Page 1 of 111

DOE Review Release 03/22/2017

Site Profile for Nuclear Materials and 

Equipment Corporation, Apollo and Parks 

Township, Pennsylvania

ORAUT-TKBS-0041 Rev. 04 

Effective Date: 03/13/2017

Supersedes: Revision 03

Subject Expert(s): Dennis L. Strenge

Document Owner 

Approval: Signature on File

Dennis L. Strenge, Document Owner

Approval Date: 03/07/2017

Concurrence: Signature on File

John M. Byrne, Objective 1 Manager

Concurrence Date: 03/07/2017

Concurrence: Signature on File

Scott R. Siebert, Objective 3 Manager

Concurrence Date: 03/07/2017

Concurrence: Vickie S. Short Signature on File for

Kate Kimpan, Project Director

Concurrence Date: 03/08/2017

Approval: Signature on File

James W. Neton, Associate Director for Science

Approval Date: 03/13/2017

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

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

New Total Rewrite Revision Page Change
Document No. ORAUT-TKBS-0041 Revision No. 04 Effective Date: 03/13/2017 Page 2 of 111

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. 04 Effective Date: 03/13/2017 Page 3 of 111

EFFECTIVE 

DATE

REVISION 

NUMBER DESCRIPTION

08/25/2016 03 Revision initiated to respond to the S. Cohen & Associates review. 

Updated Table 2-3 to revise operating periods for Parks Township 

activities. Revised Section 5.1 and Table 5-3 to provide additional 

information on plutonium mixtures and evaluation of internal dose 

from plutonium mixtures and associated radionuclides. Added 

Section 5.2.2.4 to provide guidance on evaluation of intakes based 

on uranium fluorimetric bioassays. Updated Table 5-8 to include a 

few more MDA values for lung counts of 239Pu and 241Am. Guidance 

was added to use the maximum MDA values for years when the 

bioassay records do not provide an MDA value. Revised Section 

5.2.4 to eliminate use of ORAUT-OTIB-0054 for evaluation of intake 

of associated fission and activation products. Only the doses from 

listed radionuclides are to be evaluated. Revised Section 5.2.5 to 

indicate recycled uranium contaminants should be included in the 

evaluation of the material solubility type that gives the highest 

internal dose. Added Section 6.5.2.4 and Tables 6-8 and 6-9 to 

provide guidance on estimation of neutron dose before 1969 using 

neutron-to-photon ratios. Revised Table 7-1 to include whole-body 

dose and skin dose with guidance to apply the isotropic exposure 

dose conversion factors. This is consistent with the evaluations 

performed for chronic lymphocytic leukemia. Included exposure to 

thorium in the table. Added Table 7-2 to provide guidance on 

assignment of onsite external dose for organs affected by shallow 

dose. Revised Table 7-3 to use general air sampling data rather 

than breathing zone data as the basis for intakes of residual uranium. 

Added Table 7-4 for residual thorium intakes on the same basis. 

Revised the air concentration for residual air concentrations at Apollo 

and Parks Township uranium facilities to be based on a 

resuspension factor of 1x10-5 (rather than 1 × 10-6) and updated 

Tables 7-3, 7-4, and 7-7. Added Attachment B to provide guidance 

on evaluation of the various dosimetry reports available in the 

dosimetry records for NUMEC workers. 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.

03/13/2017 04 Revision initiated to add Section 5.1.3 for evaluation of internal dose 

from 238Pu when bioassay data are available based on DCAS-RPT0005, Alternative Dissolution Models for Insoluble Pu-238. Updated 

guidance in Section 5.2.5 for evaluation of recycle uranium 

contaminants referencing current revision of the Feed Materials 

Production Center – Occupational Internal Dose, Rev 02. 

Incorporates formal ORAUT and NIOSH review comments. Table 6-

2 and Sections 6.3.1 and 6.3.2 were updated in response to SC&A 

Finding 12 regarding dosimetry inconsistencies. Training required: 

As determined by Objective Manager. Initiated by Dennis L. Strenge.
Document No. ORAUT-TKBS-0041 Revision No. 04 Effective Date: 03/13/2017 Page 4 of 111

TABLE OF CONTENTS

SECTION TITLE PAGE

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

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

1.1 Purpose...................................................................................................................... 12

1.2 Scope......................................................................................................................... 12

1.3 Special Exposure Cohort............................................................................................ 12

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

1.3.2 Parks Township Site Class........................................................................... 13

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 ................................................................................... 16

2.1.3.1 General Process Operations ..................................................... 16

2.1.3.2 Uranium Process Activities........................................................ 16

2.1.3.3 High-Temperature Gas-Cooled Reactor Critical Assembly 

Fuel Elements ........................................................................... 18

2.1.3.4 Uranium Nitrate Solution for the University of California ............ 18

2.1.3.5 Incinerator ................................................................................. 18

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

2.1.3.7 Research Activities in the Early Years ....................................... 19

2.1.4 Source Term ................................................................................................ 19

2.1.5 Remediation, Decontamination, and Decommissioning................................ 20

2.1.5.1 Shallow Land Disposal Area...................................................... 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................................ 33

3.0 Occupational Medical Dose.................................................................................................... 34

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

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

4.2 Environmental External Dose ..................................................................................... 35

5.0 Occupational Internal Dose.................................................................................................... 35

5.1 Internal Exposure Sources ......................................................................................... 35

5.1.1 Uranium Source Information ........................................................................ 35

5.1.2 Plutonium Fuel Source Information .............................................................. 35

5.1.3 Plutonium-238 Source Fabrication ............................................................... 38
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5.2 In Vitro Bioassay......................................................................................................... 38

5.2.1 Plutonium Urine and Feces Bioassay........................................................... 38

5.2.1.1 Plutonium Urine Minimum Detectable Concentrations and 

Frequencies .............................................................................. 39

5.2.1.2 Plutonium Urine Analytical Procedure ....................................... 39

5.2.1.3 Plutonium Fecal Minimum Detectable Concentrations and 

Frequencies .............................................................................. 40

5.2.2 Uranium Urine and Feces Bioassay ............................................................. 42

5.2.2.1 Uranium Urine Analytical Procedure.......................................... 42

5.2.2.2 Urine MDCs and Frequencies ................................................... 43

5.2.2.3 Uranium Fecal Minimum Detectable Concentrations and 

Frequencies .............................................................................. 45

5.2.2.4 Uranium Fluorimetric Bioassay Evaluations............................... 45

5.2.3 Thorium Exposures...................................................................................... 46

5.2.4 Mixed Fission Products ................................................................................ 46

5.2.5 Unmonitored Radionuclides from Recycled Uranium ................................... 46

5.3 In Vivo Counting ......................................................................................................... 47

5.4 Reports of Overexposures and Incidents.................................................................... 48

5.4.1 Apollo Site Incidents..................................................................................... 49

5.4.2 Parks Township Site Incidents ..................................................................... 49

5.5 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.......................................................................................... 52

6.2.3 Neutron Exposures ...................................................................................... 52

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

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

6.3.2 Neutron Dosimeters ..................................................................................... 54

6.3.3 Limits of Detection........................................................................................ 54

6.3.4 Radiological Records ................................................................................... 55

6.4 Limitations in Measured Dose..................................................................................... 56

6.4.1 Low-Energy Photons.................................................................................... 56

6.4.2 Neutron Radiation ........................................................................................ 59

6.5 Dose Reconstruction Recommendations .................................................................... 59

6.5.1 Recorded Dose Practices............................................................................. 59

6.5.2 Adjustments to Recorded Dose.................................................................... 60

6.5.2.1 Beta Dose Adjustments............................................................. 60

6.5.2.2 Photon Dose Adjustments ......................................................... 60

6.5.2.3 Neutron Weighting Factor Adjustments ..................................... 60

6.5.2.4 Neutron Dose During NTA Film Period...................................... 60

6.5.3 Missed and Unmonitored Dose .................................................................... 61

6.5.3.1 Shallow Dose and Deep Dose................................................... 61

6.5.3.2 Neutron Dose............................................................................ 61

6.5.4 Uncertainty................................................................................................... 61

6.5.5 Radiation Dose Fraction............................................................................... 62

7.0 Estimation of Exposure to Residual Activity ........................................................................... 62

7.1 External Dose from Residual Activity in the Workplace............................................... 63

7.1.1 Apollo Site.................................................................................................... 63

7.1.2 Parks Township Site .................................................................................... 64
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7.2 External Ambient Dose from Residual Activity ............................................................ 65

7.3 Occupational Medical Dose ........................................................................................ 65

7.4 Internal Dose from Residual Activity in the Workplace................................................ 65

7.4.1 Apollo Site.................................................................................................... 65

7.4.2 Parks Township Site .................................................................................... 68

7.5 Remediation Worker Dose during Residual periods ................................................... 72

8.0 Attributions and Annotations .................................................................................................. 73

References ........................................................................................................................................ 74

Glossary ............................................................................................................................................ 83

ATTACHMENT A DOSIMETRY REPORTS ................................................................................ 94

LIST OF TABLES

TABLE TITLE PAGE

2-1 Apollo site area descriptions .................................................................................................. 15

2-2 Apollo site source and SNM possession limits ....................................................................... 20

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

2-4 Parks Township Building A source and SNM possession limits.............................................. 28

2-5 Parks Township Building B source and SNM possession limits.............................................. 29

2-6 Parks Township Building C source and SNM possession limits.............................................. 30

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 .................. 36

5-2 Uranium source term information ........................................................................................... 36

5-3 Activity composition of plutonium mixtures ............................................................................. 37

5-4 Plutonium and americium urine bioassay MDCs and frequencies by period........................... 40

5-5 Plutonium fecal bioassay MDCs and frequencies by period ................................................... 41

5-6 Uranium urine bioassay MDCs and frequencies by method and period.................................. 44

5-7 Uranium fecal bioassay MDCs and frequencies by period...................................................... 45

5-8 Lung-counting MDAs of uranium based on enrichment .......................................................... 48

5-9 In vivo MDAs for 239Pu and 241Am........................................................................................... 48

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

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

6-3 WB MDLs and potential missed photon, beta, or neutron dose .............................................. 55

6-4 Annual occupational radiation exposures at the Apollo site.................................................... 55

6-5 Plutonium photon (and beta) energy factors........................................................................... 58

6-6 Recorded dose practices........................................................................................................ 59

6-7 Interpretation of reported data................................................................................................ 59

6-8 Beta, photon, and neutron radiation energies and percentages for IREP input....................... 62

7-1 External annual dose from residual surface contamination at the Apollo site ......................... 63

7-2 Assignment of external onsite residual dose .......................................................................... 64

7-3 Uranium air concentration and annual intake in the Apollo site residual period ...................... 66

7-4 232Th air concentration and annual intake in the Apollo site residual period............................ 68

7-5 Plutonium air concentration and annual alpha activity intake in the Parks Township site 

residual period .......................................................................................................................70

7-6 Alpha activity fraction for plutonium material at the Parks Township plutonium facility ........... 71

7-7 Uranium air concentration and annual intake in the Parks Township residual period.............. 72

7-8 Decontamination and Decommissioning during Residual Periods .......................................... 72
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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....................................................................... 42

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 ............. 57

6-3 Dosimeter filter ratios .............................................................................................................57

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

fabrication from 1,000 MWd/t plutonium................................................................................. 58

7-1 Reported plutonium air concentrations................................................................................... 69

A-1 Landauer annual report for one worker (showing gamma, 1964)............................................ 96

A-2 Landauer monthly dose results (code 2314)........................................................................... 97

A-3 Landauer semi-monthly dose results (code 2315).................................................................. 98

A-4 Eberline report for one exchange period ................................................................................ 99

A-5 Monthly calculation for June 1972........................................................................................ 100

A-6 Special dosimetry calculation sheet ..................................................................................... 101

A-7 Monthly summary of gamma and neutron ............................................................................ 102

A-8 Monthly monitoring results by wear period ........................................................................... 103

A-9 Annual report with monthly and quarterly summaries ........................................................... 104

A-10 Annual report with quarterly summaries ............................................................................... 105

A-11 Annual report by month with gamma and neutron doses...................................................... 106

A-12 Annual report with monthly results later years...................................................................... 107

A-13 Worker lifetime summary...................................................................................................... 108

A-14 Cumulative dose through November 1971 ........................................................................... 109

A-15 Plutonium guard summary, 1971 to 1977............................................................................. 110
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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

DOL U.S. Department of Labor

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

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

hr hour

HTGR high-temperature gas-cooled reactor

ICRP International Commission on Radiological Protection

in. inch

IREP Interactive RadioEpidemiological Program

keV kiloelectron-volt, 1,000 electron-volts

kg kilogram
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L liter

lb pound

LEU low-enriched uranium

LLRW low-level radioactive waste

LOD limit of detection

m meter

M moderate (absorption type)

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

NS&E Nuclear Science & Engineering

NTA nuclear track emulsion, type A

NU natural uranium

NUMEC Nuclear Materials and Equipment Corporation

ORAU Oak Ridge Associated Universities

OW open window

pCi picocurie

PL plastic

PNC Power Reactor and Nuclear Fuel Development Corporation

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

SS super slow (absorption type); refers to highly insoluble plutonium

t ton (metric)

TLD thermoluminescent dosimeter
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TRU transuranic

U.S.C. United States Code

WB whole body

wt % weight %

yr year

ZPPR Zero Power Plutonium (later Physics) Reactor

ZPR-III Zero Power Reactor

μCi microcurie

μg microgram

§ section or sections
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1.0 INTRODUCTION

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 Department of Energy (DOE) or Atomic Weapons Employer (AWE) 

facilities or categories of DOE or AWE facilities. They will be revised in the event additional relevant 

information is obtained about the affected DOE or AWE facility(ies). These documents may be used 

to assist NIOSH staff in the evaluation of Special Exposure Cohort (SEC) petitions and the completion 

of the individual work required for each dose reconstruction.

In this document the word “facility” is used to refer to an area, building, or group of buildings that 

served a specific purpose at a DOE or AWE facility. It does not mean, nor should it be equated to, an 

“AWE facility” or a “DOE facility.” The terms AWE and DOE facility are defined in sections 7384l(5) 

and (12) of the Energy Employees Occupational Illness Compensation Program Act of 2000 

(EEOICPA), respectively. A DOE facility is defined as “any building, structure, or premise, including 

the grounds upon which such building, structure, or premise is located ... in which operations are, or 

have been, conducted by, or on behalf of, the [DOE] (except for buildings, structures, premises, 

grounds, or operations ... pertaining to the Naval Nuclear Propulsion Program);” and with regard to 

which the DOE has or had a proprietary interest, or “entered into a contract with an entity to provide 

management and operation, management and integration, environmental remediation services, 

construction, or maintenance services.” 42 U.S.C. § 7384l(12). On the other hand, an AWE facility 

means “a facility, owned by an atomic weapons employer, that is or was used to process or produce, 

for use by the United States, material that emitted radiation and was used in the production of an 

atomic weapon, excluding uranium mining or milling.” 42 U.S.C. § 7384l(5). The Department of 

Energy (DOE) determines whether a site meets the statutory definition of an AWE facility and the 

Department of Labor (DOL) determines if a site is a DOE facility and, if it is, designates it as such.

Accordingly, a Part B claim for benefits must be based on an energy employee’s eligible employment 

and occupational radiation exposure at a DOE or AWE facility during the facility’s designated time 

period and location (i.e., covered employee). After DOL determines that a claim meets the eligibility 

requirements under EEOICPA, DOL transmits the claim to NIOSH for a dose reconstruction. 

EEOICPA provides, among other things, guidance on eligible employment and types of radiation 

exposure to be included in an individual dose reconstruction. Under EEOICPA, eligible employment 

at an AWE facility is categorized as employment either (1) during “a period when the employer was 

processing producing, for the use by the United States, material that emitted radiation and was used 

in the production of an atomic weapon, excluding uranium mining and milling,” (i.e., the operational 

period); or (2) during a period that NIOSH has determined that “there is a potential for significant 

residual contamination outside of the period in which weapons-related production occurred,” (i.e., the 

residual contamination period). 42 U.S.C. § 7384l(3).

Based on the abovementioned definition for eligible employment during an AWE facility’s operational 

period, NIOSH includes radiation exposures incurred in the performance of duty, such as medical 

X-rays received as a condition of employment for participating in DOE projects, at an AWE facility in 

dose reconstructions. This may include 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 operational period. In contrast, 

only two categories of radiation exposure as defined in 42 U.S.C. § 7384n(c)(4) should be included in 

dose reconstructions for claims involving employment during the residual contamination period. First, 

NIOSH must include exposures to radiological contaminants resulting from activities that had a 

nuclear-weapon nexus or conducted by or on behalf of the DOE (with an exclusion of activities related 

to, among other things, the Naval Nuclear Propulsion Program) that took place during the operational 

period. 42 U.S.C. § 7384n(c)(4)(A). Second, radiation doses from sources not included in the first 
Document No. ORAUT-TKBS-0041 Revision No. 04 Effective Date: 03/13/2017 Page 12 of 111

category but which cannot be distinguished through reliable documentation should also be included in 

dose reconstructions. 42 U.S.C. § 7384n(c)(4)(B). Furthermore, because all DOE-related activities 

have ceased during the residual contamination period, NIOSH does not include doses from medical 

X-rays performed during the residual contamination period (NIOSH 2007) in dose reconstructions.

Likewise, NIOSH does not consider the following types of exposure as those incurred in the 

performance of duty as a condition of employment for DOE-related activities at an AWE facility. 

Therefore these exposures are not included in dose reconstructions for either the operational or 

residual contamination period (NIOSH 2010):

• 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

1.1 PURPOSE

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. 

1.2 SCOPE

Attributions and annotations, indicated by bracketed callouts and used to identify the source, 

justification, or clarification of the associated information, are presented in Section 8.0. 

1.3 SPECIAL EXPOSURE COHORT

This section describes the classes in the 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;
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• 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 contractor, 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;

• 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 contractor, 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 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 
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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 production of nuclear fuel from HEU (>20% 235U) 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. 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 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 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. 

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 
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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. 

Table 2-1. Apollo site area descriptions.

a

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

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 Not reported Not reported 1957–1983

QC Quality control Not reported 1957–1983

PS Not reported Not reported 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

Not reported 1957–1983

E Vault Storage of SNM Not reported 1957–1983

F Vault Storage for SNM of all 

enrichments

Not reported 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 Not reported 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

Laundry 

Building

Cleaning protective 

clothing

Co-60, plutonium, thorium, and 

uranium

1960–1983 

(closed 1984)

Block Building Storage Storage of processing equipment 1957–1983

Incinerator Area 62 (Hoskinson 

H-100) 

30–35 lb/hr, 300-g U-235 limit/batch 1960–1983

a. Sources: Author unknown (ca. 2004); NUMEC (1963); B&WNES (1997); Caldwell (1966, 1968a).
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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, 1960b, 1960c, 1961a, 1961b). 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. 

2.1.3.1 General Process Operations

Apollo process operations were varied. Because the Apollo site was a research center and an allpurpose process center, a variety of radioactive materials and special nuclear material (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 was 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 

reports AEC 1960a, 1960b, 1960c, 1961a, 1961b), which cover the period from December 1959 to 

January 1961, and Procedures for Recovery of Uranium Scrap from 1962 (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 to UO3 at a controlled temperature. The UO3 product was transferred in 

small polyethylene containers to the Ceramics Fabrication Area for further processing (AEC 1960a, 

p. 3). The HASL-92 survey from August 1960 noted that the filter cake was dried by a rotary kiln 

rather than the earlier fry pan method (AEC 1960b, p. 2). The HASL-114 survey from June 1961 

noted that a calciner was added for reduction of ADU to U3O8 (AEC 1961b, p. 2). 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 
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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. The 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.

Research and development (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, 

1960b, 1960c, 1961a, 1961b), 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:

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 NUMEC (1962), scrap was dissolved in two designated areas, CRP-2 and CRP-3. The 

product solutions from the dissolution methods were processed to generate insoluble UF4, and 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. 
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2.1.3.3 High-Temperature Gas-Cooled Reactor Critical Assembly Fuel Elements

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

fabricate 3,000 graphite fuel elements to be used in the General Atomics high-temperature gas-cooled 

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. An April 1960 letter from AEC approved the 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 June 9, 1961, letter report describes a trip to the Apollo site on June 7, 1961, to observe equipment 

for producing uranyl nitrate solution for the University of California (Katine 1961). A vague description 

of NUMEC processes and facilities was reported. The report mentions that nitric acid and aluminum 

nitrate solutions were used 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 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 

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 being 

discharged 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 the 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 ft/min. Material would be handled wearing 

latex gloves. 
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3. The powder would be slugged to 4 to 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, an automatic press, or both in a hood 

with a face velocity of 100 ft/min. 

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 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 at the Apollo site: 

(1) federal and State of Pennsylvania 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 HASL reports and process feasibility reports, which contain information on radionuclides, quantities, 

and recommended safety precautions for the described activity; and (3) inventory and 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 for the Apollo site are listed in Table 2-2.

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, 1960b, 1960c, 1961a, 1961b). The 

typical amounts of uranium in use in an area ranged from milligrams to hundreds of kilograms. 
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Table 2-2. Apollo site source and SNM possession limits (Reitler 1972). 

Areas Source/chemical or physical form

Maximum 

possession

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

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

Processing areas, laboratories, and vaults Plutonium as fully clad or encapsulated material 500 kg

Mass Spectrometry Laboratory Uranium in any enrichment 350 g

Mass Spectrometry Laboratory Plutonium in any form 0.5 g

LLRW storage areas Within fenced areas in approved storage 

containers

35 g U-235

LLRW storage areas In buildings meeting safeguards and security 

requirements

50 kg U-235

Nuclear Decontamination Corporation Any byproduct material 20 mCi

Nuclear Decontamination Corporation Any source material 20 g

Nuclear Decontamination Corporation Any SNM 20 mCi

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, 233U, 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 the site handled 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 (LLRW) disposal 

facility in Barnwell, South Carolina (B&WNES 1997). 

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 onsite 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 still 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). 
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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). 

The site became a Formerly Utilized Sites Remedial Action Program site with management of the 

program transferred to the U.S. Army Corps of Engineers in 1998. 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. 

Figure 2-2. B&W shallow land disposal area (USACE 2006).

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 in 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 (Fabs) 2 through 9 and the Drum 

Warehouse as well as termination and remediation of a drum storage area. The main structures were 
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Buildings A, B, and C, the outside Storage Areas, and the Trailer Storage Area (Author unknown 

ca. 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 

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 Zero Power Reactor (ZPR)-III fuel wafers (Author unknown ca. 2004). The 

ZPR-III fuel program used 8.5% 240Pu plutonium, and the ZPPR fuel program used primarily 12% 

240Pu (11,500 fuel plates), but also some 27% 240Pu (700 fuel plates), which indicates a 

commercial fuel source having higher burn-up (B&WNES 1996). The highly enriched 240Pu was 

not received on site until 1969. 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 ca. 2004). Production and process areas are 

summarized in Table 2-3 and discussed further in Section 2.2.2. 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. Scrap material was still 

present on site after completion of a fuel manufacturing campaign. 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.

a

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

1961–1980

A–Fab 1 Plutonium conversion Plutonium nitrate, plutonium 

oxide, depleted UO2

1961–1980

A–Fab 1 Fuel fabrication for FFTF Plutonium nitrate, plutonium 

oxide, depleted UO2

1972–1980

A–East Side of Fab 1 Routine repair and 

maintenance of contaminated 

equipment

All 1961–1980

A–Fab 2 Fuel fabrication for ZPR-III Plutonium nitrate, plutonium 

oxide, depleted UO2

1964–1966

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

oxide, depleted UO2

1966-1970

A–Fab 3 Manufacturing operations, 

metallography

Plutonium nitrate, plutonium 

oxide

1963–1980

A–Fab 3 Quality control of FFTF fuel Plutonium nitrate, plutonium 

oxide

1972–1980
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Building or area 

location Operations Radionuclides

Period of 

operation

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, Pu-238/239, PuO2, 

plutonium and americium metal

1963–1980

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

August 1, 1967 

A–Fab 5 Analytical laboratory work All, small quantities of radioactive 

samples

1979–1980

A–Fab 6 Scrap recovery Plutonium – various forms 1968–1980

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

1961–1969

Building A–

Hot Cell Room

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

Be-7, Po-210, Co-60, Am-241, 

Pu-238/239, PuO2,

1969–1980

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, Pu-238, 

ThO2

1961–1980

Building B–

Hafnium Plant

Metal production Hafnium and Zr-Be alloy 

(nonradioactive)

1961–1980

Building B–

Plutonium Annex

Conversion of Pu-238 nitrate 

to oxide

Pu-238 nitrate, Pu-238 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

DU, UO2, U3O8, UF4, fully clad U233, U-235, and Pu-2391963–1974

Building B–

Metals Plant, 

second floor

Pu-238 pacemakers Pu-238-powered heart 

pacemakers

1963–1970

Building B–Machine 

Shop

Occasional machining of clad 

or unclad uranium, and clad 

plutonium and U-233; 

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 and oxide complexes, 

SNM oxides (UO3, UO2 and 

U3O8)

1973–1978

Outdoor scrap 

storage area

Storage UF6 cylinders 1971–1980

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

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

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

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. 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 ca. 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 hours a 

day. 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.
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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 Fabs 1 through 7 and Fab 9, and in several small 

laboratories adjacent to the Fabs. 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 for portions of Cores 1 and 2 and the entire Cores 3 and 4. The FFTF fuel was a mixed oxide 

(MOX) 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 using 

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 HEPA-filtered 

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 one fume hood for mixing chemicals before inserting them 

into the cell and a second 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 cells were 
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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 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 during building 

decontamination and 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.
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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 using 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. DU machining 

was performed in the Machine Shop. In addition, equipment from the Apollo and Parks Township 

sites was repaired and refurbished. 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, but over the years other 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 such as the 

development of 238Pu-fueled heart pacemakers were performed under an AEC sealed-source 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.

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

placed in a furnace. The material was then 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. 

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

Plutonium and EU 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

Plutonium (in nonpyrophoric form, 

containing at least 3 wt % Pu-240)

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 U-235

Use: 1 kg effective

Rouse 1979

EU above 5 wt % U-235 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
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Source/chemical or physical form Period Maximum possession Reference

Pu-239 as encapsulated calibration or 

reference sources

1984–1991 50 g Austin 1984

Pu-239 as encapsulated calibration or 

reference sources

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

U-233 as evaporated calibration or 

reference sources

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

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

EU of 5 wt % U-235 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

EU above 5 wt % U-235 1981–1991 Possession: <5 kg

Use: 1 kg effective

Austin 1981

EU above 5 wt % U-235 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
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Source/chemical or physical form Period Maximum possession Reference

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

Byproduct material encapsulated 1984–1991 5 Ci per source of each 

isotope

Austin 1984; 

Haughney 1991

Byproduct material encapsulated 1991–end 10 Ci per source of each 

isotope

Haughney 1991

Byproduct material any form, 

contaminated waste on or in 

equipment 

1991–end 5 Ci Haughney 1991

Byproduct material any form, 

contaminated waste on or 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

Byproduct material encapsulated 1991–end 20 g Haughney 1991

Pu-239 as encapsulated calibration or 

reference sources

1981–1991 50 g Austin 1981

Pu-239 as encapsulated calibration or 

reference sources

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

U-233 as evaporated calibration or 

reference sources

1991–end 2 g Haughney 1991

U-235 as evaporated calibration or 

reference sources

1991–end 5 g Haughney 1991

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 None

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

EU above 5 wt % U-235 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

U-235 as encapsulated or 

electroplated calibration or reference 

sources

1981–1991 Possession: <5 kg

Use: 1 kg effective

Austin 1981
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Source/chemical or physical form Period Maximum possession Reference

U-235 as encapsulated or 

electroplated calibration or reference 

sources

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

Pu-239 as encapsulated calibration or 

reference sources

1981–1991 5 g Austin 1981

Pu-239 as encapsulated calibration or 

reference sources

1991–end 285 g Haughney 1991

Pu-239 as electroplated calibration or 

reference sources

1981–end 5 g Austin 1981

Pu-239 as electroplated calibration or 

reference sources

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

Plutonium plant 

outdoor storage area

EU up to 5 wt % U-235 in 

UF6 cylinders

1973–1984 200,000 kg UF6 Browne 1978

Plutonium plant 

outdoor storage area

EU up to 5 wt % U-235 in 

UF6 cylinders

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

Plutonium plant 

storage trailer

Plutonium in nonpyrophoric 

form with at least 3 wt % 

Pu-241

1991–end <200 g Haughney 1991

Plutonium plant 

storage trailer

Plutonium and U-235 in 

approved shipping 

containers with valid 

certificates of compliance

1984–1991 Any quantity Haughney 1991

Plutonium plant 

storage trailer

EU up to 5 wt % U-235, any 

physical or chemical form 

covered by authorized 

activities

1979–1991 25,000 kg U Rouse 1979

Plutonium plant 

storage trailer

NU or DU, any physical or 

chemical form covered by 

authorized activities

1979–end 100,000 kg U Rouse 1979

Plutonium plant 

storage trailer

Uranium, any enrichment 

U-235

1991–end Possession: 

<250 g U

Haughney 1991

Plutonium plant 

storage trailer

Byproduct material any 

form, contaminated waste

1991–end 1,000 Ci of any 

isotope

Haughney 1991

Plutonium plant 

storage trailer

Byproduct material, any 

form, contaminated waste 

on or in equipment

1991–end 5 Ci Haughney 1991
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Place of use

Source/chemical or 

physical form Period

Maximum 

possession Reference

Plutonium plant 

storage trailer

Byproduct material, any 

form, contaminated waste 

on or in equipment, and 

metallic materials from 

other licensees

1991–end 5,000 Ci Haughney 1991

Plutonium plant 

storage trailer

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 U2351991–end Possession: 

<350 g U

Haughney 1991

Outside storage 

areas

Byproduct material, any 

form, contaminated waste

1991–end 1,000 Ci of any 

isotope

Haughney 1991

Outside storage 

areas

Byproduct material, any 

form, contaminated waste 

on or 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

Any state except 

Agreement States

Byproduct material, any 

form, contaminated waste 

on or in equipment

1991–end 5 Ci Haughney 1991

Any state except 

Agreement States

Byproduct material, any 

form, contaminated waste 

on or in equipment, and 

metallic materials from 

other licensees

1991–end 5,000 Ci Haughney 1991

The use of SNM was governed by AEC regulations and 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 
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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 fuel source term. 

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 for portions of Cores 1 and 2 

and all of Cores 3 and 4. FFTF fuel was a MOX 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 % 241

Pu 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, 137

Cs, 

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 sitewide 

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 

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 use of Building A 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 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 
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radiation survey of the Metals Facility was performed in 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, and the equipment was packaged 

for burial. After the equipment was removed, B&W initiated a clean-up 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 in 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 2016a), no occupational medical dose should be 

assigned.

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. 
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4.2 ENVIRONMENTAL EXTERNAL DOSE

Information on ambient external dose levels at the Apollo and Parks Township sites was 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 under EEOICPA. The 

health and safety coverage for both sites was 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 DU, NU, LEU (3.5%), and HEU (93%). Exposure to mixed fission and activation products 

was possible at some locations such as the 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 DU, NU, LEU (3.5%), and HEU 

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

fabrication (including 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. 

5.1.1 Uranium Source Information

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.

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 

Township. See Tables 2-1 and 2-3 for more information. The dose reconstructor should use the 

solubility type that results in the highest dose.

5.1.2 Plutonium Fuel Source Information

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

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

processed amount 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, pp. 328–329). Older materials, even when starting out as 

soluble, can have a tendency to oxidize when left in contact with air. Oxides, metals, and old 
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contamination should 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 (type SS) 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 2015). If the plutonium is type SS, follow guidance in ORAUT (2010a) 

for assignment of the 241Am solubility type. The type SS guidance is to be applied only to plutonium 

materials that are primarily 239Pu.

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) U-234 DU, NU

Uranium UO3 & UF4 (M) U-235 LEU (3.5%)

Uranium U3O8 & UO2 (S) U-238 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) 20% PuO2 and 80% UO2 About 4.5% 235U

MOXe UO2 (M, S) [7% plutonium – fuel 

grade/5% plutonium – 

weapons grade]

About 4.5% 235U

Fission and 

activation products

Unknown Be-7, Co-60, Sr-90, 

Ru/Rh-106, Cs-137, Tc-99 

(from ruthenium), Ir-192

Not applicable

a. Sources: Author unknown (ca. 2004); NUMEC (1963).

b. Type SS refers to highly insoluble plutonium.

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.

Table 5-2. Uranium source term information. 

Source 

term

Specific 

activity 

(pCi/μg)

U-234 

activity 

fraction

U-234 

activity

U-235 

activity 

fraction

U-235 

activity

U-236 

activity 

fraction

U-236 

activity

U-238 

activity 

fraction

U-238 

activity

NUa 0.683 0.489 0.334 0.023 0.016 Negligible Negligible 0.489 0.334

93%a 68.1 0.968 65.9 0.030 2.04 0.002 0.136 0.0003 0.020

3.5%a 2.20 0.818 1.80 0.034 0.075 Negligible Negligible 0.147 0.323

2%b 1.20 0.648 0.778 0.041 0.049 0.0009 0.001 0.311 0.373

Typical DUa 0.402 0.155 0.062 0.011 0.004 0.0005 0.0002 0.834 0.335

a. Source: Integrated Modules for Bioassay Analysis (IMBA). 

b. Source: American National Standards Institute N13.22 (HPS 2013). 

There are essentially three types of plutonium-based material: (1) reactor grade, (2) weapons grade, 

and (3) fuel grade, which falls between reactor and weapons grades. 

Fuel-grade plutonium was used for FFTF fuel fabrication and most of the ZPPR fuel fabrication 

(11,500 plates). However, commercial reactor-grade plutonium was also used for fabrication of 700 

ZPPR fuel plates starting in April of 1969. For work in the FFTF fuel fabrication, lacking specific 
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information on the actual composition of the processed plutonium, the composition of Hanford fuelgrade plutonium can be used because this was the source of plutonium for FFTF fuel fabrication 

(Author unknown 2004). The site also used Hanford 6% weapons-grade plutonium for some mixedoxide fuel fabrication. An 8.5% plutonium mix was used for a contract with Japan’s Power Reactor 

and Nuclear Fuel Development Corporation (PNC) and for ZPR-III fuel. The activity compositions for 

the four grades of plutonium are listed in Table 5-3 for fuel aged up to 20 years (ORAUT 2010b, 

2015). The age of plutonium to assume for a given analysis depends on the radionuclide measured in 

the bioassay analysis. When gross alpha, 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. 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. If the exposure occurred only in FFTF fuel fabrication, 

then the Hanford fuel-grade composition can be used. If the work location is not known (such as for 

plutonium scrap recovery), then use of the Hanford 6% weapons-grade plutonium will provide a dose 

estimate that is favorable to the claimant.

Table 5-3. Activity composition of plutonium mixtures (Ci/g).

a

Hanford 6% weapons-grade plutonium specific activityb

Nuclide Aged 0 yr Aged 5 yr Aged 10 yr Aged 15 yr Aged 20 yr

Pu-238 8.56E–03 8.23E–03 7.91E–03 7.60E–03 7.31E–03

Pu-239 5.77E–02 5.77E–02 5.77E–02 5.77E–02 5.77E–02

Pu-240 1.36E–02 1.36E–02 1.36E–02 1.36E–02 1.36E–02

Pu-241 8.24E–01 6.48E–01 5.09E–01 4.00E–01 3.15E–01

Pu-242 1.97E–06 1.97E–06 1.97E–06 1.97E–06 1.97E–06

Am-241 0.00E+00 5.83E-03 1.04E–02 1.39E–02 1.66E–02

8.5% plutonium specific activityc

Nuclide Aged 0 yr Aged 5 yr Aged 10 yr Aged 15 yr Aged 20 yr

Pu-238 1.22E–02 1.17E–02 1.12E–02 1.08E–02 1.04E–02

Pu-239 5.55E–02 5.55E–02 5.55E–02 5.55E–02 5.55E–02

Pu-240 1.94E–02 1.94E–02 1.94E–02 1.94E–02 1.94E–02

Pu-241 1.78E+00 1.40E+00 1.10E+00 8.64E–01 6.80E–01

Pu-242 2.80E–06 2.80E–06 2.80E–06 2.80E–06 2.80E–06

Am-241 0.00E+00 1.26E–02 2.25E–02 3.00E–02 3.58E–02

Hanford 12% fuel-grade plutonium specific activityb

Nuclide Aged 0 yr Aged 5 yr Aged 10 yr Aged 15 yr Aged 20 yr

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

Commercial 27% fuel-grade plutonium specific activityb

Nuclide Aged 0 yr Aged 5 yr Aged 10 yr Aged 15 yr Aged 20 yr

Pu-238 1.71E-01 1.64E-01 1.58E-01 1.52E-01 1.46E-01

Pu-239 3.41E-02 3.41E-02 3.41E-02 3.41E-02 3.41E-02

Pu-240 5.90E-02 5.89E-02 5.89E-02 5.89E-02 5.89E-02

Pu-241 1.34E+01 1.05E+01 8.28E+00 6.51E+00 5.12E+00

Pu-242 1.97E-04 1.97E-04 1.97E-04 1.97E-04 1.97E-04

Am-241 0.00E+00 9.49E-02 1.69E-01 2.26E-01 2.79E-01
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a. Aging refers to the time since separation of Am-241 from the plutonium mix.

b. Source: ORAUT (2010b).

c. Source: Gerrish (1965).

5.1.3 Plutonium-238 Source Fabrication

Plutonium-238 heat and heart pacemaker sources were prepared at NUMEC from 1963 to 1980. 

Operations involving 238Pu included the following: 

1. Conversion of 238Pu nitrate to the oxide in Building B – Plutonium Annex,

2. 238Pu pacemakers manufacture in Building B – Metals Plant, second floor, from 1963 to 1970,

3. 238Pu heat source manufacture in Building A – Fabrication Area 4 from 1963 to 1980,

4. Processing of 238Pu in Building B – Multipurpose Fabrication Building, and

5. Storage of fabricated sealed sources in Building A – Hot Cell Room from 1969 to 1980,

The work with 238Pu was performed in filtered gloveboxes. When a worker involved with 238Pu

processing or source fabrication has bioassay data for intake of 238Pu, the estimated intake and dose 

should be evaluated using guidance in DCAS-RPT-005, Alternative Dissolution Models for Insoluble 

Pu-238 (NIOSH 2016). That report describes a special dissolution model specific to intakes of 

insoluble 238Pu based on bioassay data for workers at other sites. The model parameters in Table 3-1 

for Los Alamos National Laboratory are to be used for evaluation of intakes by NUMEC workers. The 

intakes are to be evaluated using material absorption types M and S with the special dissolution 

model. The analysis that provides the highest dose should be applied in the dose estimate.

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 and the Parks Township site 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 (NIOSH 2007a, 2008a). 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.

5.2.1 Plutonium Urine and Feces Bioassay

Plutonium might have been present at the NUMEC sites in several forms that include types M, S, and 

possibly SS 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. Several different bioassay vendors performed plutonium urine and fecal analyses as 

observed from worker dosimetry records.
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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 for a period before the dates in the table, the values for the 

earliest date should be used. No bioassay monitoring results were found for 1985 to 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).
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Table 5-4. Plutonium and americium urine bioassay MDCs and frequencies by period.

a,b

Date Laboratory Analyte 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

01/1966–

12/1968

Eberline Pu-238 Quarterly/as 

needed

0.06 dpm/sample 0.03 dpm/sample

01/1966–

12/1975

Eberline Pu-239 Quarterly/as 

needed

0.06 dpm/sample 0.03 dpm/sample

01/1966–

12/1975

Eberline Am-241 Quarterly/as 

needed

0.06 dpm/sample 0.03 dpm/sample

01/1976–

04/1980

Controls for 

Environmental 

Pollution

Pu-238 As needed (f) (f)

01/1976–

04/1980

Controls for 

Environmental 

Pollution

Pu-239 Quarterly/as 

needed

(f) (f)

01/1976–

04/1980

Controls for 

Environmental 

Pollution

Am-241 Quarterly/as 

needed

(f) (f)

05/1980–

09/1985

Controls for 

Environmental 

Pollution

Pu-238 As needed (f) (f)

05/1980–

09/1985

Controls for 

Environmental 

Pollution

Pu-239 Quarterly/as 

needed

(f) (f)

05/1980–

09/1985

Controls for 

Environmental 

Pollution

Am-241 Quarterly/as 

needed

(f) (f)

1995 Eberline Pu-239/240 Unknown 0.1 pCi/L Not applicable

1996 -

1998

Thermo NU 

Tech

Pu-239/240 Unknown 0.1 pCi/L Not applicable

1999 Quanterra Pu-238 Unknown 0.0025–0.044 

pCi/L

Not applicable

1999 Quanterra Pu-239/240 Unknown 0.0025–0.045 

pCi/L

Not applicable

11/1974–

08/1975

Eberline Gross alpha 

(plutonium + 

americium)

Quarterly/as 

needed

<10.0 

dpm/sample

Not applicable

a. Based on review of worker dosimetry reports in Boyd (2006a, 2006b, 2006c, 2006d, 2006e, 2006f).

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.

Procedures used in later years to analyze for plutonium in urine 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 found. The estimated MDCs are 

listed in Table 5-5 for NUMEC facilities. 
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Table 5-5. Plutonium fecal bioassay MDCs and frequencies by period.

a,b

Date Laboratory Analyte Frequencyb MDCc Error

01/1966–

01/1977

Eberline Pu-239, Pu-238 or 

Am-241

Quarterly/as 

needed

0.1 dpm/sample 0.05 dpm/sample

05/1975–

09/1975

Eberline Gross alpha 

(plutonium + 

americium)

As needed 0.1 dpm/sample 0.05 dpm/sample

02/1977–

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, 2006b, 2006c, 2006d, 2006e, 2006f).

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.

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). 

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).

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), usually both fecal and urine 

samples were collected and the bioassay results were correlated with BZA samples. The suspected 

exposed worker was removed from radiation work and fecal and urine samples were collected. This 

was the method 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).
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Figure 5-1. Correlation of fecal bioassay with air sampling 

(Caldwell, Potter, and Schnell 1967).

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 from before late 1959.

Bioassay data in the available reports indicate uranium was analyzed in urine from about late 1959 

through 1988 and in 1999. 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 
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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 from Controls for Radiation, and 5 μg/L for results from Eberline in available 

bioassay reports. 

No information on sample analysis methods for other periods is available.

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,

2006b, 2006c, 2006d, 2006e, 2006f). If an MDC value is needed for periods 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.

The MDC value for the Controls for Radiation is 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 (Author unknown undated c). This value (1 μg/L) 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 but are sometimes 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,

• 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 
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Table 5-6. Uranium urine bioassay MDCs and frequencies by method and period.

a,b

Date Laboratory Analyte Frequencyb MDCc Error

03/1961–

02/1966

Controls for 

Radiation

Total uranium Quarterly/as 

needed

<1 μg/Ld Unknown

09/1972–

12/1976

Eberline Total uranium Quarterly/as 

needed

<5 μg/sample Unknown

01/1977–

11/1993

Controls for 

Environmental 

Pollution

Total uranium Unknown (e) (e)

1995 Eberline Total uranium Unknown <5 μg/L Unknown

1996 -

1998

Thermo NU 

Tech

Total uranium Unknown <5 μg/L Unknown

1999 Quanterra Total uranium Quarterly/as 

needed

<0.006 μg/L Unknown

04/1962–

01/1967

Controls for 

Radiation

Gross alpha Quarterly/as 

needed

26 dpm/L 13 dpm/L

02/1967–

08/1972

Tracerlab Gross alpha 

uranium

Quarterly/as 

needed

0.2 dpm/sample 0.1 dpm/sample

09/1972–

01/1974

Eberline Gross alpha 

uranium

Quarterly/as 

needed

<50.0 dpm/sample 0.05 dpm/mL

02/1974–

04/1974

Eberline Gross alpha 

uranium

Quarterly/as 

needed

<10 dpm/sample Unknown

02/1974–

12/1976

Eberline Gross alpha 

uranium

Quarterly/as 

needed

2 dpm/sample 1 dpm/sample

03/1964–

06/1967

Controls for 

Radiation

EU Quarterly/as 

needed

4 dpm/L 2 dpm/L

07/1967–

08/1972

Tracerlab EU Quarterly/as 

needed

0.2 dpm/sample 0.1 dpm/sample

01/1977–

02/1987

Controls for 

Environmental 

Pollution

EU Quarterly/as 

needed

(e) (e)

a. Based on review of worker dosimetry reports in Boyd (2006a, 2006b, 2006c, 2006d, 2006e, 2006f).

b. Records indicate quarterly monitoring for uranium workers unless an intake was suspected, which initiated more 

frequent special sample analyses.

c. When an MDC is not available in the records, assume 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.

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, then a radiometric analysis was 

performed if the level approached 50 μg/L. 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 an AEC hazard evaluation in 1959, a number of personnel had 

urine concentration results 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% 235U (NUMEC 1963). At 

some time in 1963 this was decreased to 300 dpm/L, and by October 1964 it 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). 
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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 found. 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. 

Table 5-7. Uranium fecal bioassay MDCs and frequencies by period.a

Date Laboratory Radionuclide Frequencyb MDCc

06/1967–06/1972 Tracerlab Radiometric uranium Quarterly/as needed 2 dpm/sample

07/1972–01/1976 Eberline Total uranium Quarterly/as needed <5 μg/sample

02/1976–10/1985 Controls for 

Environmental Pollution

Total uranium Quarterly/as needed (d)

a. Based on review of worker dosimetry reports in Boyd (2006a, 2006b, 2006c, 2006d, 2006e, 2006f).

b. Records indicate quarterly monitoring for uranium workers unless an intake was suspected, which initiated 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.

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.

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). 

5.2.2.4 Uranium Fluorimetric Bioassay Evaluations

The evaluation of intakes based on uranium fluorimetric bioassays requires conversion from units of 

mass to activity. This requires information on the 235U enrichment level because the specific activity of 

the uranium material varies with enrichment. If the enrichment is known for the exposed worker, then 

the specific activity for that enrichment should be used as defined in Table 5-2. If the enrichment is 

not known, then an estimate of activity that is favorable to the claimant can be made by assuming the 

material to be HEU (93%). This is a reasonable approach because HEU was used frequently at the 

Apollo facilities in the early years when urine bioassays were performed by fluorimetric analysis. 
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If the bioassay results contain both fluorimetric and radiometric results, then the radiometric results 

should be used because the radiometric analysis method provides a more sensitive estimate of 

uranium activity and no conversion of units is required.

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, Tracerlab reported 

the error as 0.1 dpm/sample for 232Th for a 100-mL urine sample, which provides a minimum 

detectable activity (MDA) value of 0.2 dpm/100 ml. The fecal analysis error for the same workers was 

reported as 0.1 dpm/100 ml, which provides an MDA value of 0.2 dpm/100 ml. 

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 (2014b), after preparation of the thorium dioxide from ore, the amount of 

progeny isotopes decreases and later builds in from continued decay of the 232Th. The thorium 

material processed at NUMEC was obtained from W. R. Grace which began separation of thorium ore 

in 1958 (ORAUT 2011). The maximum age after separation for the material at NUMEC at the end of 

thorium processing in 1970 would be 13 years. At 13 years, the amount of 228Ra is 79.1% of the 

232Th activity and the amount of 228Th is 70.1% of the 232Th activity. In addition, 228Ac and 224Ra would 

be present at the same percentage as 228Ra and 228Th, respectively. The internal dose is evaluated 

based on the estimated intake of 232Th plus the above percentages of the progeny. 

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 FPs. 

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 

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 

mixed FPs in urine. Urine bioassay data for mixed FPs should be used, if included in the case files, to 

estimate intakes of FPs. The mixed FP urine bioassay results do not indicate if the measurements 

are based on beta or gamma analysis, so the intake should be based on 90Sr as a representative beta 

emitter and 137Cs as a representative gamma emitter and the analysis providing the highest dose 

should be used. 

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 
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reprocessed to recover the uranium, which was returned to the DOE inventories along with trace 

contaminants. The intake of RU contaminant radionuclides can be estimated using the guidance in 

Section 5.3.2 and activity fractions in Tables 5-33 and 5-34 of ORAUT-TKBS-0017-5, Feed Materials 

Production Center – Occupational Internal Dose (ORAUT 2016b). The assignment of material 

solubility type is based on the guidance in Table 3-1 of ORAUT-OTIB-0060 (ORAUT 2014a). The 

determination of the form of uranium (types F, M, or S) most favorable to the claimant must include 

the recycled components for each uranium solubility type. Because plutonium is a minor contaminant 

in the RU matrix, the consideration of type SS 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 FPs 235U and 241Am, then 

estimated 239Pu 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 

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 the 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 

echoencephaloscope. The evaluation of 239Pu activity was based on the assumption that only 239Pu 

was present and that all 17-keV X-rays were from 239Pu. The difficulty in measuring the low-energy Xrays results in MDA values that represent significant lung burdens.

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.
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Table 5-8. Lung-counting MDAs of uranium based on enrichment.

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% U-235) 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 and 241Am (nCi).

a 

Year

Pu-239 

minimum

Pu-239 

maximum

Pu-239 

counts

Am-241 

minimum

Am-241 

maximum

Am-241 

counts

1968 Not reported Not reported Not reported 0.13 0.38 17

1969 Not reported Not reported Not reported 0.16 0.16 1

1970 10 10 1 Not reported Not reported Not reported

1971 Not reported Not reported Not reported Not reported Not reported Not reported

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 0.06 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, 2006c, 2006d, 2006g, 2006h, 2006i). Values for 1968 through 

1971 are based on the Helgeson system with remaining values from the University of Pittsburgh system.

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. If values are not provided in the records, then the 

maximum value from Table 5-9 should be used (nominally 35 nCi for 239Pu and 0.40 nCi for 241Am).

5.4 REPORTS OF OVEREXPOSURES AND INCIDENTS

The NUMEC Health Physics department 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.
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Overexposures were required to be reported to AEC or 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 it only if the bioassay data indicated an overexposure (Shapiro 1969).

5.4.1 Apollo Site Incidents

The following incidents occurred at the Apollo site 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 containing about 8.8 kg of HEU were damaged. 

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 was 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). 

5.4.2 Parks Township Site Incidents

Incidents occurred at the Parks Township site and might have contributed to employee exposures. 

Information about individual involvement in incidents is likely to be in worker dosimetry records and 

should be consulted for evaluation of intakes of radionuclides during such events. 

5.5 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 by the AEC Oak Ridge Operations Office R&D 
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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 in the 

1966 to 1967 timeframe indicated that there was little difference between fixed-station BZA and GA 

samplers. The correspondence between lapel sampler data and early fecal clearance for plutonium 

showed very good agreement, but fixed-station BZA and GA sampling usually underestimated 

airborne concentrations. Fifty percent of the lapel air sample results at the Apollo site showed 

concentrations 7 times greater than stationary air samples. The median of the ratio of lapel BZA to 

GA concentration results was found to be about 7 at the Apollo and Parks Township sites (Caldwell, 

Potter, and Schnell 1967). 

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 in 1963, 75 shortterm BZA samples were obtained in Apollo process buildings every week but no routine GA 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 were 1 × 10-10 μCi/mL or 220 dpm/m3 for in-plant 

airborne uranium (NUMEC 1963). 

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. 

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 than a maximum dose estimate. However, at the 

Apollo and Parks Township sites during the operational period, 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 (Leavitt 2007, 

2008). 

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 
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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 

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 (i.e., 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.

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.

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 Gamma and neutron radiation depending on source

Mixed plutonium and EU fuel fabrication Gamma and neutron radiation depending on source

HEU and LEU scrap recovery Gamma and neutron radiation depending on source

UO2 pellet production started in 1961 Gamma and neutron radiation depending on source

R&D for coating uranium particles started in 1961 Gamma and neutron radiation depending on source

Thorium operations and pellet production started 

in 1963

Beta radiation and more significant photon radiation

Laundry operations Uranium and thorium residues
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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 

4 mrad/hr at 30 cm from a drum of aged yellowcake (NIOSH 2014, p. 27).

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 occurred; levels are generally described in the Guide 

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. A description of the dosimetry report 

types that might be included in the NUMEC dosimetry records is given in Attachment B.

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 concern about film dosimetry over-response to the low-energy 

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. The Landauer K and P dosimeters were used to monitor for beta and 

gamma exposure. External results were found for the fourth quarter of 1982 through 1985 and July to 

December 1991 with dosimetry provided by Teledyne Isotopes (BWTX 1983, BWTX 1984, BWTX 

1985, BWXT 1991).
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Table 6-2. Dosimetry for external WB, wrist, and extremity exposures. 

Beta/Photon Dosimeters – WB 

Period Monitoring technique Dosimeter description

1957–

05/1968

Photographic film badge 

provided by Nuclear 

Science & Engineering, 

Controls for Radiation, 

Eberline, or Landauer

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 were similar to other dosimetry systems at that time as 

presented in the 1954 AEC dosimeter performance study (AEC 

1955).

06/1968–

1975

NUMEC or Eberline TLD Comprised of two TLD-700 chips, two TLD-600 chips, and one CaF2

for monitoring beta, X-ray, gamma and neutron exposure.

1976–

present

Landauer or equivalent 

TLD (Z1 dosimeter -1990)

Comprised of three TLD-700 chips for monitoring beta, X-ray, and 

gamma exposure, and a Neutrac 144 chip for monitoring fast 

neutron exposure. 

10/1982 –

1985 and 

7/1991 –

12/1991

Teledyne Isotopes TLD 

Badge

TLD badge for monitoring beta, gamma, and neutron exposure 

(BWTX 1983, BWTX 1984, BWTX 1985, BWXT 1991). Details of 

the dosimeter are not available, other than detection limits.

1995 – 2000 Landauer TLD (K 

dosimeter)

Comprised of three TLD-100 chips.

2000 Landaer TLD (P 

dosimeter)

Luxel dosimeter for monitoring X-ray, beta, and gamma exposure

(Boyd 2006j).

Beta/Photon Dosimeters – Wrist and Ring

Period Monitoring technique Dosimeter description

07/1963–

ca. 05/1968

Landauer (type M – wrist 

beta-gamma) film badges 

or equivalent

Film dosimeter known as type M responsive to beta and gamma 

radiation.

Ca. 06/1968–

1983

TLD wrist badge Comprised of three TLD-100 chips.

10/1982 –

1985 and 

7/1991 –

12/1991

Teledyne Isotopes TLD 

Badge

TLD badge for monitoring beta and gamma exposure (BWTX 1983, 

BWTX 1984, BWTX 1985, BWXT 1991). Details of the dosimeter 

are not available, other than detection limits.

Neutron Dosimeters – WB

Period Monitoring technique Dosimeter description

1957–

05/1968

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 1,000-

power microscope or projection capability.

06/1968–

1975

NUMEC or Eberline TLD Comprised of two TLD-700 chips, two TLD-600 chips, and one CaF2

for monitoring beta, X-ray, gamma and neutron exposure.

1975–1995 Landauer Neutrak 

Extended Range 

dosimeter (Z1 dosimeter)

Comprised of three TLD-700 chips for monitoring beta, X-ray, and 

gamma exposure, and a Neutrac 144 chip for monitoring fast 

neutron exposure. 

10/1982 –

1985 and 

7/1991 –

12/1991

Teledyne Isotopes TLD 

badge

Combined gamma, beta, and neutron TLD (BWTX 1983, BWTX 

1984, BWTX 1985, BWXT 1991). Details of the dosimeter are not 

available, other than detection limits.
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6.3.2 Neutron Dosimeters

Workers were monitored for neutron exposures with nuclear track emulsion, Type A (NTA) film from 

commercial vendors 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). Starting in 1968, TLDs 

were used to evaluate external dose. The in-house TLD program from 1968 through 1975 included 

monitoring for neutrons with the 5-chip dosimeter (two TLD-700 chips, two TLD-600 chips, and one 

CaF2). The Teldyne dosimeters included capability for neutron monitoring as indicated by records for 

1982 to 1985 and 1991 (BWTX 1983, BWTX 1984, BWTX 1985, BWXT 1991). The Landauer Z1 

dosimeter included neutron monitoring capability with a Neutrac 144 chip. The provider of dosimetry 

services is not always clear for the period from 1975 through the end of operations as the summary 

reports generated by the site do not indicate the vendor. There is no indication that monitoring for 

neutrons was performed after 1995.

6.3.3 Limits of Detection

External dosimetry technology minimum detectable levels (MDLs) are expected to have been 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 were higher. The film badge dosimetry at NUMEC was 

likely similar to dosimeters used at Hanford during the period 1957 to 1968 (ORAUT 2010b). 

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 through 1975 (ORAUT 2010b,

2010c). 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, External Dose Reconstruction Implementation Guideline (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. 

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.
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Table 6-3. WB LODs and potential missed photon, beta, or neutron dose. 

Vendor or processor/area monitored Period of use LODa (rem)

Annual missed doseb (rem)

(frequency in parenthesis)

Nuclear Science & Engineering or 

Controls for Radiation film and NTA film

1957–1963 0.03 photonsc

0.03 betac

0.05 neutrons

0.18 beta & photons (monthly)

0.30 neutrons [fast] (monthly)

Landauer film and NTA film 1964–1965 0.03 photonsc

0.03 betac

0.05 neutrons

0.18 beta & photons (monthly)

0.30 neutrons [fast] (monthly)

Eberline film and NTA film 1966–06/1968 0.03 photonsc

0.03 betac

0.05 neutrons

0.18 beta & photons (monthly)

0.30 neutrons [fast] (monthly)

NUMEC or Eberline (Film-Apollo) until 

about 1970 and TLD for neutron

07/1968–1975 0.02 photonsd

0.02 betad 

0.05 neutrons

0.12 beta & photons (monthly)

0.30 neutrons (monthly)

Landauer TLD and Teledyne Isotopes 1976–1995 0.01 photonse

0.01 betae

0.02 neutrons

0.06 beta & photons (monthly)

0.12 neutrons (monthly)

a. Estimated LODs 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. LODs for photons and beta during these periods are based on Hanford dosimeter values (ORAUT 2010b).

d. LOD during this period is probably twice the recording level of 0.010 rem.

e. Landauer LOD values from Boyd (2006j) and Teledyne Isotopes LOD values from BWXT (1991).

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 indicates that these represent the highest 

quality records for retrospective dose assessments (NIOSH 2007b). 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 AEC or 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 (numbers of individuals with 

WB doses in each range) (Breuer 1977, 1978).

Year

Total 

number 

monitored

Number with 

measured 

dose

<0.1 

rem

0.1–0.25 

rem

0.25–0.5 

rem

0.5–0.75 

rem

0.74–1.0 

rem

1.0–12 

rem

>12.0 

rem

1976 42 42 27 14 1 0 0 0 0

1977 39 39 15 16 6 0 2 0 0
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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

Information from the mid-1960s indicates there are potential limitations of the NUMEC film dosimeter 

to accurately measure low-energy photon radiation, such as that in 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 241

Am 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 

Figure 6-1. Gamma spectrum external to plutonium 

chemical processing line and plutonium ceramics 

line (Caldwell and Judd 1966, p. 14). 
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Figure 6-2. Gamma spectrum external to plutonium 

ceramics line and plutonium melt glovebox (Caldwell 

and Judd 1966, p. 15). 

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.

Figure 6-3. Dosimeter filter ratios (Caldwell and Judd 1966, p. 18).

The following interpretation was made by Caldwell and Judd (1966):
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1. Hot cell workers were exposed to 60Co and FPs. The energy response of the film badge was

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

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 2.0, NUMEC would use a plutonium spectrum calibration

curve that represented a typical plutonium gamma spectrum.

3. The OW:PL ratio was about 1.0 and was within a 6% standard deviation. This meant that the

large OW:CD ratio was 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 also included as well as 

233U and 241Am, which have similar overall photon and beta properties.

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. 19).

0

20

40

60

80

Fraction of 

Dose

20 kev60 kev 115

kev

195

kev

335

kev

Energy in kev

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

Energy NUMEC plutonium (Am-241 and U-233) Hanford plutonium

Photon <30 keV 15 25

Photon 30–250 keV 82 75

Photon >250 keV 3 0

Beta >15 keV 100 100
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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 2014c). However, at this time, the neutron dosimeter readings should be used 

without correction for this effect (also see Section 6.5.2.4). 

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.

Dosimeter type Period

Dosimeter measured 

quantities Compliance dose quantities

Photon/electron film 

dosimeter + NTA 

neutron dosimeter

1957–

1968

Gamma (G)

Neutron (N)

Beta (B)

WB or total = gamma (photon) + neutron

Beta separate

Extremity = gamma (+ neutron)

Photon/electron film 

dosimeter + TLD 

neutron dosimeter

1968–

1983

Deep = gamma (DBG) 

and neutron (N) 

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

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 

LOD.

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 

LOD.

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 

LOD. 

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
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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).

The guidance from ORAUT-OTIB-0017, Interpretation of Dosimetry Data for Assignment of Shallow 

Dose (ORAUT2005c) is as follows:

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 (DCFs) 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 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 2014c). This range includes 100% for HEU, EU, NU, and plutonium work 

locations. The recorded neutron doses during fuel work and neutron doses estimated using the 

neutron-to-photon ratios should, therefore, be multiplied by a factor of 1.91 (for ICRP Publication 60 

correction). 

6.5.2.4 Neutron Dose During NTA Film Period

NUMEC used NTA film to monitor for neutron exposure until approximately June 1968 when TLDs 

were introduced at the site. The NTA film dosimeters had a reduced response to neutrons at energies 

below about 500 to 800 keV (ORAUT 2014a). The missed neutron dose due to the energy threshold 

effect is accounted for by use of neutron-to-photon ratios.

A study was performed in 1975 (BWTX 1975) using TLDs to determine the photon and neutron doses 

while working in the FFTF fuel fabrication area of the Parks Township Plutonium Facility. In addition, 

dosimeters were placed at various fixed locations. The results of this study indicated the neutron-tophoton ratio for plutonium workers had a geometric mean of 0.34 and a GSD of 1.71. This ratio is 
Document No. ORAUT-TKBS-0041 Revision No. 04 Effective Date: 03/13/2017 Page 61 of 111

supported in the Parks Township Plutonium Facility 1979 to 1981 Health and Safety reports (BWTX 

1981; 1982a), which indicate that neutron-to-photon ratios varied from 0.23 to 0.42 with an average of 

0.33. The reported glovebox ratios had a geometric mean of 1.00 and a GSD of 1.49. 

The difference between the ratios is likely a result of shielding on the gloveboxes that would have 

reduced the photon component reaching the dosimeter.

Further, information was obtained from a September 1969 event in which a worker involved with 

manufacturing neutron sources had a neutron-to-photon ratio of 2.33 (determined using estimated 

neutron dose values) (Caldwell 1968d). 

Based on this information, NUMEC workers with neutron dosimetry based on NTA film dosimeters 

(i.e., through June 1968) should have neutron dose assigned using the ratio most appropriate for their 

work and job location (i.e., use a ratio of 0.34 for typical workers, ratio of 1.00 for glovebox workers, 

and ratio of 2.33 for workers involved with manufacturing neutron sources). This dose is assigned 

using a lognormal distribution with the GSD provided, except for the manufacturing neutron source 

ratio, which should be applied as a constant. If the worker’s recorded neutron dose is higher than the 

neutron dose calculated from applying the ratio to the photon dose, the recorded neutron dose should 

be assigned. 

6.5.3 Missed and Unmonitored Dose

The potential for missed dose exists when workers were 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 OCASIG-001 (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 should be followed 

for assignment of missed shallow and deep doses (ORAUT 2005c). 

6.5.3.2 Neutron Dose

The potential missed neutron dose can be estimated from LOD values in Table 6-3 for monitored 

workers using the same approach. If the neutron-to-photon method is used to evaluate the neutron 

dose per guidance in Section 6.5.2.4, then use the neutron-to-photon ratio to assign missed neutron 

dose. When measured NTA film dose is used per guidance in Section 6.5.2.4, then use the LOD 

method to assign missed neutron dose. 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 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:
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• The technology used to measure worker dose at NUMEC is similar to the technology that was

used by commercial and government 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 Begin End

Radiation 

type

Energy 

selection Percent

Uranium facilities 01/01/1957 12/31/1983 Beta >15 keV 100

Uranium facilities 01/01/1957 12/31/1983 Photon 30–250 keV 100

Plutonium facilities 01/01/1959 12/31/1980 Photon 30–250 keV 100

Plutonium facilities 01/01/1959 12/31/1980 Neutron 0.1–2 MeV 100

Thorium handling 01/01/1957 12/31/1983 Beta >15 keV 100

Thorium handling 01/01/1957 12/31/1983 Photon 30–250 keV 25

Thorium handling 01/01/1957 12/31/1983 Photon >250 keV 75

Photon and neutron sources 01/01/1957 12/31/1983 Beta >15 keV 100

Photon and neutron sources 01/01/1957 12/31/1983 Photon 30–250 keV 50

Photon and neutron sources 01/01/1957 12/31/1983 Photon >250 keV 50

Photon and neutron sources 01/01/1957 12/31/1983 Neutron 0.1–2 MeV 100

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 is 

from 1984 through 1995, and the residual period for the Parks Township site is from 1981 through 

2004.
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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 suggests there was 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, but 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 have been 

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 for uranium of 6.39 × 105 dpm/m2 (GSD = 6.95) and a mean 

surface concentration of thorium and progeny of 1.54 × 105 dpm/m2 (GSD = 5.0). These values 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 DCF for exposure to uniform activity on a ground plane: 

( ) ( ) ( ) ( ) ( ) (7-1) 2 2 Dose residual level rem/yr = dpm/m rem/hr / dpm/m DCF   × exposure time hr/yr   ×

The external DCF for exposure to isotopes of uranium, thorium, 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 WB exposure and skin exposure. 

Values are presented for uranium activity, thorium activity, and the total from both sources. These 

values are based on an exposure time of 2,000 hr/yr. The dose to a specific organ can be obtained 

by multiplying the WB exposure value by the isotropic exposure DCF for the organ of interest. The 

listed WB dose values are based on the effective WB dose from Federal Guidance Report 12 using 

ICRP Publication 60 weighting factors (ICRP 1991). The values have been adjusted to ensure that 

application of the exposure isotropic dose factors results in an organ dose that is not less than the 

Federal Guidance Report 12 value for the organ.

Table 7-1. External annual dose (rem) from residual surface contamination at the Apollo site.

Organ Uranium annual dosea Thorium annual doseb Total annual dose

Whole body 0.0005 0.0032 0.004

Skin 0.0354 0.0088 0.044

a. Uranium values are input into IREP as a lognormal distribution; value listed is the geometric mean; the GSD

is 6.95.

b. Thorium values are input into IREP as a lognormal distribution; value listed is the geometric mean; the GSD

is 5.0.

The uranium 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 
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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 m 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. For dose to shallow organs (i.e., breast, testes, and penis), the skin dose values should be 

assigned with a correction factor of 0.30 (ORAUT 2005c). The shallow dose to the lens of the eye 

should be based on the assigned skin dose values with a correction factor of 0.33 (ORAUT 2010d).

The method for assignment of external dose from residual materials at the Apollo site is as indicated 

in Table 7-2. The following is based on guidance in the technical information bulletin for shallow dose 

(ORAUT 2005c) and the procedure for occupational onsite external dose (ORAUT 2006). This 

method includes application of the exposure isotropic DCFs and results in a best estimate of dose.

Table 7-2. Assignment of external onsite residual dose.

Organ(s) Dose assignment

All except skin, breast, testes, and penis Listed WB dose times organ isotropic exposure DCF

Skin Listed skin dose

Breast, testes, and penis Listed WB dose times organ isotropic exposure DCF

Breast, testes, and penis 30% of listed skin dose

Lens of eye 33% of listed skin dose

The dose should be entered into the IREP input as a lognormal distribution with a GSD of 5.0 (IREP 

Parameter 2) as photons of energy from 30 to 250 keV. This provides a favorable estimate of POC 

for all organs, even though some of the photon energy is likely to be of higher energy. Although much 

of the dose to the skin is from electrons, assignment of the dose as photons provides a result that is 

favorable to claimants. 

7.1.2 Parks Township Site

Because the residual period for the Parks Township site includes the period from 1981 through 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. There might be external dosimetry data for later 

years of the residual period, but operations then were not part of the nuclear weapons-related 

program and are not covered for dose reconstruction under EEOICPA (Leiton 2011). 

Activities 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 favorable to the claimant. 

An estimate of the external dose from residual plutonium activity at the Parks Township site has been 

made using the same approach as 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 2.26 × 104 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 and DCFs for exposure to uniform activity on a ground 

plane. The annual dose was evaluated using the Hanford fuel-grade plutonium isotopic compositions 

in Table 7-6 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 
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need be assigned for the residual period for Parks Township workers based on potential exposure to 

plutonium for Building A.

Parks Township also performed work with HEU in Building C. No information has been found about 

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, 1960b, 1960c, 1961a, 1961b), the residual 

activity levels for the Apollo site provide an estimate of external dose from residual activity at the 

Parks Township HEU operations in Building C that is favorable to claimants. The annual external 

dose values for the Parks Township residual period are the same as those given in Table 7-1 for 

uranium.

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 and thorium contamination. Valid bioassay data are unlikely to be available for 

the residual period. However, monitoring data can be used instead of the default intake assumptions 

given below to limit dose, as appropriate (NIOSH 2008b). Intakes of both uranium and thorium should 

be included during the Apollo residual period.

ORAUT-OTIB-0070, Guidance in Dose Reconstruction During Residual Radioactivity Periods at 

Atomic Weapons Employer Facilities (ORAUT 2012), 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 involved 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, 1960b, 1960c, 1961a, 1961b). The results of these studies are described in 

Section 5.0. 

The mean average daily air concentration can be used to estimate the residual surface concentration 

using guidance from ORAUT (2012). The mean average air concentration is based on the long-term 

average using GA monitoring results for the site. The most appropriate data are those taken using 

continuous air monitors that include all variations in air concentration over the year. Such data were

found for uranium for the year 1967 for several of the operations at the Apollo uranium facility. 

Limited data were found for 1966 for the Hammer Mill and the ceramics area, and for a few periods in 

1962 and 1963 for the CRP-3 beryllium glovebox area and process area. The results from these 

measurements and the HASL reports indicate an air concentration value that is favorable to claimants 

to use for the residual period is represented by the maximum median value of 329 dpm/m3

. The 
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maximum estimated GSD for all datasets is 6.95 for the HASL GA data. Therefore, the residual 

activity is based on a lognormal distribution with a median of 329 dpm/m3 and a GSD of 6.95. The 

deposition amount is estimated using a deposition velocity of 0.00075 m/s with deposition assumed to 

occur for 30 days (NIOSH 2011). Using this approach, a surface concentration of uranium is 

estimated as follows: 

3 5 2 329 dpm/m 2 592 000 s/30 d 0.00075 m/s = 6.39 10 dpm/m ×× × , , (7-2)

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 2012). The depletion factors applied to each year are 

listed in Table 4-2 of ORAUT (2012). 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 the factors in Table 4-2 of ORAUT (2012).

The air concentration for each year is estimated using a resuspension factor (ORAUT 2012) of 1 × 10-

5

/m. This resuspension factor is appropriate for situations where moderate work activity continued 

during the residual period, such as for the Apollo and Parks Township uranium facilities. Application 

of this resuspension factor and the above-described depletion factors to the residual contamination 

level of 6.39 × 105 dpm/m2 results in the air concentration and annual intakes in Table 7-3. The intake 

evaluation is based on exposure for 2,000 hr/yr and an inhalation rate of 1.2 m3

/hr.

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 6.95, corresponding to the distribution of the 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 (F, M, and S), and the dose from 

the highest type should be used in the IREP input.

Table 7-3. 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 6.39 1.53E+04 16,400

1985 5.00 1.20E+04 12,900

1986 3.92 9.40E+03 10,100

1987 3.07 7.36E+03 7,900

1988 2.40 5.77E+03 6,200

1989 1.88 4.51E+03 4,800

1990 1.48 3.54E+03 3,800

1991 1.16 2.78E+03 3,000

1992 0.90 2.16E+03 2,300

1993 0.71 1.70E+03 1,800

1994 0.55 1.33E+03 1,430

1995 0.43 1.04E+03 1,120

a. Doses are assigned as a lognormal distribution with a GSD of 6.95.

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 

ingestion of uranium activity based on guidance in OCAS-TIB-009, Estimation of Ingestion Intakes

(NIOSH 2004). 
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The daily intake rate (dpm/d) is estimated to be 0.2 times the average daily air concentration 

(dpm/m3

). Using the air concentration at the end of the operating period of 329 dpm/m3, an intake 

rate of 16,400 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 (2012). This provides an assessment of ingestion intake that is favorable to claimants.

Similar data were found for thorium operations in 1964 and 1965, which were also conducted in the 

Apollo uranium facility. The GA samples were not continuous, but were taken for periods from a few 

minutes to about an hour in areas where work was being conducted with thorium. The 71 data points 

indicated a median thorium concentration of 112 dpm/m3

, a mean of 160 dpm/m3, and a GSD of 2.4. 

The samples were counted immediately at the end of the sample period and also after several hours 

to allow for decay of natural atmospheric radon and progeny. The results used in this analysis were 

those taken after decay because they represent the alpha activity from long-lived radionuclides. 

Based on these data, residual activity for thorium can be evaluated using a lognormal distribution of 

112 dpm/m3 and a GSD of 5.0 based on recommendations of Davis and Strom (2008). The higher 

GSD is used because it includes uncertainties associated with measurement errors and other errors 

not represented by the variability in the numerical values alone. The thorium at Apollo was received 

as thorium oxide that had undergone only one separation. The measured activity was assumed to be 

232Th with the minimum 228Th following separation, an assumption favorable to claimants. This results 

in an initial activity of 79 dpm/m3 232Th. Because exposures in the residual period are about 20 to 37 

years after the initial processing of thorium at NUMEC, the 232Th is assumed to be in equilibrium with 

all progeny (ORAUT 2014b). The activity of 232Th and each progeny isotope (after decay to the start 

of the residual period) is assumed to be 79 dpm/m3

. The deposition amount is estimated using a 

deposition velocity of 0.00075 m/s with deposition assumed to occur for 30 days (NIOSH 2011). 

Using this approach, a surface concentration of 232Th is estimated as follows: 

3 5 2 79 dpm/m 2 592 000 s/30 d 0.00075 m/s = 1.54 10 dpm/m ×× × , , (7-3)

This mean surface concentration is described as a level at the end of the operating period that is 

favorable to claimants. 

The 232Th air concentration for each year is estimated using a resuspension factor (ORAUT 2012) of 

1 × 10-5/m as discussed above for uranium. Application of this resuspension factor and the abovedescribed depletion factors to the residual thorium contamination level of 1.32 × 106 dpm/m2 results in 

the air concentration and annual intakes in Table 7-4. The intake evaluation is based on exposure for 

2,000 hr/yr and an inhalation rate of 1.2 m3

/hr. The potential ingestion intake is also included in the 

table. Because the remediation work could have been as long as 37 years after the initial separation 

of the thorium from the ore by W. R. Grace (ORAUT 2011), the activity of 232Th progeny could 

approach that of the 232Th. The amount of 228Ra, 228Ac, 228Th, and 224Ra are assumed to be 100% of 

the 232Th activity. Therefore, the intake activities in the table should be assigned to each of the 

following radionuclides 232Th, 228Ra, 228Ac, 228Th, and 224Ra as alpha radiation in IREP. 
Document No. ORAUT-TKBS-0041 Revision No. 04 Effective Date: 03/13/2017 Page 68 of 111

Table 7-4. Th-232 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 1.54 3.69E+03 4,000

1985 1.20 2.89E+03 3,100

1986 0.94 2.26E+03 2,400

1987 0.74 1.77E+03 1,900

1988 0.58 1.39E+03 1,500

1989 0.45 1.08E+03 1,200

1990 0.35 8.52E+02 900

1991 0.28 6.67E+02 700

1992 0.22 5.20E+02 600

1993 0.17 4.09E+02 400

1994 0.13 3.20E+02 340

1995 0.10 2.50E+02 270

a. Doses are assigned as a lognormal distribution with a GSD of 5.0.

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). Bioassay monitoring 

data for uranium might be available for the first 3 years of the Parks Township residual period (1981 

through 1983). If the worker might have worked at the Apollo site during this period, the data could 

relate to DOE work and should 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 describes methods to estimate intake of radionuclides during 

residual periods (ORAUT 2012). 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 dataset 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.
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Figure 7-1. Reported plutonium air concentrations.

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

The median plutonium air concentration from the above-described study and the guidance from 

ORAUT (2012) were used to estimate the residual surface concentration. The deposition amount was 

estimated using a deposition velocity of 0.00075 m/s, with deposition assumed to occur for 30 days 

(NIOSH 2011). Using this approach, a surface concentration of plutonium was estimated as follows: 

×× × 3 4 2 11 6 dpm/m 2 592 000 s/30 d 0.00075 m/s = 2.26 10 dpm/m . ,, (7-4)

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 2012). The depletion factors that were applied to each year are described in Table

4-2 of ORAUT (2012). 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 (2012).

The air concentration for each year was estimated using a resuspension factor of 1 × 10-6/m (ORAUT 

2012). This resuspension factor is based on the Parks Township Plutonium facility having been 

cleaned prior to continued use, resulting in less loose material being available for resuspension. 

Application of this resuspension factor and the above-described depletion factors to the residual 

contamination level of 2.26 × 104 dpm/m2 resulted in the air concentration and annual intakes in Table

7-5. The intake evaluation was based on exposure for 2,000 hr/yr and an inhalation rate of 1.2 m3

/hr. 
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Table 7-5. Plutonium air concentration and annual alpha activity 

intake in the Parks Township site residual period.

Year

Air concentration 

(dpm/m3)

Inhalation 

intake 

(dpm/yr)a

Ingestion 

intake (dpm/yr)a

1981 0.023 54 580

1982 0.018 42 450

1983 0.014 33 360

1984 0.011 26 280

1985 0.008 20 220

1986 0.0066 16 170

1987 0.0052 13 130

1988 0.0041 10 100

1989 0.0032 7.6 82

1990 0.0025 6.0 64

1991 0.0020 4.7 50

1992 0.0015 3.7 39

1993 0.0012 2.9 31

1994 0.0009 2.3 24

1995 0.00074 1.8 19

1996 0.00058 1.4 15

1997 0.00045 1.1 12

1998 0.00035 0.84 9.0

1999 0.00028 0.67 7.1

2000 0.00022 0.52 5.6

2001 0.00017 0.41 4.4

2002 0.00013 0.32 3.4

2003 0.00010 0.25 2.7

2004 0.00008 0.20 2.1

a. Doses are assigned as a lognormal distribution with a GSD of 4.97.

Because the plutonium is based on gross alpha GA monitoring results, the activity represents the total 

alpha activity. The dose from the residual activity should be evaluated using the fractional isotopic 

alpha activity of each radionuclide given in Table 7-6. Activity fractions for this conversion are given in 

Table 7-6 for the plutonium used for ZPR-III and PNC, the Hanford fuel-grade plutonium used for 

FFTF and ZPPR fuel fabrication, and for commercial reactor-grade plutonium used for some of the 

ZPPR fuel fabrication. The total alpha activity values in Table 7-5 are to be multiplied by the activity 

fractions in Table 7-6 to determine the intake of each radionuclide. The age of the material must also 

be considered because the maximum dose (per unit alpha activity) is obtained for the less aged 

material. The beginning of plutonium processing at the NUMEC plutonium facility was in 1962. 

Therefore, the age of the residual material in 1981 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. Guidance on selection of material age is given in ORAUT (2014a). If 

the exposure occurred only in FFTF fuel fabrication, then the Hanford fuel-grade composition may be 

used. If the work location is not known (such as for plutonium scrap recovery), or if the work involved 

ZPPR fuel fabrication, then both fuel types should be considered and the fuel type giving the highest 

dose should be used.

The intakes in Table 7-5 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 air concentrations that were used to 

estimate the annual plutonium intake. The dose should be evaluated for types M and S plutonium, 

and the dose from the highest type should be used in the IREP input. Because the residual plutonium 
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Table 7-6. Alpha activity fraction for plutonium material at the Parks Township plutonium facility.

Hanford fuel-grade plutonium

Isotope Aged 0 yr Aged 5 yr Aged 10 yr Aged 15 yr Aged 20 yr

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 (beta) 31.9 20.6 14.2 10.2 7.5

Am-241 0.00 0.19 0.29 0.35 0.40

ZPR-III and PNC plutonium

Isotope Aged 0 yr Aged 5 yr Aged 10 yr Aged 15 yr Aged 20 yr

Pu-238 0.14 0.12 0.10 0.09 0.09

Pu-239/240 0.86 0.76 0.69 0.65 0.62

Pu-241 (beta) 20.4 14.1 10.1 7.5 5.6

Am-241 0.00 0.13 0.21 0.26 0.30

Commercial reactor-grade plutonium

Isotope Aged 0 yr Aged 5 yr Aged 10 yr Aged 15 yr Aged 20 yr

Pu-238 0.65 0.46 0.38 0.32 0.29

Pu-239/240 0.35 0.26 0.22 0.20 0.18

Pu-241 (beta) 50.6 29.7 19.7 13.8 10.0

Am-241 0.00 0.27 0.40 0.48 0.55

is in the form of aged material, type SS plutonium should be considered and adjustments made based 

on guidance in ORAUT-OTIB-0049 (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 (2012). This provides an assessment of 

ingestion intake that is favorable to claimants.

The intakes in Table 7-5 relate to DOE work in Building A of the Parks Township site. Work with 

uranium in Building C could also have contributed to internal dose for workers during the residual 

period. However, no information was found to allow determination of the residual uranium activity 

from work in Building C. For workers who are known to have been exposed from Building C residual 

activity, the internal dose can be evaluated based on intakes given in Table 7-3, with intakes shifted 

back 3 years to account for the difference in residual period between the Apollo and Parks Township 

sites. These intakes are given in Table 7-6 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 7-5) and uranium (Table 7-7) can be assigned as favorable to the claimant.
Document No. ORAUT-TKBS-0041 Revision No. 04 Effective Date: 03/13/2017 Page 72 of 111

Table 7-7. 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 6.39 1.53E+04 16,400

1982 5.00 1.20E+04 12,900

1983 3.92 9.40E+03 10,100

1984 3.07 7.36E+03 7,900

1985 2.40 5.77E+03 6,200

1986 1.88 4.51E+03 4,800

1987 1.48 3.54E+03 3,800

1988 1.16 2.78E+03 3,000

1989 0.90 2.16E+03 2,300

1990 0.71 1.70E+03 1,800

1991 0.55 1.33E+03 1,430

1992 0.43 1.04E+03 1,120

1993 0.34 8.16E+02 870

1994 0.27 6.38E+02 680

1995 0.21 5.00E+02 540

1996 0.16 3.91E+02 420

1997 0.13 3.07E+02 330

1998 0.10 2.39E+02 260

1999 0.079 1.89E+02 200

2000 0.061 1.47E+02 160

2001 0.048 1.15E+02 123

2002 0.038 9.02E+01 97

2003 0.029 7.07E+01 76

2004 0.023 5.54E+01 59

a. Doses are assigned as a lognormal distribution with a GSD of 6.95.

7.5 REMEDIATION WORKER DOSE DURING RESIDUAL PERIODS

Dose to remediation workers during decontamination and decommissioning (D&D) activities may 

involve exposure to DOE-related materials that exist during the residual periods for the NUMEC sites. 

If a NUMEC worker is known to have been involved in remediation work during the residual periods, 

then bioassay monitoring and external monitoring records can be applied in the evaluation of dose. 

The periods of final D&D for the NUMEC facilities performed during the residual periods are indicated 

in Table 7-8.

Table 7-8. Decontamination and Decommissioning during Residual Periods. 

Facility Start End Comment

Apollo – Uranium 1984 1995 Final D&D operations

Parks – Plutonium 1981 1983 Glovebox and fume hood removal

Parks – Plutonium 1998 2004 Final D&D operations

Parks – HEU 1998 2004 Final D&D operations

Because work during the residual period with non-DOE materials was not on the same scale as 

during the operational period, the bioassay results during the D&D periods for remediation workers 

can be assumed to be related totally to DOE work. Bioassay monitoring from CEP cannot be used to 

evaluate internal dose. The CEP results for urine and fecal sample analysis covers the residual 

period up to 1993. No results have been observed for 1994. Eberline provided internal bioassay 

sample analysis in 1995 and Quanterra provided sample analysis starting in 1996. The bioassay 

results from Eberline and Quanterra can be used to estimate internal exposure. In vivo monitoring 

results can be used in the internal dose assessment.
Document No. ORAUT-TKBS-0041 Revision No. 04 Effective Date: 03/13/2017 Page 73 of 111

If external monitoring results are available during the residual D&D periods listed in Table 7-8, the 

results can be assumed to be related to DOE materials if it is known that the worker was involved in 

remediation work. If the worker was not involved in remediation work then internal and external doses 

should be based on the guidance for exposure to residual contamination.

When no external monitoring results are available for a remediation worker during the D&D periods, 

the residual external doses should be applied as defined earlier. External exposure comes primarily 

from the material on surfaces and is not impacted by airborne materials during operations that may 

disturb some of the residual materials.

When bioassay monitoring results are not available for a remediation worker during the D&D periods, 

the residual internal doses defined earlier should be multiplied by a factor of 10 and applied for the 

years the worker was involved in D&D activities. This accounts for enhanced air concentrations 

resulting from disturbing the residual activity. For uranium facilities, this corresponds to a 

resuspension factor of 1.E-4 and for the plutonium facility it corresponds to a resuspension factor of 

1.E-5. The difference in resuspension factors results from the more thorough remediation performed

for the plutonium facility at the end of the operating period. This approach is favorable to the 

claimants as the contamination is assumed to be all related to the residual activity from earlier DOE 

work; contributions from non-DOE operations is not considered. Also, remediation workers would be 

required to wear clothing and protective equipment to minimize exposure to contamination. 

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. 04 Effective Date: 03/13/2017 Page 74 of 111

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Laboratory, the Canoga Avenue Facility, the Downey Facility, and the De Soto Avenue Facility 

(sometimes referred to as Energy Technology Engineering Center [ETEC] or Atomics 

International) – Occupational External Dose, ORAUT-TKBS-0038-6, Rev. 02, Oak Ridge, 

Tennessee, April 26.

ORAUT (Oak Ridge Associated Universities Team), 2010d, Monitoring the Eye Lens Dose: Statement 

by the German Commission on Radiological Protection with Scientific Reasoning (English 

Translation), Geschäftsstelle der Strahlenschutzkommission, Bonn, Germany, February 2. 

[SRDB Ref ID: 135905]

ORAUT (Oak Ridge Associated Universities Team), 2011, An Exposure Matrix for W.R. Grace and 

Company in Erwin, Tennessee, ORAUT-TKBS-0043, Rev 02, Oak Ridge, Tennessee, 

September 16.

ORAUT (Oak Ridge Associated Universities Team), 2012, Dose Reconstruction During Residual 

Radioactivity Periods at Atomic Weapons Employer Facilities, ORAUT-OTIB-0070, Rev. 01, 

Oak Ridge, Tennessee, March 5. 

ORAUT (Oak Ridge Associated Universities Team), 2014a, Internal Dose Reconstruction, ORAUTOTIB-0060, Rev. 01, Oak Ridge, Tennessee, September 8.

ORAUT (Oak Ridge Associated Universities Team), 2014b, Guiding Reconstruction of Intakes of 

Thorium Resulting from Nuclear Weapons Programs, ORAUT-OTIB-0076, Rev. 00, Oak 

Ridge, Tennessee, July 10.

ORAUT (Oak Ridge Associated Universities Team), 2014c, Argonne National Laboratory – East – 

External Dosimetry, ORAUT-TKBS-0036-6, Rev. 00, Oak Ridge, Tennessee, October 16. 

ORAUT (Oak Ridge Associated Universities Team), 2015, Hanford Site – Occupational Internal Dose, 

ORAUT-TKBS-0006-5, Rev. 06, Oak Ridge, Tennessee, November 16.

ORAUT (Oak Ridge Associated Universities Team), 2016a, Guidance on Assigning Occupational 

X-Ray Dose Under EEOICPA for X-Rays Administered Off Site, ORAUT-OTIB-0079, Rev. 01, 

Oak Ridge, Tennessee, March 18.
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ORAUT (Oak Ridge Associated Universities Team), 2016b, Feed Materials Production Center – 

Occupational Internal Dose, ORAUT-TKBS-0017-5, Rev. 02, Oak Ridge, Tennessee, 

September 30. 

PDEP (Pennsylvania Department of Environmental Protection), 2008, “Babcock & Wilcox (B&W); 

Apollo, PA, NRC license SNM-145 (Terminated), NRC Docket 070-135,” Commonwealth of 

Pennsylvania. [SRDB Ref ID: 43949]

Puechl, K. H., 1965, “Application for Renewal of Special Nuclear Materials License SNM-414 

Covering Operations in the Metals Plant,” letter with attachment to D. A. Nussbaumer (U.S. 

Atomic Energy Commission), Nuclear Materials and Equipment Corporation, Apollo, 

Pennsylvania, June 14. [SRDB Ref ID: 28510]

Reitler, E. K. Jr., 1972, “Renewal Application for SNM-145, Docket 70-135,” letter with attachment to 

R. B. Chitwood (U.S. Atomic Energy Commission), Nuclear Material and Equipment 

Corporation, Apollo, Pennsylvania, October 31. [SRDB Ref ID: 19961]

Reitler, E. K. Jr., 1973a, “Application for Amendment to SNM-414, Docket 70-364 to Operate 

NUMEC’s Type II Facility,” letter with attachment to R. B. Chitwood (U.S. Atomic Energy 

Commission), Nuclear Materials and Equipment Corporation, Apollo, Pennsylvania, 

February 20. [SRDB Ref ID: 19774]

Reitler, E. K. Jr., 1973b, “Incident Report -- SNM-145,” letter to F. E. Kruesi (U.S. Atomic Energy 

Commission), Nuclear Materials and Equipment Corporation, Apollo, Pennsylvania, 

February 23. [SRDB Ref ID: 19878]

Rouse, L. C., 1979, “70-364, License No. SNM-414, Amendment No. 5,” letter with attachment to M. 

A. Austin (Babcock & Wilcox Company), U.S. Nuclear Regulatory Commission, Washington, 

D.C., August 28. [SRDB Ref ID: 10435, p. 74]

Shapiro, Z. M., 1969, untitled letter to L. D. Low (U.S. Atomic Energy Commission), Nuclear Materials 

and Equipment Corporation, Apollo, Pennsylvania, June 12. [SRDB Ref ID: 19858]

Taylor, G. A., K. W. Crase, T. R. LaBone, and W. H. Wilkie, 1995, A History of Personnel Radiation 

Dosimetry at the Savannah River Site, WSRC-RP-95-234, Westinghouse Savannah River 

Company, Aiken, South Carolina, May. [SRDB Ref ID: 10931]

Thornton, W. T., and W. A. Johnson, 1964, Health Protection Review of Nuclear Materials and 

Equipment Corporation, October 1964, U.S. Atomic Energy Commission, Oak Ridge 

Operations Office, Oak Ridge, Tennessee, October. [SRDB Ref ID: 11961, p. 2]

USACE (U.S. Army Corps of Engineers), 2006, “Shallow Land Disposal Area, Parks Township 

Pennsylvania,” Pittsburgh District, Pittsburgh, Pennsylvania, September 19. [SRDB Ref ID: 

30955]
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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 insoluble in the 

respiratory tract (slow solubilization). Also called solubility type.

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, 
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excluding uranium mining and milling, and (B) is designated by the Secretary of Energy as an 

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. Also called radiobioassay. 

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. Examples include the tailings or wastes from the extraction or 

concentration of uranium or thorium from ore.

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 interval or 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 undesired locations including air, soil, buildings, animals, and persons.
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control rod

Neutron-absorbing device in a reactor used to slow or speed the reaction. Also called safety 

rod.

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.

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 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. 

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.
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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. 

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)

(1) Radionuclides produced by fission or by the subsequent radioactive decay of 

radionuclides. (2) Fragments other than neutrons that result from the splitting of an atomic 

nucleus.

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.
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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.

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.
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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.

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).

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, 
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monitoring includes internal and external dosimetry including interpretation of the 

measurements.

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 made by the Eastman Kodak Company that is sensitive to fast neutrons. The developed 

image has tracks caused by neutrons that become visible under oil immersion with about 

1,000-power magnification. The number of tracks in a given area is a measure of the dose 

from that radiation.

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 the Energy Employees Occupational Illness 

Compensation Program Act of 2000. 

open window (OW)

Area of a film dosimeter that has little to no radiation shielding (e.g., only a holder and visible 

light protection). The open window measures nonpenetrating as well as penetrating dose, 
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which minimizes the potential for beta radiation to contribute to the interpreted penetrating 

dose. See film dosimeter. 

penetrating dose

Dose from moderate to higher energy photons and neutrons that penetrates the outer layers of 

the skin. See dose. 

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. These are 

noted as Hp(0.07) and Hp(10), respectively. In 1993 the International Commission on 

Radiological Measurement and Units recommended Hp(d) as the 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 Program Act of 2000, 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

(1) Any object or substance that emits radiation. (2) Package of radioactive material 

constructed to have specific radiation properties used, for example, for medical purposes or to 

calibrate dosimeters.

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. 
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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. 

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:

(A) The employee was so employed for a number of work days aggregating at least 250 

work days before February 1, 1992, at a gaseous diffusion plant located in Paducah, 

Kentucky, Portsmouth, Ohio, or Oak Ridge, Tennessee, and, during such 

employment— 

(i) was monitored through the use of dosimetry badges for exposure at the plant of 

the external parts of employee’s body to radiation; or

(ii) worked in a job that had exposures comparable to a job that is or was monitored 

through the use of dosimetry badges.

(B) The employee was so employed before January 1, 1974, by the Department of 

Energy or a Department of Energy contractor or subcontractor on Amchitka Island, 

Alaska, and was exposed to ionizing radiation in the performance of duty related to 

the Long Shot, Milrow, or Cannikin underground nuclear tests.
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(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. The sample 

standard deviation is the square root of the sample variance. This means that it has the same 

linear units as the original data values or a measure of central tendency, instead of the 

squared units of the sample variance.

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.

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. Also called crystals.

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 interval or level, and error. Also called standard 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.
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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. 
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ATTACHMENT A 

DOSIMETRY REPORTS

TABLE OF CONTENTS

SECTION TITLE PAGE

A.1 Reports Provided by Landauer............................................................................................... 95

A.2 Reports Provided by Eberline................................................................................................. 98

A.3 NUMEC Dosimetry Reports ................................................................................................. 100

A.3.1 Monthly Dose Calculation Sheets ............................................................................. 100

A.3.2 Monthly Special Report Calculation Sheet................................................................ 101

A.3.3 Monthly Summary Gamma and Neutron................................................................... 102

A.3.4 Monthly Monitoring Results by Wear Period ............................................................. 103

A.3.5 Annual Report with Monthly and Quarterly Summaries............................................. 104

A.3.6 Annual Report with Quarterly Summaries................................................................. 105

A.3.7 Annual Report by Month with Gamma and Neutron Doses....................................... 106

A.3.8 Annual Report with Monthly Results Later Years...................................................... 107

A.3.9 Worker Lifetime Summary ........................................................................................ 108

A.3.10 Cumulative Dose through November of 1971 ........................................................... 108

A.3.11 Plutonium Guard Summary 1971 – 1977.................................................................. 110

A.4 Guidance for Dose Assignment............................................................................................ 110

LIST OF FIGURES 

FIGURES TITLE PAGE

A-1 Landauer annual report for one worker (showing gamma, 1964)............................................ 96

A-2 Landauer monthly dose results (code 2314)........................................................................... 97

A-3 Landauer semi-monthly dose results (code 2315).................................................................. 98

A-4 Eberline report for one exchange period ................................................................................ 99

A-5 Monthly calculation for June 1972........................................................................................ 100

A-6 Special dosimetry calculation sheet ..................................................................................... 101

A-7 Monthly summary of gamma and neutron ............................................................................ 102

A-8 Monthly monitoring results by wear period ........................................................................... 103

A-9 Annual report with monthly and quarterly summaries ........................................................... 104

A-10 Annual report with quarterly summaries ............................................................................... 105

A-11 Annual report by month with gamma and neutron doses...................................................... 106

A-12 Annual report with monthly results later years...................................................................... 107

A-13 Worker lifetime summary...................................................................................................... 108

A-14 Cumulative dose through November 1971 ........................................................................... 109

A-15 Plutonium guard summary, 1971 to 1977............................................................................. 110
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ATTACHMENT A 

DOSIMETRY REPORTS (continued)

The external dosimetry records for NUMEC come in a variety of report types that change over the 

years. External dose monitoring has been provided by the NUMEC health physics department and 

the commercial suppliers Nuclear Science & Engineering (NS&E), R. S. Landauer Jr. and Company, 

and Eberline Instrument Corporation (Section 6.0). The following discussion describes the various 

reports found in NUMEC external dosimetry records. Although NS&E was reported to provide 

external dosimetry services sometime from 1959 to 1968, no NS&E reports have been found. 

A.1 REPORTS PROVIDED BY LANDAUER

The NUMEC records might contain summary reports for one monitoring period with dose values for 

several workers as well as annual reports for one worker. Annual reports for one individual were 

provided by Landauer in the early 1960s. Figure A-1 is an example of this report for 1964 by quarter 

of the gamma dose and the lifetime dose. Section 8 of the report indicates the dose is gamma dose. 

Section 9 of the report indicates the dose was measured by film badge. Each report provides only 

one type of dose: gamma, beta, or neutron. The lifetime total dose values can be useful in 

reconciling gaps in dosimetry information. The entries can be assumed to represent the badge 

exchange period, in this case quarterly.
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ATTACHMENT A 

DOSIMETRY REPORTS (continued)

Figure A-1. Landauer annual report for one worker (showing gamma, 1964).(NUMEC 

1964a, p. 3)
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ATTACHMENT A 

DOSIMETRY REPORTS (continued)

Landauer provided results of dosimeters for groups of workers for each badge exchange period. The 

form code 2314 indicates workers on a monthly badge exchange frequency and form code “2315” 

indicates workers on a half-month badge exchange frequency. Examples of these forms are given in 

Figures A-2 and A-3.

Figure A-2. Landauer monthly dose results (code 2314) (NUMEC 1964b, p. 72). 

Dose values are reported as Gamma and X-Ray, Beta or Thermal Neutron, Neutron, and Total for the 

period. Column 4 of the report indicates the type of exposure. A 1 indicates total body, 2 indicates 

skin exposure, and 3 and 4 indicate extremity. The Beta or Thermal Neutron column is beta dose if 

the exposure type is 2 (skin); it is thermal neutron if the exposure type is 1 (total body). If a worker 

was monitored for both total body and skin exposures, there would be two lines for the individual in 

the report. The report also indicates the cumulative total dose for the current quarter (or previous 13 

weeks), the current year, and the lifetime totals. Note that the lifetime totals represent the totals 

recorded and reported by Landauer and might not include all dose recorded by the site. However, it is 

possible the site provided additional dose values to Landauer that are included in the totals as 

indicated by a note in column 3. An additional feature of the report is the information on the far right 

side. This includes the month and year that monitoring was started (in the Landauer records) along 

with the number of badge periods from the start of monitoring until the present report. Missing badge 

reports are indicated. This information can be useful in reconciling dosimetry records and the number 

of badge cycles to include in the dose reconstruction.
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ATTACHMENT A 

DOSIMETRY REPORTS (continued)

Figure A-3. Landauer semi-monthly dose results (code 2315) (NUMEC 1964b, p. 86). 

A.2 REPORTS PROVIDED BY EBERLINE

One type of Eberline report was found in the dosimetry records. This report provides the dosimeter 

results for one reporting period for several workers. Figure A-4 shows results of Eberline film badges 

for workers at the Parks Township plutonium laboratory. The doses are reported for gamma, beta, 

and neutron exposures. For the gamma and beta columns, a blank indicates zero dose. For neutron 

exposures, a blank in the Neutron Tracks columns indicates no monitoring for neutrons. A value of 

zero, a positive value, or an entry of M (less than detection limit) indicates the worker was monitored 

with a neutron film badge and a value should be in the neutron dose column (or blank if the dose was 

zero). Two lines are included for each worker. The values depend on the section of the report. In the 

Dose for Period section, the top values give the gamma, beta, and neutron doses. The lower value in 

the first column is the sum of the gamma and neutron dose. When both the upper and lower values of 

the first column are the same, there is no neutron dose. In the next section the upper values are for 

Accumulated Dose for Cal Qtr and the lower values are for Accumulated Dose for Cal. Yr. This 

section gives WB dose (gamma plus neutron), skin dose, and extremity dose. A blank in the 

extremity column does not indicate a zero value because there might have been no monitoring for 

extremity doses. 

The lower line in the last section of the report gives the Lifetime Occupational Dose (rem). Other 

entries in this section are usually blank and provide no useful information. 
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ATTACHMENT A 

DOSIMETRY REPORTS (continued)

Figure A-4. Eberline report for one exchange period (NUMEC 1969, p. 4). 

The report lists entries for date of badge issue and badge return. The issue date is generally the first 

of the month. The return date is generally about 2 weeks after the end of the month in which the 

badge was issued. The wear period is likely for just the month in which the badge was issued

because the return date likely represents the time the badge was returned to Eberline for processing.

The quarterly, annual, and lifetime totals can be useful in reconciling gaps in dosimetry information. 

Although the report has a column for badge exchange frequency, this column seldom has an entry.
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ATTACHMENT A 

DOSIMETRY REPORTS (continued)

A.3 NUMEC DOSIMETRY REPORTS

NUMEC provided external dose monitoring for much of the covered periods for both the Apollo and 

Parks Township facilities. However, no reports are available from before 1964 when the records 

include the Landauer reports as described above. For some of the periods when Landauer and 

Eberline provided dosimetry, there could also be records of dosimetry from NUMEC. In these cases, 

the dose reconstructor must reconcile the two sets of dosimetry records. The records of dosimetry 

from NUMEC are discussed in this section.

A.3.1 Monthly Dose Calculation Sheets

This report indicates the calculations performed to convert dosimeter readout data to external dose 

from shallow beta/gamma, deep gamma, and neutron exposures. An example of a monthly 

calculation sheet is given in Figure A-5 for June of 1972. The headings indicate the results are for 

TLD badges. The dose values are reported to the tenth of an mrem. These values might be included 

in other summary reports but might be rounded to whole numbers of millirem. There are reports for 

the months of 1970 to 1975. Some of the dose entries indicate “see special report,” and no values 

are given for these individuals (see Figure A-6). There is generally one line for each worker with dose 

values on the right side of the page. Note that one worker has entries for three monitoring periods, 

with along with total dose for the month. When this is noted in the records, the worker was issued 

three dosimeters during the month and the total dose from the three dosimeters was reported. For 

such cases, assignment of zero doses should include consideration of all reported monitoring periods 

during the month.

Figure A-5. Monthly calculation for June 1972 (Boyd 2006b, p. 235). 
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ATTACHMENT A 

DOSIMETRY REPORTS (continued)

A.3.2 Monthly Special Report Calculation Sheet

This report shows calculation of external dose for workers with special dosimetry badges (Figure A-6.) 

The report is for one individual for the fourth quarter of 1974. The results for each badge wear period 

are given for shallow dose (SBG), deep gamma dose (DBG), and neutron dose (n). A summary of 

doses is listed at the bottom of the page.

Figure A-6. Special dosimetry calculation sheet (Boyd 2006b, p. 72). 
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ATTACHMENT A 

DOSIMETRY REPORTS (continued)

A.3.3 Monthly Summary Gamma and Neutron

This report gives the total monthly dose to individuals as handwritten results of WB dose from gamma 

and neutron (if any) exposures (Figure A-7). The Instrument location indicates TLDs were used to 

monitor external dose. Although the report has columns for each day of the month, only the monthly 

totals are given. The reported values include error bounds, which appear to be just 20% of the 

reported value.

Figure A-7. Monthly summary of gamma and neutron (Boyd 2006b, p. 426). 
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ATTACHMENT A 

DOSIMETRY REPORTS (continued)

A.3.4 Monthly Monitoring Results by Wear Period

Monthly reports with daily dose entries might be available for the late 1960s to the early 1970s giving 

doses reported as they were identified (Figure A-8). The report is for WB exposures from February 

1968. The entries include two numbers; the top number is the dose for the current period and the 

bottom number is the cumulative dose to the current day of the month. The final monthly total dose is 

listed on the left side of the form. This particular form includes a monthly total based on the TLD 

results and another total based on film badges. The Instrument: entry in the upper right corner of the 

form indicates the reported daily values are based on TLDs. The totals on the left for the TLD agree 

with the daily cumulative totals on the form. The upper left of the form indicates the dose values are 

for WB. Other reports might be available for finger ring doses. Care must be taken when reviewing 

the forms to be sure the doses are for the WB. Finger ring doses are only useful for evaluation of 

dose to cancers on the hands.

Figure A-8. Monthly monitoring results by wear period (NUMEC 1968a, p. 6). 
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ATTACHMENT A 

DOSIMETRY REPORTS (continued)

A.3.5 Annual Report with Monthly and Quarterly Summaries

This report gives an annual summary for workers by month, quarter, and year of WB external 

exposure (Figure A-9). The report gives the monthly dosimeter results for workers at the Parks 

Township ZPPR fuel production operation through the end of 1968. Other similar reports might be 

available that give results for only a part of the year. The workers on this form are grouped by job 

category. The first set is for Acc’t (material accounting) workers and the second set is for Ceramics 

workers. This information could be useful for identifying work locations. The dose values could 

include both gamma and neutron dose, but the breakdown is not indicated. The column Total Dose 

Thru 1968 gives the lifetime dose for each individual (one line per individual worker). The total 

quarterly, annual, and lifetime dose values can be useful in reconciling gaps in dosimetry information. 

Figure A-9. Annual report with monthly and quarterly summaries (NUMEC 1968b, p. 6). 
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ATTACHMENT A 

DOSIMETRY REPORTS (continued)

A.3.6 Annual Report with Quarterly Summaries

This report gives a summary of the WB doses by quarter for 1968 (Figure A-10). The report is for the 

Advanced Materials Center, which is the Parks Township site. The report has columns for WB dose, 

extremity dose, and skin dose. Blanks indicate the worker was not monitored for a particular dose. 

The total quarterly and annual dose values can be useful in reconciling gaps in dosimetry information. 

The report of quarterly values does not necessarily indicate the worker was on a quarterly badge 

exchange frequency. Each line of the report is for one worker. The report indicates that all workers 

were monitored for WB dose, but not all were monitored for extremity dose, and only the first was 

monitored for skin dose.

Figure A-10. Annual report with quarterly summaries (Caldwell 1969, p. 3). 
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ATTACHMENT A 

DOSIMETRY REPORTS (continued)

A.3.7 Annual Report by Month with Gamma and Neutron Doses

This report gives the annual summary of penetrating dose by month and quarter (Figure A-11). There 

are three entries for each individual and each period. The top value is the gamma WB dose, the 

middle value is the neutron dose, and the bottom value is the sum of the gamma and neutron doses. 

The report also gives the lifetime WB dose (gamma plus neutron) through the end of the reporting 

period. 

Figure A-11. Annual report by month with gamma and neutron doses (Boyd 2006b, p. 350). 
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ATTACHMENT A 

DOSIMETRY REPORTS (continued)

A.3.8 Annual Report with Monthly Results Later Years

This report gives the dosimetry results for several workers through part of a year (Figure A-12). The 

dose values include three entries per period and worker. The top value is the deep gamma dose, the 

second value is the neutron dose, and the bottom value is the sum of the gamma and neutron doses. 

The entries for the first two individuals indicate the workers were on a quarterly badge exchange 

frequency with entries given for March, June, and September. The next two individuals were on a 

monthly badge exchange frequency with entries for each month. The last individual was monitored 

starting in September. This possibly indicates a new employee, or the start of monitoring for a 

previously employed worker. This report is for Parks Township plutonium facility workers involved 

with “Pu Decontamination.” Because the period is in the residual period for the Parks Township 

facility, the dose values for this report would not be included in a dose reconstruction. However, if it is 

known the worker spent time at the Apollo facility, the values could be included in the assigned doses. 

The records could contain similar reports for the Apollo and Parks Township facilities for earlier years 

for which the dose values would be included. 

Figure A-12. Annual report with monthly results later years (BWXT 1982b, p. 33). 
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ATTACHMENT A 

DOSIMETRY REPORTS (continued)

A.3.9 Worker Lifetime Summary

The records could include a summary of the exposures received by an individual during all periods of 

employment (Figure A-13). This report includes summaries of external doses. External doses from 

gamma, beta, and neutrons are usually included. The report includes the employment period and the 

monitoring periods. The sample indicates an employment period that started and ended outside the 

monitoring period (Period of Exposure). This information can be useful in reconciling gaps in 

dosimetry records. The report also indicates the SNM license number under which the work was 

conducted. License SNM-145 indicates employment at the Apollo facility, and license SNM-414 

indicates employment at the Parks Township facility. Some employees are shown to have worked at 

both facilities. 

Figure A-13. Worker lifetime summary (Boyd 2006d, p. 9). 

A.3.10 Cumulative Dose through November of 1971

This report was prepared to show the lifetime doses to workers when the company was purchased by 

B&W (Figure A-14). The dose values are lifetime doses through November of 1971. Values are 

given for shallow dose (SBG), deep gamma dose (DBG), and neutron dose. The total WB dose is the 

sum of the deep gamma and neutron doses. Extremity exposures are also reported. The lifetime 

totals can be useful in reconciling gaps in dosimetry information. 
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ATTACHMENT A 

DOSIMETRY REPORTS (continued)

Figure A-14. Cumulative dose through November 1971 (Boyd 2006a, p. 3). 
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ATTACHMENT A 

DOSIMETRY REPORTS (continued)

A.3.11 Plutonium Guard Summary 1971 – 1977

A summary of doses for workers who served as plutonium guards at the Parks Township plutonium 

laboratory during 1971 through 1977 is included in the NUMEC records (Figure A-15). If dosimetry 

records are not included for a worker who was on the security force at the Parks Township plutonium 

laboratory during this period, then an analysis of the information in this report could be of use in 

estimating external dose.

Figure A-15. Plutonium guard summary, 1971 to 1977 (Boyd 2006a, p. 6). 

A.4 GUIDANCE FOR DOSE ASSIGNMENT

The assignment of dose depends on the information available. If the claim files have reported 

dosimeter readings for all dosimeter cycles, the assignment of dose is not a problem because the 

reported dosimeter values can be used as provided. The dose reconstructor still needs to review the 

dosimeter reports to ensure that all zero dosimeter readings are included because some of the reports 

have a blank rather than a reported value of zero. The dosimeter reports need to be carefully 

reviewed along with the employee’s work locations to determine if there was potential for neutron 

exposure and if neutron dose needs to be assigned. The NUMEC monthly calculation sheets 

(Figures A-5 and A-6), NUMEC monthly report sheets (Figure A-7), Landauer reports (Figures A-1, 
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ATTACHMENT A 

DOSIMETRY REPORTS (continued)

A-2, and A-3), Eberline reports (Figure A-4), annual summary reports (Figures A-11 and A-12), and 

lifetime reports (Figures A-13 and A-14) could give an indication if the employee was monitored for 

neutron exposure. 

The claims usually do not have data for all dosimeter assignments. For some claims, there might not 

be complete agreement among the dose values reported in NUMEC, Landauer, and Eberline reports. 

When the dosimeter information is incomplete, the dose reconstructor should attempt to reconcile the 

available information to ensure that the reported total dose (if available) has been assigned. For 

employees with both deep photon and neutron exposures, the relative amount of each dose type 

needs to be reviewed when the missing dose is assigned to periods with missing dosimeter readings. 

Most reported WB dose values for work at the Parks Township plutonium laboratory and source 

fabrication facility potentially include neutron dose.