation

Square root of the variance, or the measure of spread in a group of numbers.

thermoluminescence

Property that causes a material to emit light as a result of heat.

thermoluminescent dosimeter (TLD)

Device for measuring radiation dose that consists of a holder containing solid chips of material 

that, when heated, release the stored energy as light. The measurement of this light provides 

a measurement of absorbed dose.
Document No. ORAUT-TKBS-0041 Revision No. 02 Effective Date: 11/26/2012 Page 86 of 89

thermoluminescent dosimeter chip

Small block or crystal of lithium fluoride in a thermoluminescent dosimeter. A TLD-600 

dosimeter contains a chip made from more than 95% 6

Li for neutron radiation detection, and a 

TLD-700 dosimeter contains a chip made from more than 99.9% 7

Li for photon and beta 

radiation detection.

transuranic (TRU) elements

Elements with atomic numbers above 92 (uranium). Examples include plutonium and 

americium.

uncertainty

Standard deviation of the mean of a set of measurements. The standard error reduces to the 

standard deviation of the measurement when there is only one determination. See accuracy, 

confidence level, and error.

unmonitored dose

Potential unrecorded dose that could have resulted because a worker was not monitored. See 

missed dose.

whole-body counter

Equipment used to perform in vivo bioassay. Radiation emitted from radioactive material 

deposited throughout the body is measured.

whole-body (WB) dose

Dose to the entire body excluding the contents of the gastrointestinal tract, urinary bladder, 

and gall bladder and commonly defined as the absorbed dose at a tissue depth of 

10 millimeters (1,000 milligrams per square centimeter). Also called penetrating dose. See

dose.

X-ray

See X-ray radiation.

X-ray radiation

Electromagnetic radiation (photons) produced by bombardment of atoms by accelerated 

particles. X-rays are produced by various mechanisms including bremsstrahlung and electron 

shell transitions within atoms (characteristic X-rays). Once formed, there is no difference 

between X-rays and gamma rays, but gamma photons originate inside the nucleus of an atom.
Document No. ORAUT-TKBS-0041 Revision No. 02 Effective Date: 11/26/2012 Page 87 of 89

LIST OF TABLES

TABLE TITLE PAGE

A-1 HASL-82 weighted BZA survey, December 1959................................................................... 87

A-2 HASL-92 weighted BZA survey, June 1960 ........................................................................... 88

A-3 HASL-103 weighted BZA survey, October 1960..................................................................... 88

A-4 HASL-106 weighted BZA survey, December 1960................................................................. 89

A-5 HASL-114 weighted BZA survey, May 1961........................................................................... 89

Table A-1. HASL-82 weighted BZA survey, December 1959.

a

Operator

Number of 

persons

Average daily weighted 

exposures in dpm/m3

UF6 to UO3 Operator (Kiln and Filter) 4 790

Ceramics Reduction Furnace Operator 1 560

Ceramics Laboratory Operator 3 6,300

Sinter Furnace Operator 3 130

Ceramics laboratory–Group Leader 3 940

Co-precipitator Operator 2 950

U-Zr Recovery Operator–Chip 1 43

U-Zr Recovery Operator–Dissolving 1 49

U-Zr Recovery Operator–Reduction 1 39

Wet Chemistry Laboratory–Group Leader 1 37

Wet Chemistry Laboratory–Chemist 2 37

Wet Chemistry Laboratory–Technician 3 37

Gas Analysis–Chemist 1 7.5

Gas Analysis–Technician 1 7.5

Spectrographic Laboratory–Group Leader 1 17

Spectrographic Laboratory–Technician 3 17

Metallographic Group Leader 1 7

Grinding and Polishing Technician 1 7

Coated Sphere Operator 3 12

a. Data from AEC (1960a).

ATTACHMENT A

BREATHING-ZONE AIR SAMPLE SURVEY DATA

Page 1 of 3
Document No. ORAUT-TKBS-0041 Revision No. 02 Effective Date: 11/26/2012 Page 88 of 89

Table A-2. HASL-92 weighted BZA survey, June 1960.

a

 

Operator

Number of 

persons

Average daily weighted 

exposures in dpm/m3

UF6 to UO3 Operator (Kiln and Filter) 3 370

UF6 to UF3 Operator (Hydrolysis and Waste Filter) 6 73

Ceramic Reduction–Kiln Operator 3 96

Powder Preparation–Ceramics Fabrication 1 820

Ceramics Laboratory Operator 1 5,500

Sinter Furnace Operator 3 94

Centerless Grinder 2 100

Ceramic Fabrication–Prepress Operator (Graphite) 2 6,300

Ceramic Fabrication–Press Operator (Graphite) 2 490

Ceramic Fabrication–Pellet Press Operator (Uranium) 2 73

Ceramics Laboratory–Group Leader 3 600

Inspection – Graphite 6 57

Inspection – Uranium 2 57

U-Zr Recovery Operator–Extraction 2 160

U-Zr Recovery Operator–Powder 1 820

Wet Chemistry Laboratory (Group Leader, Chemists, 

Technicians) 10 9

Gas Analysis–(Chemist, Technician) 2 39

Spectrographic Laboratory–(Group Leader, Technician) 6 19

Metallographic (Group Leader and Technician) 3 10

Coated Sphere Operator 6 11

Maintenance 21 110

Sweepers 2 150

Sampler 3 110

a. Data from AEC (1960b).

Table A-3. HASL-103 weighted BZA survey, October 1960.

a

Operator

Number of 

persons

Average daily weighted 

exposures in dpm/m3

Ceramics Laboratory 3 36

Ceramic Laboratory–Group Leader 3 24

Ceramics Fabrication–Reduction Kiln Operator 3 65

Ceramics Fabrication–Pellet Press 1 19

Sinter Furnace Operator 3 16

Ceramics Reduction Tube Operator 1 590

Centerless Grinder 2 23

Coated Sphere Operator 6 31

CRP-3 Leach Operator 4 26

CRP-3 Filter and Feed Prep Operator 4 39

CRP-3 Extraction Operator 3 28

CRP-3 ADU Filter and Reduction Kiln Operator 3 27

a. Data from AEC (1960c).

ATTACHMENT A

BREATHING-ZONE AIR SAMPLE SURVEY DATA

Page 2 of 3
Document No. ORAUT-TKBS-0041 Revision No. 02 Effective Date: 11/26/2012 Page 89 of 89

Table A-4. HASL-106 weighted BZA survey, December 1960.

a

Operator

Number of 

persons

Average daily weighted 

exposures in dpm/m3 % U

Ceramics Laboratory Technician 2 680 93

Ceramics Laboratory Powder Production 6 910 93

Ceramics Laboratory Clerk 1 310 93

Ceramics Laboratory Group Leader 1 310 1.8-93

Ceramics Fabrication–Centerless Grinder 2 190 3.5

Ceramics Fabrication–Pellet Press 2 120 3.5

Ceramics Fabrication–Sinter Furnace 3 120 3.5-93

Ceramics Fabrication–Quality Control 4 61 3.5

CP-2 (UF6 to UO2) Filter and Drying 3 190 93

CP-2 (UF6 to UO2) Hexdrolysis 3 140 93

U-Zr powder 3 220 93

U-Zr Extraction Cascade 3 180 93

CRP-3 Extractor 3 110 1.8

CRP-3 Precipitate and Filter 1 82 1.8

CRP-3 Helper 1 88 1.8

Coated Sphere Operator 6 46 93

a. Data from AEC (1961a).

Table A-5. HASL-114 weighted BZA survey, May 1961.

a

Operator

Number of 

persons

Average daily weighted 

exposures in dpm/m3 % U

Ceramics Laboratory Reduction Tubes 1 130 5.7

Ceramics Fabrication–Pellet Press 2 35 5.7

Ceramics Fabrication–Centerless Grinder 2 200 5.7

Ceramics Fabrication–Sinter Furnace 3 27 3.4-5.7

Ceramics Fabrication–Quality Control 2 33 5.7

Ceramics Fabrication–Group Leader 2 31 5.7

CP-2 (UF6 to UO3) Hydrolysis 3 24 3.4

CP-2 Ammoniation and Filtering ADU 3 27 3.4

CP-2 Filtrate 3 20 3.4

CP-2 Calciner 3 57 3.4

CP-2 Reduction kiln Operator 3 31 3.4

CRP-2 (U-Zr) Dissolving and Filtering 3 39 5.7

CRP-2 Ammoniation, ADU Filter Conversion 3 27 5.7

CRP-2 Helper 3 33 5.7

CRP-2 Extraction Cascade 3 27 5.7

Coated Sphere Operator 6 22 Normal

a. Data from AEC (1961b).

ATTACHMENT A

BREATHING-ZONE AIR SAMPLE SURVEY DATA

Page 3 of 3