ORAU TEAM

Dose Reconstruction

Project for NIOSH

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

Page 1 of 67

Document Title:

Site Profile for Nuclear Materials and Equipment 

Corporation, Apollo, Pennsylvania

Document Number: ORAUT-TKBS-0041

Revision: 00

Effective Date: 09/26/2008

Type of Document: TBD

Supersedes: None

Subject Expert(s): Paul J. Demopoulos and William Joyce

Site Expert(s): N/A

Approval: Signature on File Approval Date: 08/25/2008

Dennis L. Strenge, Document Owner

Concurrence: Signature on File Concurrence Date: 08/25/2008

John M. Byrne, Task 3 Manager

Concurrence: Signature on File Concurrence Date: 08/25/2008

Edward F. Maher, Task 5 Manager

Concurrence: Signature on File Concurrence Date: 08/25/2008

Kate Kimpan, Project Director

Approval: Signature on File Approval Date: 09/26/2008

James W. Neton, Associate Director for Science

 New Total Rewrite Revision Page Change

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

REVISION AND DISCARD / DESTROY ALL COPIES OF THE PRIOR REVISION.
Document No. ORAUT-TKBS-0041 Revision No. 00 Effective Date: 09/26/2008 Page 2 of 67

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.

All trademarks are the property of their respective owners.
Document No. ORAUT-TKBS-0041 Revision No. 00 Effective Date: 09/26/2008 Page 3 of 67

TABLE OF CONTENTS

SECTION TITLE

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

PAGE

7

1.0 Introduction ............................................................................................................................10

1.1 Purpose......................................................................................................................10

1.2 Scope.........................................................................................................................10

1.3 Special Exposure Cohort............................................................................................11

2.0 Site description ......................................................................................................................11

2.1 Apollo Site General Description and Operational History ............................................ 11

2.2 Apollo Facilities ..........................................................................................................12

2.3 Apollo Process Descriptions .......................................................................................12

2.3.1 General Process Operations........................................................................... 15

2.3.2 Various Uranium Process Activities................................................................. 15

2.3.3 High-Temperature Gas-Cooled Reactor Critical Assembly 

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

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

2.3.5 Apollo Incinerator............................................................................................ 18

2.3.6 Thorium Operations ........................................................................................ 19

2.3.7 Research Activities at Apollo NUMEC, the Early Years................................... 20

2.3.8 Source Term ................................................................................................... 20

2.4 Remediation, Decontamination and Decommissioning of NUMEC Apollo .................. 21

2.4.1 Shallow Land Disposal Area ........................................................................... 22

2.4.2 Apollo Administrative Building ......................................................................... 22

3.0 Medical X-Rays......................................................................................................................23

4.0 Environmental Occupational Dose ......................................................................................... 25

4.1 Environmental Internal Dose ......................................................................................25

4.2 Environmental External Dose .....................................................................................25

5.0 Occupational Internal Dose.................................................................................................... 25

5.1 Internal Exposure Sources at Apollo...........................................................................25

5.2 In vitro Bioassay .........................................................................................................27

5.2.1 Plutonium Urine and Feces Bioassay.............................................................. 27

5.2.1.1 Plutonium Urine MDCs and Frequencies.......................................... 28

5.2.1.2 Plutonium Urine Analytical Procedure............................................... 28

5.2.1.3 Plutonium Fecal MDCs and Frequencies.......................................... 29

5.2.2 Uranium Urine and Feces Bioassay................................................................ 30

5.2.2.1 Uranium Urine Analytical Procedure................................................. 30

5.2.2.2 Urine MDCs and Frequencies........................................................... 31

5.2.2.3 Uranium Fecal MDCs and Frequencies ............................................ 33

5.2.3 Thorium Exposures......................................................................................... 34

5.2.4 Mixed Fission Products................................................................................... 34

5.2.5 Unmonitored Radionuclides from Recycled Uranium ...................................... 35

5.3 In vivo counting ..........................................................................................................35

5.4 Apollo Process Uranium Air Sampling studies ............................................................37

5.4.1 Apollo 40 MPC-hr Reports to AEC of Overexposures ..................................... 39
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5.4.2 Other Incidents................................................................................................ 39

5.5 Apollo Uranium and Plutonium Air Sampling Programs .............................................. 40

6.0 Occupational External dose ................................................................................................... 40

6.1 External Exposure Sources at Apollo..........................................................................40

6.2 Radiological Exposure Sources..................................................................................41

6.2.1 Beta Radiation ................................................................................................ 41

6.2.2 Photon Radiation ............................................................................................ 41

6.2.3 Neutron Exposures ......................................................................................... 41

6.3 Personnel Monitoring..................................................................................................42

6.3.1 Beta/Gamma Radiation................................................................................... 42

6.3.2 Neutron Radiation........................................................................................... 42

6.3.3 Limits of Detection .......................................................................................... 42

6.3.4 Radiological Records ...................................................................................... 44

6.4 Limitations in Measured Dose.....................................................................................45

6.4.1 Low-Energy Photons....................................................................................... 45

6.4.2 Neutron Radiation........................................................................................... 48

6.5 Dose Reconstruction Recommendations....................................................................48

6.5.1 Recorded Dose Practices ............................................................................... 48

6.5.2 Adjustments to Recorded Dose....................................................................... 49

6.5.2.1 Beta Dose Adjustments .................................................................... 49

6.5.2.2 Photon Dose Adjustments ................................................................ 49

6.5.2.3 Neutron Dose Adjustments............................................................... 49

6.5.3 Missed and Unmonitored Dose ....................................................................... 50

6.5.3.1 Shallow Dose and Deep Dose.......................................................... 50

6.5.3.2 Neutron Dose ................................................................................... 50

6.5.4 Uncertainty...................................................................................................... 50

6.5.5 Radiation Dose Fraction.................................................................................. 51

7.0 Estimation of Exposure to Residual Activity ........................................................................... 51

7.1 External Dose from Residual Activity in the Workplace............................................... 51

7.2 External Ambient Dose from Residual Activity ............................................................53

7.3 External Medical X-ray Dose ......................................................................................53

7.4 Internal Dose from Residual Activity in the Workplace................................................53

8.0 Attributions and Annotations .................................................................................................. 54

References...........................................................................................................................................55

ATTACHMENT A, BREATHING-ZONE AIR SAMPLE SURVEY DATA............................................. 63
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LIST OF TABLES

TABLE TITLE

2-1 NUMEC Apollo site area and descriptions..............................................................................

PAGE

14

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

2-3 Decommissioning criteria for NUMEC Apollo ......................................................................... 21

3-1 Annual organ doses due to the assumed annual medical PA chest X-ray .............................. 24

5-1 Fuel types, chemical form, isotope, and enrichment of Apollo process material ..................... 25

5-2 Uranium source term information ........................................................................................... 26

5-3 Activity composition of Hanford reference fuel-grade plutonium mixture................................. 27

5-4 Plutonium and americium urine bioassay MDC, frequency, and period.................................. 28

5-5 Plutonium fecal bioassay MDC, frequency and period ........................................................... 29

5-6 Uranium urine bioassay MDC, frequency and period ............................................................. 32

5-7 Uranium fecal bioassay MDC, frequency and period.............................................................. 33

5-8 Lung-counting MDAs of uranium based on enrichment in picocuries ..................................... 36

5-9 In vivo MDAs for 239Pu and 241Am........................................................................................... 36

5-10 HASL report summary............................................................................................................ 37

6-1 MDLs and potential missed photon, beta, or neutron dose..................................................... 43

6-2 Dosimetry for external whole-body, wrist, and extremity exposures ....................................... 43

6-3 Annual occupational radiation exposures at the Apollo facility................................................ 44

6-4 Highest beta-gamma and neutron exposures measured for 1973 .......................................... 45

6-5 Plutonium photon energy factors............................................................................................ 47

6-6 NTA energy threshold correction factors ................................................................................ 48

6-7 Recorded dose practices........................................................................................................ 48

6-8 Interpretation of reported data................................................................................................ 48

6-9 Neutron dose correction factors ............................................................................................. 50

6-10 Beta, photon, and neutron radiation energies and percentages for IREP input....................... 51

7-1 External annual dose from residual surface contamination..................................................... 52

7-2 Uranium air concentration and annual intake in the residual period........................................ 54

A-1 HASL 103 weighted BZA survey, October 1960..................................................................... 63

A-2 HASL 92 weighted BZA survey, June 1960............................................................................ 64

A-3 HASL 82 weighted BZA survey, December 1959 ................................................................... 65

A-4 HASL 106 weighted BZA survey, December 1960 ................................................................. 66

A-5 HASL 114 weighted BZA survey, May 1961........................................................................... 67
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LIST OF FIGURES

FIGURE TITLE

2-1 Site layout, NUMEC Apollo ....................................................................................................

PAGE

13

2-2 Site floor plan at NUMEC Apollo ............................................................................................ 14

2-3 Schematic of UO2 production process.................................................................................... 16

2-4 B&W Shallow Land Disposal Area ......................................................................................... 23

5-1 Correlation of fecal bioassay with air sampling....................................................................... 30

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

ceramics line..........................................................................................................................46

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

glovebox ................................................................................................................................46

6-3 Dosimeter filter ratios ............................................................................................................. 46

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

fuel fabrication from MWD/T plutonium .................................................................................. 47
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ACRONYMS AND ABBREVIATIONS

ADU ammonium diuranate

AEC U.S. Atomic Energy Commission

AWE Atomic Weapons Employer

B&W Babcock & Wilcox (Company)

Bq becquerel

BZA breathing-zone air

CFR Code of Federal Regulations

cm centimeter

CP Chemical Processing

CRP Chemical Reprocessing

d day

DCF dose conversion factor

DOE U.S. Department of Energy

dpm disintegrations per minute

DU depleted uranium

EEOICPA Energy Employee Occupational Illness Compensation Program Act of 2000

EU enriched uranium

eV electron-volt

F fast (absorption type)

FP fission product

ft foot

g gram

GA general air

GSD geometric standard deviation

HASL Health and Safety Laboratory

HEPA high-efficiency particulate air

HEU highly enriched uranium

HTGR high-temperature gas-cooled reactor

hr hour

I insoluble

ICRP International Commission on Radiological Protection

IMBA Integrated Modules for Bioassay Analysis

in. inch

IREP Interactive RadioEpidemiological Program

keV kiloelectron-volt, 1,000 electron-volts

kg kilogram

L liter

lb pound

LEU low-enriched uranium
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M moderate (absorption type)

m meter

mCi millicurie

MDA minimum detectable activity (or amount)

MDC minimum detectable concentration

MDL minimum detectable level

MeV megaelectron-volt, 1 million electron-volts

mg milligram

min minute

ml milliliter

mo month

MOX mixed oxide

MPC maximum permissible concentration

mR milliroentgen

mrad millirad

mrem millirem

nCi nanocurie

NDC Nuclear Decontamination Corporation (Building; also Laundry Building)

NIOSH National Institute for Occupational Safety and Health

NRC U.S. Nuclear Regulatory Commission

NTA nuclear track emulsion, type A

NUMEC Nuclear Materials and Equipment Corporation

ORAU Oak Ridge Associated Universities

ORISE Oak Ridge Institute for Science and Education

PA posterior-anterior

pCi picocurie

POC probability of causation

QF quality factor

R&D research and development

RU recycled uranium

s second

S slow (absorption type)

SA specific activity

SEC Special Exposure Cohort

SLDA Shallow Land Disposal Area

SNM Special Nuclear Material

SRDB Ref ID Site Research Database Reference Identification (number)

TLD thermoluminescent dosimeter

TRU transuranic

U.S.C. United States Code

wk week
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wt % weight %

yr year

μCi microcurie

μg microgram
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1.0

1.1 PURPOSE

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 historic background information and guidance to assist in the preparation of 

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

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

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

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

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

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

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

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

contract period and/or during the residual period.

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

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

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

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

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

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

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

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

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

exposures to be occupationally derived (NIOSH 2007a):

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

conventional structures

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

reasons

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

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

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

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

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

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

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

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

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

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

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

1.2 SCOPE

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

Nuclear Materials and Equipment Corporation (NUMEC) Apollo plant. This site profile for NUMEC 

presents information useful for reconstruction of doses NUMEC employees received. 
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1.3 SPECIAL EXPOSURE COHORT

A Special Exposure Cohort (SEC) 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 workdays aggregating at least 250 workdays or in combination with workdays within the 

parameters established for one or more other classes of employees in the SEC. The SEC does not 

include work at the Parks Township site.

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

feasible to reconstruct doses for the following situations:

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

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

source term data;

• Potential ambient radiation dose from stack releases;

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

and 

• Internal doses where the bioassay data was based on the NUMEC Apollo contactor, Controls 

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

Although the SEC Petition Evaluation Report for petitions SEC-00047 and SEC-00080 (NIOSH 

2007c) 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 HHS 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 non-presumptive cancers may be considered 

for partial dose reconstruction (DHHS 2007).

The SEC covers employees of the NUMEC Apollo facility, which may include employees who worked 

at both the Apollo and Parks Township facilities. This site profile covers only the Apollo NUMEC 

facilities and may be used to perform partial dose reconstructions for individuals who worked at the 

Apollo facility.

2.0

2.1 APOLLO SITE GENERAL DESCRIPTION AND OPERATIONAL HISTORY

SITE DESCRIPTION

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

License number C-3762 that the U.S. Atomic Energy Commission (AEC) issued in 1957. From 1958 

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

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

was dedicated to production of uranium fuel. The major activities at NUMEC Apollo included: (1) the 

conversion of LEU hexafluoride (<5% 235U by weight) to uranium oxide (UF6 to UO2) for use in lightwater moderated water reactors; (2) the conversion of HEU to produce HEU (>20% 235U) nuclear fuel 

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

LEU and HEU) from the AEC in the 1960s (B&WNES 1997). 
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Between 1958 and 1983 the Apollo Site manufactured LEU uranium dioxide fuel for use in nuclear 

power plants. The process consisted of converting low-enriched UF6 to UO2. In 1963, a second 

product line was added to produce HEU (>20% 235U) uranium fuel for U.S. Navy propulsion reactors. 

Other operations included analytical laboratories, HEU and LEU scrap recovery, forming UO2 into 

pellets and other shapes, uranium storage, and research and development (R&D) (B&WNES 1997). 

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

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

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

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

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 site is approximately 33 miles northeast of Pittsburgh in the borough of Apollo in 

Armstrong County, Pennsylvania. The 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. These buildings were 

on the east side of the site between Warren Avenue and the Kiskimenetas River. They were 

bordered on the north, south, and west by a metals processing facility that was not owned by B&W. 

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

contained an analytical laboratory that was used to analyze radioactive and nonradioactive product. A

small portion of the building basement housed operations that manufactured instruments for the 

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

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

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

1997). The parking lot area was bounded by the Kiskimenetas River on the west, Warren Avenue on 

the east, and the offsite area on the north. The parking lot area contained the Laundry Building and a 

Small Block Building made of concrete blocks. 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. Figure 2-1 is a general layout of the Apollo site (Author unknown 2004). 

2.2 APOLLO FACILITIES

The Apollo site was broken down into production and process areas and clean areas. Personnel 

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

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

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

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

Production and process areas are described in Table 2-1. Some of these areas are illustrated in 

Figure 2-2.

2.3 APOLLO PROCESS DESCRIPTIONS

Brief summaries for the principal operations are given below and additional details can be found in the 

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

references (Forscher 1963; AEC 1960a,b,c, 1961a,b). Inherent in all of the operations is nuclear 

criticality safety that governs not only the operations and storage but also the movement of material 

within the facility. Nuclear criticality safety is maintained at the facility through the control of one or 

more of the conventional parameters of geometry, mass, concentration, and control of neutron 

interaction between sub-critical units. The standard administrative policy at this and other similar 
Document No. ORAUT-TKBS-0041 Revision No. 00 Effective Date: 09/26/2008 Page 13 of 67

facilities is to control two such parameters whenever possible. For birdcage units, the two control 

parameters are geometry (birdcage dimensions) and a 235U mass limit that can be placed within the 

Figure 2-1. Site layout, NUMEC Apollo.
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Figure 2-2. Site floor plan at NUMEC Apollo.

Table 2-1. NUMEC Apollo site area and descriptions.

Building or area 

location Description Operations/radionuclides Period of operation

CF-1 Ceramic fabrication UO2, ThO2, (metal, powder & oxide) 1957–1970

ThO2–1963-1970

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

U3O8

Early 1959 to 1972

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

UF6, UF4, U nitrate, UO2 and U3O8

1957–1983

HEU-1957–1978

LEU-1957–1983

PC-2 Process chemistry HEU, EU, DU, U nitrate, UO2, UF6, 

UF4, and U3O8

1957–1983

PC-3 Process chemistry HEU, EU, DU, U nitrate, UO2, UF6, 

UF4, and U3O8

1957–1983

CP-1 Chemical processing HEU, EU, DU, U nitrate, UO2, UF6, 

UF4, and U3O8

1957–1983

CP-2 Chemical processing HEU, EU, DU, U nitrate, UO2, UF6, 

UF4, and U3O8

1957–1983

CRP-1 Chemical reprocessing HEU, EU, DU, U 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 Handing Equipment, HEU, 

EU, DU, U nitrate, UO2, UF6, UF4, 

and U3O8

1957–1983

GF-1 1957–1983
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Building or area 

location Description Operations/radionuclides Period of operation

QC Quality control 1957–1983

PS 1957–1983

A Vault Process security material. 

Controlled by CP-2

HEU, EU, DU, U nitrate, UO2, UF6, 

UF4, and U3O8

1957–1983

C Vault Receiving and shipping 

storage area

1957–1983

E Vault Storage of SNM 1957–1983

F Vault Storage for SNM of all 

enrichments

1957–1983

G Vault Solution storage area, inprocess storage of materials 

for CP-1

HEU, EU, DU, U nitrate, UO2, UF6, 

UF4, and U3O8

1957–1983

H Vault Storage of SNM 1957–1983

Waste Treatment 

Area

Filter press section HEU, EU, DU, U nitrate, UO2, UF6, 

UF4, U3O8, FPs, and TRU elements

1957–1983

GPH Room 2 Health and safety counting 

room

HEU, EU, DU, U nitrate, UO2, UF6, 

UF4, U3O8, FPs, and TRU elements

1957–1983

Office Building Off-site Uranium 1957–1983

Laundry facility Co-60, Pu, Th, and U 1960–1983 (Closed 

1984)

Block Building Storage of processing equipment 1957–1983

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

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

birdcage. A review of the available literature shows that no criticality accidents have occurred during 

Apollo site operations.

2.3.1

NUMEC Apollo process operations were varied. Since NUMEC Apollo acted as a research center 

and an all-purpose process center a variety of radioactive materials and Special Nuclear Materials 

(SNM) were processed. Most of the work performed involved commercial work for commercial 

reactors but much of the work was in the development of better fuel configurations for the burgeoning 

Navy nuclear program. Even though this work is not considered weapons related work the SNM and 

radiation level exposures resulting from this work during the AEC operational years (1957-1983) are 

to be included in radiation dose reconstruction as per NIOSH policy, since this work cannot be 

distinguished from AWE related work. Only some examples are listed below to illustrate the breadth 

of the SNM operations that occurred at the NUMEC Apollo site.

General Process Operations 

2.3.2

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

Survey Reports 82, 92, 103, 106, and 114 (Occupational Exposure to Radioactive Dusts reports) 

which cover the period from December 1959 to January 1961 and the Procedure for Recovery of 

Scrap Uranium from about 1963. (AEC 1960a,b,c 1961a,b; NUMEC ca. 1963). The following 

paragraphs provide a description of the process. 

Various Uranium Process Activities 

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

Room. A 5-in. diameter UF6 cylinder was removed from a shipping/storage container known as a 

“birdcage” and placed in a heating mantle, connected to the system and brought to operating 

temperature and pressure. The UF6 gas was fed at a controlled rate into a water-circulating column 
Document No. ORAUT-TKBS-0041 Revision No. 00 Effective Date: 09/26/2008 Page 16 of 67

where the UF6 was hydrolyzed to an aqueous solution of UO2F2 and HF. This solution was then piped 

to a second column where it 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 

scraped off, placed into shallow metal containers called fry pans, and transferred to drying hoods 

where the ADU was decomposed at a controlled temperature to UO3. The UO3 was transferred from 

the pans to small polyethylene containers in an unvented glovebox and subsequently transferred to 

Ceramics Fabrication for further processing (AEC 1960a, p. 3). The HASL 92 (AEC 1960b, p. 2) 

survey from August 1960 noted that the filter cake was dried by means of a rotary kiln rather than the 

previous fry pan method. HASL 114 (AEC 1961b, p. 2) survey from June 1961 noted that a calciner 

was added for reduction of ADU to U3O8. The dried cake was then 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 may have been a manual operation. No information could be found to indicate that the 

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

Reduction of UO3 to UO2 was performed in the Ceramics Fabrication Area or reduction room where 

UO3 from the Chemical Conversion plant was loaded into a reduction tube and reduced to UO2 by 

passing wet hydrogen through the heated tube. UO2 was removed from the reduction tube, placed 

into a polyethylene container in a glovebox, weighed and transferred to the Ceramic Laboratory for 

additional fabrication. HASL 92 (AEC 1960b) noted that the UO3 was reduced to UO2 using a rotary 

kiln instead of the reduction furnace. 

A schematic illustrating the UO2 production process is presented in Figure 2-3. 

Figure 2-3. Schematic of UO2 production process (Franke and Makhijani 1998). 
Document No. ORAUT-TKBS-0041 Revision No. 00 Effective Date: 09/26/2008 Page 17 of 67

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

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

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

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

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

UO2 was then classified, weighed and packaged. 

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

were performed in the Ceramics Fabrication Area. 

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

weighed, then blended in a twin shell blender and prepressed by means of a Drake-press in a large 

polyethylene tent located in the ceramic laboratory. The prepressed slugs were subsequently fed into 

a granulator and granulated. The granulated powder was fed into the hopper of the Colton press and 

compact pressed. The resultant pellets were cured, inspected, and packed for shipping. All steps 

after prepressing were performed in the Ceramics Fabrication Area.

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

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

According to the 1960-1961 HASL survey reports, approximately 0.25-in. layer of UO2 powder was 

placed into a furnace tube (1-in. inner diameter by 2 ft) and heated to the desired temperature. A ratio 

of niobium pentachloride to hydrogen was then introduced. The pentachloride was reduced by the 

hydrogen so that niobium metal deposited on the UO2 particles. Vibration of the system allowed for 

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 was performed in the Chemical 

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

uranium (U3O8) from uranium-zirconium scrap as follows: 

One-gallon cans containing uranium-zirconium chips and oil were opened. The oil was drained and 

the chips transferred to wire mesh baskets. Chips were then degreased with trichloroethane in a 

ventilated degreasing tank. The chips were then transferred to 6-L beakers and dissolved in 

hydrofluoric acid in chemical fume hoods. The zirconium remains in solution and part of the uranium 

precipitates as UF4. After complete dissolution of chips, the batch was heated and hydrogen peroxide 

added to oxidize the insoluble UF4 to soluble UF6. The batch was then filtered to remove any 

remaining insoluble material. The filtrate was then 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 the HASL 92 (AEC 1960b) survey report, the experimental development of recovering 

U3O8 (93%) by solvent extraction was under development at Apollo (AEC 1960b). In addition, UF4

was being converted to U3O8 by ignition with the U3O8 subsequently 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 
Document No. ORAUT-TKBS-0041 Revision No. 00 Effective Date: 09/26/2008 Page 18 of 67

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

packaging as UO3. 

According to the Procedure for Recovery of Scrap Uranium, undated, NUMEC Apollo, Pennsylvania

(submitted to AEC around 1962 or 1963) (NUMEC ca. 1963), incoming scrap was received in 

criticality safe "bird cages" and stored at Advanced Materials Center Parks Township near Leechburg, 

Pennsylvania, 5 miles north of Apollo site on route 66. Specific lots were then moved to the Apollo 

site on NUMEC trucks and logged into the Apollo process storage area located on the second floor of 

the Apollo site. One drum at a time was then carried through the checking process and wheeled to 

the head end of the processing area. An accountability check of the information supplied by the 

shipper was then performed including a gross piece count. Net weight checks were performed on 

chips, fuel elements and fuel assemblies and recorded on Form CRP-2-A. Pickle liquors and other 

solutions containing over 5 g/L were received in Eversafe 5-in. containers (dissolved scrap procedure 

followed for liquids). Pickle liquors and other solutions containing less than 5 g/L were received in 55-

gallon drums (dissolved scrap procedure followed for liquids). Degreasing of material was performed

as necessary. Fines were removed from oil as necessary (NUMEC ca. 1963). 

NUMEC dissolved the scrap in two designated areas: CRP-2 and CRP-3. Dissolution was performed 

in dissolving hoods in 5-in.-diameter, stainless-steel beakers. The total quantity in each hood was 

limited to 350 g of 235U. Uranium-zirconium chips and pieces were mixed with hydrofluoric acid. U-Al 

alloy pieces were mixed with nitric acid. UO2 and U3O8 were mixed with nitric acid. BeO-UO2 scrap 

was mixed with HF-HNO3. The resulting solutions were placed into a mixing column and the 

container was then sampled, assayed, and weighed (NUMEC ca. 1963). The product solution from 

the various dissolution methods was processed to generate insoluble UF4, and ultimately converted 

to the final product of U3O8 or UO2.

2.3.3

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

fabricate 3000 graphite fuel elements to be used in the General Atomics High-Temperature GasCooled Reactor (HTGR) critical assembly. The job was to involve between 95 to 120 kg of 93%-

enriched U3O8 supplied by another company. Fuel element composition and specifications were 

provided in the letter. The total 235U content of 2,850 fuel elements was to be 79.339 kg. There is 

some process description in the letter but not much detail. A letter dated April 1960 to Shapiro 

(NUMEC) from Wesley Johnson, Deputy Manager (AEC) indicated approval of Feasibility Report for 

the General Atomic HTGR critical assembly fuel elements (Katine 1960). 

High-Temperature Gas-Cooled Reactor Critical Assembly Fuel Elements 

2.3.4

A letter report dated June 9, 1961, written by J. E. McLaughlin, Director, Radiation Physics Division, 

HASL, describes a trip report to the NUMEC Apollo Facility on 6-7-61 to observe equipment for 

producing uranyl nitrate solution for the University of California as described in Feasibility Report No. 

27, dated 6-9-61. A vague description of NUMEC processes and facilities was reported. The report 

mentions nitric acid and aluminum nitrate solutions employed in the solvent extraction process (Katine 

1961). 

Uranium Nitrate Solution for the University of California 

2.3.5

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

consisted of a Hoskinson H-100 incinerator equipped 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). 

Apollo Incinerator
Document No. ORAUT-TKBS-0041 Revision No. 00 Effective Date: 09/26/2008 Page 19 of 67

Packages of contaminated waste awaiting incineration were stored in designated areas. Each 

package was labeled with the 235U content. Packages were burned at a rate of 30-35 lb/hr. Ashes 

were collected in stainless steel, 1-gallon containers. The ash receiver can was changed when a total 

of 300 g of 235U had been charged or when it was full. 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 gallon pail. The pail was labeled and placed on a 

temporary storage rack pending laboratory analysis. On receipt of analytical data, the pail label was 

completed with the uranium (

235U) content and 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 gloved box at 

negative pressure (Caldwell 1968a). (As of 1967 this was likely not the case and work in this area 

was perhaps the highest area for intakes. See section 3.1.2.) 

Combustible gases passed through the after-burner to a water-operated venturi-type fume scrubber. 

This separated the fly ash from the gas stream. Downstream of the venturi, the gases passed 

through a packed tower where fine particulates were scrubbed from the gas stream by a countercurrent flow of water. These gases were then discharged through a 15-ft stack. A make-up air duct 

was used to maintain negative head on the scrubber and for controlling the flow of flue gases through 

the system. Exhaust air from the ash handling glovebox passed through a prefilter and final highefficiency particulate air (HEPA) filter before discharge through a roof stack (Caldwell 1968a). 

2.3.6

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

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

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

Thorium Operations

NUMEC correspondence with the AEC, Oak Ridge Operations Office, indicates the following NUMEC 

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 having a face velocity of 100 fpm. Material would be handled wearing 

latex gloves. 

3. Slug the powder to 4-5 g/cc then granulate through 14-mesh screen. 

4. Blend each batch of powder in a "V" type blender located in a ventilated glovebox. 

5. Press the ThO2 pellets using a hand press and/or automatic press within a hood with face 

velocity of 100 fpm. 

6. Sinter the ThO2 pellets in hydrogen atmosphere with the out gases of the furnace passing 

through a filtered exhaust ventilation system. 

7. All pellets will be center-less ground in ventilated hood. 

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

polyethylene bag. 
Document No. ORAUT-TKBS-0041 Revision No. 00 Effective Date: 09/26/2008 Page 20 of 67

Air sampling was performed by the site to characterize thorium exposures during this period. 

According to health protection program review conducted in 1964, thorium operations involving the 

blender and weighing hood were resulting in excessive airborne concentrations (AEC 1964). 

2.3.7

There were various research projects conducted at the Apollo site that involved mostly the fabrication 

of new types of fuel in support of the Naval Reactors Branch through the Knolls Atomic Power Plant 

and Bettis reactor research laboratory. The research involved chemical process development with 

various forms of uranium compounds and metal.

Research Activities at Apollo NUMEC, the Early Years

2.3.8

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

handled at the Apollo 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 HASL reports, and process feasibility reports, which contain information regarding 

radionuclides, quantities and recommended safety precautions for the described activity; and (3) 

inventory/material handling (accountability) records.

Source Term

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

Source Material License number C-3762 issued by AEC in 1957 (Docket No 70-135). Some 

possession limits at different periods of time are listed for the Apollo site in Table 2-2.

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

activation products (Reitler 1972). Currently 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 any given area or process. Much of the work was R&D work so 

unique source terms could be encountered in a particular job. 

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

reactor fuel, research studies in the gloveboxes and laboratories (AEC 1960a –c and AEC 1961a-b). 

The typical amounts of uranium in use in any one area ranged from milligrams to hundreds of 

kilograms. Occasional work involving other uranium chemical forms, such as UF6 or uranyl nitrate 

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

Areas Source/chemical or physical form

Maximum 

possession

Processing areas, laboratories 

and vaults

U-235 enrichment >5% 5,000 kg

U-235 enrichment >= 5% 75,000 kg

Plutonium as fully clad or encapsulated material 500 kg

Mass Spectrometry Laboratory Uranium in any enrichment 350 g

Plutonium in any form 0.5 g

Low-level radioactive waste 

storage areas

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

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

Nuclear Decontamination 

Corporation 

Any by-product material 20 mCi

Any source material 20 g

Any SNM 20 mCi

Source: SNM-145 Renewal Application October 31, 1972, Reitler (1972).
Document No. ORAUT-TKBS-0041 Revision No. 00 Effective Date: 09/26/2008 Page 21 of 67

was also conducted. Uranium isotopic forms included DU, normal (i.e., natural enrichment), HEU (up 

to 93%), as well as 232U, 233U and 236U. 

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

less than that of uranium overall, but the thorium activity in use in an area at any 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). 

Uranium from recycle operations would have included smaller 

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

2.4 REMEDIATION, DECONTAMINATION AND DECOMMISSIONING OF NUMEC APOLLO

The HEU processing area located 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 (B&WNES 1997). 

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

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

all of the equipment was removed and sent to Chem-Nuclear (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’s trench that served as a sump drain for washing machine 

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

All of the 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 processing plant (B&WNES 1997). 

Soil acceptable decommissioning criteria are described in the “Apollo Final Technical Report,” but 

residual period building contamination levels were not covered (B&WNES 1997). The 

decommissioning criteria that were followed for the Apollo site are presented in Table 2-3. 

The background exposure rates were found to be 9.5 μR/hr and the average concentration in the soil 

4 pCi U/g. 

Table 2-3. Decommissioning criteria for NUMEC Apollo.a

Radionuclide

Release concentration

pCi/g

Total TRU 25

Sr-90 5

Tc-99 220

Co-60 8

Am-241 30

Th-232 10

a. B&WNES (1997).

Another decommissioning study conducted by Oak Ridge Institute for Science and Education

(ORISE) for the NRC in 1993 indicates that the main building ground area exposure rates were 

between 9 to 12 μR/hr and between 10 to 13 μR/hr in the south parking lot. Total Uranium 

concentrations ranged from 6.5 to 2,200 pCi/g for single point samples and 5.3 to 280 pCi/g for the 

grid block soil samples (Adams 1993).
Document No. ORAUT-TKBS-0041 Revision No. 00 Effective Date: 09/26/2008 Page 22 of 67

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

process and related equipment were removed by this point in time. NUMEC indicated that access to 

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

to identify the remaining HEU that may 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 for Apollo remaining on and in the floors, 

walls, pad and ceiling was estimated as 35,548.55 g of total uranium with about 23,743.27 g of 235U 

(Martin 1982). 

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

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

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

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

Environmental Protection, NRC issued a letter to B&W terminating the Apollo license (PDEP 2008). 

2.4.1

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 Parks) with about 700,000 ft3 of waste buried in trenches on the site (PDEP 2008). Currently 

options for the disposition of the SLDA are in discussion with the NRC and the former licensee. The 

site may become a Formerly Utilized Sites Remedial Action Program site. As of 2007 a feasibility 

study is being conducted by the Army Corps of Engineers (USACE 2007). A graphic of the area is 

provided in Figure 2-4. 

Shallow Land Disposal Area

2.4.2

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

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

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

Apollo Administrative Building

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

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

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

was backfilled and leveled (survey conducted May 23, 1995). All results were found to meet the 

guidelines previously established for unrestricted release. Special emphasis was made to survey 

closely the northeast area of the administrative building where the wet laboratory existed and the 
Document No. ORAUT-TKBS-0041 Revision No. 00 Effective Date: 09/26/2008 Page 23 of 67

Figure 2-4. B&W Shallow Land Disposal Area (SLDA) (USAEC

2007).

southern end, the location of the instrument repair and calibration laboratory (NRC 1995). Exposure 

rate measurements, including background, ranged from 10 to 12.5 μR/hr, with an average of 

approximately 11 μR/hr. The guideline was 5 μR/hr above background meaning that the acceptable 

level was approximately 15 μR/hr (NRC 1995). The reported values essentially represent background 

levels. 

Soil samples were analyzed by the Oak Ridge Institute for Science and Education (ORISE). Their 

results indicated that no 235U was found above the detection limit (not stated). Total uranium was not 

found statistically above the counting uncertainties and measured less than 3.5 pCi U/g of soil. Total 

thorium concentrations were found to range from 1.51 to 2.02 pCi/g with an average of 1.86 pCi/g. 

The guidelines for unrestricted release of soil with residual concentrations of EU and thorium are 30 

pCi U/g and 10 pCi Th/g of soil, respectively. All soil concentrations were found to be well within 

these guidelines. The background concentrations were NOT subtracted.

3.0

From a cursory review of claimant medical records the X-ray section of the physical forms seems not 

to be filled out for most of the claimant files. NUMEC apparently did not have its own medical X-ray 

department during AEC operational years. However, some claimant case files indicate some 

posterior-anterior (PA) chest X-rays were given. The X-ray was read off-site by a radiologist, and the 

radiologist’s report could be dated up to 5 months after the date of the physical examination. This 

could have been the reason why the X-ray documentation was not retained in the medical records. 

There is no evidence that Lateral chest X-rays or lumbar spine X-rays were given to NUMEC 

employees routinely or for screening purposes as evidenced from the review of the claimant records. 

It was indicated at least as of 1963 that annual physical examinations were given as well as preemployment and termination exams (NUMEC 1963). Even though there is no current evidence that 

routine medical X-rays were performed at NUMEC, it is reasonable to assume that one X-ray should 

be assigned for preemployment, one at termination and annually in the interim. A PA chest X-ray was 

MEDICAL X-RAYS
Document No. ORAUT-TKBS-0041 Revision No. 00 Effective Date: 09/26/2008 Page 24 of 67

given to employees who worked with beryllium, twice a year as indicated in the NUMEC (Apollo) 

health and safety manual (NUMEC1963).

To date no site-specific information is available for a NUMEC Apollo medical X-ray program. 

Therefore, medical X-ray doses should be assigned based on general guidance provided in the 

Technical Information Bulletin ORAUT-OTIB-0006 (ORAUT 2005a). All X-rays should be assumed to 

be PA chest X-rays. Photofluorography was not likely because this method was suitable for screening 

large groups of people at one time. Because the medical X-rays for NUMEC employees appear to 

have been performed at a local clinic or hospital, the use of photofluorography was unlikely and 

should not be assumed to have occurred unless specifically indicated in the medical records. The 

organ dose from PA chest X-rays should be based on the values provided in Table 6-5 of ORAUTOTIB-0006 (ORAUT 2005a), which are reproduced here in Table 3-1. For organs not in the table, 

surrogate organs may be used as indicated in Table 3-2 of ORAUT-OTIB-0006 (ORAUT 2005a). 

Table 3-1. Annual organ doses due to the assumed annual medical PA chest X-ray.

Organ

Annual dose (rem)

1957–1969

Annual dose (rem)

1970–1985

Annual dose (rem)

1985–present

Bladder 2.50E-02 1.00E-04 2.60E-04

Red bone marrow 1.84E-02 male

1.72E-02 female

9.20E-03 8.90E-03

Bone surface 9.02E-02 4.51E-02 3.37E-02

Breast 9.80E-03 4.90E-03 5.80E-03

Colon/rectum 2.50E-02 1.00E-04 2.60E-04

Esophagus 9.02E-02 4.51E-02 3.37E-02

Eye/brain 6.40E-03 3.20E-03 3.90E-03

Ovaries 2.50E-02 1.00E-04 2.60E-04

Testes 5.00E-03 1.00E-06 5.00E-07

Liver/gall bladder/spleen 9.02E-02 4.51E-02 3.37E-02

Lung 8.38E-02 male

9.02E-02 female

4.51E-02 3.37E-02

Remainder organs 9.02E-02 4.51E-02 3.37E-02

Skin 2.70E-01 1.35E-01 7.00E-02

Stomach 9.02E-02 4.51E-02 3.37E-02

Thymus 9.02E-02 4.51E-02 3.37E-02

Thyroid 3.48E-02 3.20E-03 3.90E-03

Uterus 2.50E-02 1.30E-04 2.60E-04

a. Based on ORAUT (2005a).

The frequency of medical X-rays should be based on the frequency of routine physical examinations 

as provided in the employee’s records. One PA chest X-ray should be assumed to have been 

provided along with each routine physical, even though, as stated above, the routine physical records 

will likely not indicate that the employee received an X-ray. If the frequency of routine physical exams 

is not provided in the employee records, the one PA chest X-ray should be assumed to have been 

given during a prehire physical, an annual physical, and at termination. If there is indication that the 

employee worked regularly with beryllium, then two PA chest X-rays should be assigned each year.

The PA chest X-ray values in Table 6-5 of ORAUT-OTIB-0006 can be assigned as an overestimating 

approach by multiplying the values by an uncertainty factor of 1.3, represented as a constant 

distribution. For assignment of a best estimate, the values can be entered as the mean of a normal 

distribution with a standard deviation of 30% (ORAUT 2005a). All medical X-ray doses should be 

assigned as an acute exposure to photons with energy from 30 to 250 keV. 
Document No. ORAUT-TKBS-0041 Revision No. 00 Effective Date: 09/26/2008 Page 25 of 67

4.0

The SEC Petition Evaluation Report for petitions SEC-00047 and SEC-00080 (NIOSH 2007c) 

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

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. 

ENVIRONMENTAL OCCUPATIONAL DOSE

4.1 ENVIRONMENTAL INTERNAL DOSE

Adequate information on environmental air concentrations near the NUMEC Apollo site was not 

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

4.2 ENVIRONMENTAL EXTERNAL DOSE 

Information on ambient external dose levels at the NUMEC Apollo site were not found. Therefore, no 

estimates of external ambient dose can be made for workers for any time period at the NUMEC Apollo 

site. 

5.0

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

while performing tasks within buildings and structures at the NUMEC Apollo site 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 technical data to evaluate internal 

occupational doses that can reasonably be associated with worker radiation exposures covered by 

the EEOICPA legislation. The health and safety coverage for NUMEC Apollo and Parks Township 

were administered under one department. The information in this section covers exposure at both 

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

OCCUPATIONAL INTERNAL DOSE

5.1 INTERNAL EXPOSURE SOURCES AT APOLLO

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

some potential for exposure to plutonium, or thorium dust produced from the manipulation and 

chemical processing of those materials during recovery and fuel fabrication processes. Uranium 

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

and activation products was possible at some locations (laundry facility). 

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

Apollo facility.

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

a

Radionuclide or fuel

Chemical form and 

solubility type(s)b

Isotope 

(% in mass, where listed) Enrichment

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

UO3 & UF4 (M)

U3O8 & UO2, (S)

U-234

U-235

U-238

DU, Natural, 

LEU (3.5%)

Thoriumc ThO2 (M, S) Th-228,

Th-232

Natural 

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 yr 
Document No. ORAUT-TKBS-0041 Revision No. 00 Effective Date: 09/26/2008 Page 26 of 67

Radionuclide or fuel

Chemical form and 

solubility type(s)b

Isotope 

(% in mass, where listed) Enrichment

Technetium or other 

TRU elements

Same as the Th, U or Pu 

matrix

Tc-99, Np-237 N/A

MOXe PuO2 (M, S, SS)/UO2 (M, S) 20% PuO2 and 80% UO2

[7% Pu – fuel grade/5% Pu –

weapons grade]

About 4.5% 235U

Fission and Activation 

Products

Unknown Co-60, Sr-90, Ru/Rh-106, Cs137, Tc-99 (from recycled U)N/A

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

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

c. All thorium work was with unirradiated thorium material.

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

e. Mixed Oxide (MOX) work was likely limited to the Parks Township facility.

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 form and the enrichment varied over time at the Apollo 

facilities. The manufacture of uranium products occurred in most of the buildings in Apollo. See 

Table 2-1 for further information. The dose reconstructor should use the solubility type that results in 

the highest dose.

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

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

enrichment.

Table 5-2. Uranium source term information.

Uranium source 

term Reference

Specific activity

pCi/μg

Activity fractions

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

Natural uranium IMBAa 0.683 0.489 0.023 - 0.489

93.% IMBAa 68.1 0.968 0.030 0.002 0.0003

3.5% IMBAa 2.20 0.818 0.034 - 0.147

2% HPSb 1.20 0.648 0.041 0.0009 0.311

Typical DU IMBAa 0.402 0.155 0.011 0.0005 0.834

Uranium source 

term Reference

Specific activity

pCi/μg

Specific constituent activity in mixture 

(μCi/g, nCi/mg, or pCi/μg)

Natural uranium IMBAa 0.683 0.334 0.016 - 0.334

93.% IMBAa 68.1 65.9 2.04 0.136 0.020

3.5% IMBAa 2.20 1.80 0.075 - 0.323

2% HPSb 1.20 0.778 0.049 0.001 0.373

Typical DU IMBAa 0.402 0.062 0.004 0.0002 0.335

a. IMBA = Integrated Modules for Bioassay Analysis software.

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

c. ORAUT (2007a).

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

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

amount processed of each chemical form is not known.

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

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

have a tendency to oxidize when left in contact with air. Oxides, metals, and old contamination should

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

S can be used to maximize the dose to the organ of concern. Also, because highly insoluble forms of 

plutonium may have been present, guidance in Technical Information Bulletin: Estimating Doses for 

Plutonium Strongly Retained in the Lung (ORAUT 2007b) should be followed for the evaluation of 
Document No. ORAUT-TKBS-0041 Revision No. 00 Effective Date: 09/26/2008 Page 27 of 67

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

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

words, the americium should be treated as absorption type S if the plutonium is type S (ORAUT 

2007a). If the plutonium is type Super S, follow guidance in ORAUT (2007b) for assignment of the 

241Am solubility type.

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

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

information on the actual composition of the processed plutonium, an assumption of 10-yr-old 

fuel-grade plutonium is favorable to the claimant and reasonable (Table 5-3). This is noted in the 

Hanford site profile; Hanford processed much of the DOE complex plutonium (ORAUT 2007a). 

Table 5-3. Activity composition of Hanford reference

fuel-grade plutonium mixture (12%).

a

Mixture designation SA in (Ci/g)

Alpha activity 

fraction 

Years of agingb 10-yr

Pu-238 1.58E-02 1.17E-1

Pu-239 5.26E-02 3.91E-01

Pu-240 2.72E-02 2.02E01

Pu-241 1.91E+00 0.0

Pu-242 3.93E-06 2.92E-05

Am-241 3.89E-02 2.89E-01

Total alpha 1.35E-01 1.00

a. Source: ORAUT (2007a).

b. Time since separation of the Am-241 from the plutonium mix.

5.2 IN VITRO BIOASSAY

The bioassay program for NUMEC workers primarily involved urine and fecal sampling for isotopes of 

uranium, plutonium, and americium-241. Occasional analyses were performed for FPs and 232Th. 

The reported bioassay data generally includes 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 evaluation for the NUMEC Apollo site (NIOSH 2007c) indicated that Controls for Environmental 

Pollution has been implicated in the falsification of data and that its bioassay analyses provided to 

NUMEC cannot be considered reliable. Bioassay data from Controls for Environmental Pollution 

should only be used to indicate the potential for exposure to a particular radionuclide on a particular 

date. The data can not 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, except as noted below for urine 

bioassay data from Controls for Radiation in the early years (1961 – 1965).

5.2.1

Plutonium may have been present at the NUMEC facilities in several forms that include type M, type 

S, and possibly type Super S material solubility categories. The intake analysis based on bioassay 

monitoring results should evaluate intakes based on all three types and use the type providing the 

highest dose estimate. 

Plutonium Urine and Feces Bioassay
Document No. ORAUT-TKBS-0041 Revision No. 00 Effective Date: 09/26/2008 Page 28 of 67

5.2.1.1 Plutonium Urine MDCs and Frequencies

Plutonium-239 was analyzed in urine from about 1962 to 1999, and 241Am was analyzed starting in 

about 1970. The MDCs are listed in Table 5-4 for NUMEC facilities. If an MDC value is needed prior 

to the dates listed in the table, the values for the earliest date should be used. Note that no bioassay 

monitoring results were found between 1985 and 1999. Also, because one health physics department 

was responsible for the bioassay program at both the Apollo and Parks Township sites, it is difficult to 

determine from the reported bioassay results if the employee worked at the Apollo or Parks Township 

sites. It is likely that much of the plutonium bioassay results were for work at the Parks Township site. 

The accuracy of the early plutonium measurements was questioned by a manager who observed that 

some workers not associated with the plutonium fabrication work had higher plutonium urine results 

than the plutonium workers (Puechl 1963). The information indicated that the minimum detectable 

concentration (MDC) might have been closer to 5 dpm/L than to the reported values in measurement 

results. Therefore, for the first period in Table 5-4 the MDC has been assigned as 5 dpm/L. This 

value should be used as a minimum MDC over the listed values in the bioassay records.

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

a,b

Date Laboratory Radionuclide Frequencyb MDCc,d Errore

10/1961–12/1965 Controls for Radiation Pu Quarterly/as needed 5 dpm/L 0.02–0.12 dpm/L

9/1968–12/1975 Eberline Pu-239 Quarterly/as needed 0.06 dpm/sample 0.03 dpm/sample

1/1970–12/75 Eberline Am-241 Quarterly/as needed 0.06 dpm/sample 0.03 dpm/sample

1/1976–4/1980 Controls for Environmental 

Pollution

Pu-238 As needed f f

1/1976–4/1980 Controls for Environmental 

Pollution

Pu-239 Quarterly/as needed f f

1/1976–4/1980 Controls for Environmental 

Pollution

Am-241 Quarterly/as needed f f

5/1980–9/1985 Controls for Environmental 

Pollution

Pu-238 As needed f f

5/1980–9/1985 Controls for Environmental 

Pollution

Pu-239 Quarterly/as needed f f

5/1980–9/1985 Controls for Environmental 

Pollution

Am-241 Quarterly/as needed f f

1999 Quanterra Pu-238 Unknown 0.0025–0.044 pCi/L -

1999 Quanterra Pu-239/240 Unknown 0.0025–0.045 pCi/L -

11/1974–8/1975 Eberline Gross alpha (Pu + Am) Quarterly/as needed <10.0 dpm/sample -

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

b. Records indicate quarterly monitoring for plutonium workers, unless an intake was suspected initiating more frequent special sample 

analyses.

c. Assumes the MDC is twice the reported error

d. The 5 dpm/L value is based on the uncertainty issue mentioned in the text. 

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 is not to be used in internal dose assessments.

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 (Barry 1964), the early analytical procedure likely consisted of drying 500 ml of urine 

to dryness with HNO3. The residue was re-evaporated successively with nitric acid (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 

any organic residue. The planchet was counted in a Nuclear Measurement Corporation gas flow 

proportional counter for four hours. The minimum sample volume was 500 ml. Because nearly 
Document No. ORAUT-TKBS-0041 Revision No. 00 Effective Date: 09/26/2008 Page 29 of 67

weightless samples are 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 (Barry 1964).

Procedures used to analyze for plutonium in urine in later years are not known. 

From a review of the worker dosimetry records, once per quarter seemed to be the average 

frequency. Special bioassays were ordered for those workers exceeding 40 maximum permissible 

concentration-hours (MPC-hr) of exposure or nose wipes exceeding 25 dpm.

5.2.1.3 Plutonium Fecal MDCs and Frequencies 

The analytical procedure for plutonium fecal analysis has not been located. The estimated MDCs are 

listed in Table 5-5 for NUMEC facilities.

Fecal sampling was initiated in January 1966 at the NUMEC facilities. Three goals of the program 

were (1) the early detection of acute inhalation exposures, (2) estimation of detected lung burdens,

and (3) the screening for potential chronic exposures (Caldwell 1966a). The fecal analysis continued 

until about 1985 as indicated in worker dosimetry records. The fecal analysis results reported as 

dpm/sample should be considered equivalent to the daily excretion rate (dpm/day).

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

a,b

Date Laboratory Radionuclide Frequencyb MDCc Error

1/1966-1/1977 Eberline Pu-239, Pu-238 or 

Am-241

Quarterly/as 

needed

0.1 dpm/sample 0.05 dpm/sample

5/1975 – 9/1975 Eberline Gross alpha (Pu + 

Am)

As needed 0.1 dpm/sample 0.05 dpm/sample

2/1977-10/1985 Controls for Environmental 

Pollution

Pu-239, Pu-238 or 

Am-241

Quarterly/as 

needed

d d

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

b. Records indicate quarterly monitoring for plutonium workers, unless an intake was suspected initiating more frequent special sample 

analyses.

c. Assumes the MDA is twice the sensitivity or error. 

d. Bioassay data analyzed by Controls for Environmental Pollution is not to be used in internal dose assessments.

NUMEC health physicist Roger Caldwell believed that fecal sampling was the only satisfactory 

method for estimating lung burdens for insoluble actinide alpha emitters classified as Y in the 

contemporary lung model. The most important alpha emitters included 239PuO2, 241AmO2, 234UO2, and 

232ThO2 (Caldwell 1966a). 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 1966a). 

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 1966a; 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 (HPS 1966) and lapel sampler data 

(Caldwell, Potter, and Schnell 1967). 

NUMEC health physicists used BZA and general air (GA) sample results to screen for possible 

exposures. If an exposure occurred (based on BZA sample or incident) bioassays of usually both 
Document No. ORAUT-TKBS-0041 Revision No. 00 Effective Date: 09/26/2008 Page 30 of 67

fecal and urine samples were collected and then correlated with BZA samples. The suspected 

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

This was the method employed by the mid 1960s since 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). 

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

(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 

in order 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 

1966a). A number of different bioassay vendors performed plutonium urine and fecal analysis as 

observed from worker dosimetry records. 

5.2.2

Uranium was processed at the Apollo site. Enrichment levels varied with time and included depleted 

uranium (DU), natural uranium, LEU (3.5%), and HEU (93%).

Uranium Urine and Feces Bioassay

5.2.2.1 Uranium Urine Analytical Procedure

Information in HASL report 82 (AEC 1960a) indicates that before 1960 urine samples were obtained 

on a monthly and bimonthly basis, with the commercial laboratory Nuclear Engineering and Sciences 
Document No. ORAUT-TKBS-0041 Revision No. 00 Effective Date: 09/26/2008 Page 31 of 67

Corporation performing the urine analysis. The results frequently included high values that ranged 

from 50 to 150 μg/L of uranium in urine (AEC 1960a). However, the available dosimetry records do 

not contain urine bioassay results before late 1959.

Available uranium bioassay data indicate uranium was analyzed in urine from about late 1959 through 

1988, and in 1999 based on available bioassay reports. Information on the specific procedure used to 

analyze for uranium in urine is not known. Based on a bid specification (Barry 1964), the early 

analytical procedure likely 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 hydrochloric acid (HCl) and nitric acid (HNO3) then with 

hydrogen peroxide (H2O2) and finally with HNO3 to ensure destruction of all organic matter (Barry 

1964). 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 electro-deposition unit of the 

AEC laboratory design at a temperature of 950 F and 2 amps of current for 1 hour. The nickel disk 

was then dried and counted in a Nuclear Measurements Corporation gas flow proportional counter of 

the PC series (Barry 1964). 

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 that no 

absorption correction was necessary (Barry 1964).

The fluorimetric analysis would require a 5 ml sample volume and would have a sensitivity of 1 μg/L 

with a precision of ±10% (Barry 1964). However, the reported detection limit was 0.1 μg/L for results 

reported by Controls for Radiation, and 5 μg/L for results reported by Eberline in available bioassay 

reports. 

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

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 (BWXT 2006a, b,

c, d, e, f). If an MDC value is needed prior to the dates listed in the table, the values for the earliest 

date should be used. The measurements based on activity (gross alpha, gross alpha U, and EU) 

should be evaluated as total uranium activity.

The MDC value for the Controls for Radiation is set to 1 ug/L even though the vendor reports often 

indicate a value of 0.1 ug/L. Therefore, the value in Table 5-6 has been set to 1 μg/L based on the 

reported value for the analytical method by Barry (1964). This value should be used as a minimum 

value in place of the values reported in the bioassay records; larger reported values may be used in 

the intake assessment. This is consistent with the NUMEC health physics staff interpretation of the 

plutonium urine analysis results from Controls for Radiation, where they demonstrated that the 

accuracy was about 50 times less than the vendor indicated (5 dpm/day versus 0.1 dpm/day).

Care should be taken in evaluating the Controls for Radiation reported urine bioassay results as the 

records may have errors in the reported units. The reported values are in units of μg/ml although 

sometimes are listed as μg/L. 
Document No. ORAUT-TKBS-0041 Revision No. 00 Effective Date: 09/26/2008 Page 32 of 67

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

a,b

Date Laboratory Radionuclide Frequencyb MDCc Error

3/1961 –

2/1966

Controls for Radiation Total uranium Quarterly/as 

needed

<1 μg/Ld -

9/1972 –

12/1976

Eberline Total uranium Quarterly/as 

needed

<5 μg/sample -

1/1977 –

11/1987

Controls for 

Environmental Pollution

Total uranium Unknown e e

1999 Quanterra Total uranium Quarterly/as 

needed

<0.006 μg/L -

4/1962 –

1/1967 

Controls for Radiation Gross alpha Quarterly/as 

needed

26 dpm/L 13 dpm/L

2/1967-

8/1972

Tracerlab Gross alpha U Quarterly/as 

needed

0.2 dpm/sample 0.1 dpm/sample

9/1972-

1/1974

Eberline Gross alpha U Quarterly/as 

needed

<50.0 dpm/sample 0.05 dpm/ml

2/1974-

4/1974

Eberline Gross alpha U Quarterly/as 

needed

<10 dpm/sample -

2/1974 –

12/1976

Eberline Gross alpha U Quarterly/as 

needed

2 dpm/sample 1 dpm/sample

3/1964 -

6/1967

Controls for Radiation EU Quarterly/as 

needed

4 dpm/ L 2 dpm/L

7/1967 -

8/1972

Tracerlab EU Quarterly/as 

needed

0.2 dpm/sample 0.1 dpm/sample

1/1977-

2/1987

Controls for 

Environmental Pollution

EU Quarterly/as 

needed

e e

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

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

special sample analyses.

c. Assumes the MDC is twice the error, when an MDC is not available in the records. 

d. The MDC for Controls for Radiation for total uranium (1961 – 1966) is based on the reported value by Barry (1964) 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 is not to be used in internal dose assessments. 

Urine samples were typically 24-hr 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).

• 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 
Document No. ORAUT-TKBS-0041 Revision No. 00 Effective Date: 09/26/2008 Page 33 of 67

quarter for uranium workers. Special bioassays were ordered for those workers exceeding 40 MPChr of exposure or nose wipes exceeding 25 dpm.

There were approximately up to 100 urine bioassay analyses conducted each month. In the early 

years (to about 1964), urine samples were normally analyzed on a weight basis and then a 

radiometric analysis was performed if the level approached 50 μg/L. As stated above the urinary 

control levels were 50 μg/L and or 500 dpm/L for HEU (93%). According to the 1963 program review, 

the records for the past few years before 1963 indicated that there had been no restrictions as a result 

of the personnel monitoring program (AEC 1963). However, during a hazard evaluation conducted by 

the AEC in 1959, a number of personnel had a urine concentration result of 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 MAC (maximum allowable concentration) in urine was 500 dpm/L for 93%-enriched 235U (NUMEC 

1963). At some time during 1963 this was decreased to 300 dpm/L and by October of 1964 this was 

further decreased to 150 dpm/L (AEC 1964). The natural uranium urine control limit was 50 μg/L 

weight basis or 75 dpm/L activity basis (AEC 1963). 

By the mid 1960s both fecal and urine bioassay samples were being collected by NUMEC to 

determine the appropriate clearance model. The permissible natural uranium urine level of 75 dpm/d

was being used (Caldwell, Potter, and Schnell 1967). 

5.2.2.3 Uranium Fecal MDCs and Frequencies

The analytical procedure for uranium fecal analysis has not been located. The MDCs and 

approximate frequencies for the uranium urine bioassays are listed in Table 5-7 for NUMEC facilities. 

The fecal analysis results reported as dpm/sample should be considered equivalent to the daily 

excretion rate (dpm/day). When results are provided as dpm/gram along with the sample weight, the 

daily excretion value is also 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 MDC, frequency and period.a

Date Laboratory Radionuclide Frequencyb MDCd

6/1967–6/1972 Tracerlab Radiometric 

Uranium

Quarterly/as needed 2 dpm/sample

7/1972–1/1976 Eberline Total U Quarterly/as needed <5 μg/sample

2/1976–10/1985 Controls for Environmental 

Pollution

Total U Quarterly/as needed d

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

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

special sample analyses.

c. The MDC for radiometric uranium (Tracerlab) is based on a reported error value of about 1 dpm/sample, multiplied by 2.

d. Bioassay data analyzed by Controls for Environmental Pollution is not to be used in internal dose assessments. 

Fecal sampling (in addition to urine sampling) began on a large scale at the NUMEC Uranium plant in 

June 1966 (Caldwell 1966a). The fecal analysis continued until about 1985 as indicated in worker 

dosimetry records. Caldwell observed that some UO2 exposures were poorly detected in urine 

(Caldwell 1966a). According to Caldwell, literature available at the time indicated that whole-body

counting was effective for enriched uranium lung burdens greater than 7 nCi, but fecal sampling was 

necessary for smaller fractions of the permissible lung burden (Caldwell 1966a). 
Document No. ORAUT-TKBS-0041 Revision No. 00 Effective Date: 09/26/2008 Page 34 of 67

Caldwell used a permissible fecal excretion rate of 50 dpm/d for uranium assuming the ICRP 

recommended 380-day half time for chronic UO2 exposures (Caldwell, Potter, and Schnell 1967).

By 1972 or later, Caldwell believed that fecal sampling for all radionuclides was a valuable tool for 

early assessment of inhalation exposures but that information on the urine to fecal excretion ratio was 

necessary for the complete interpretation of urine data. Caldwell found that the most important use of 

fecal sampling data was for estimating the magnitude of single inhalations of uranium from accidental 

exposures. For uranium plant operations, Caldwell believed that lung burdens should be based on 

urine sampling or in vivo counting (Caldwell ca. 1972). 

5.2.3

There is not sufficient air-sampling or urinalysis information available for the NUMEC facilities in order 

to conduct a thorium intake analysis for workers in general. If the case files include thorium 

measurement results, then an intake and dose assessment can be performed. Thorium was 

processed at the Apollo facility for a few years starting in 1963. Limited information on thorium 

bioassay analyses was found in the worker dosimetry records. In 1969 the error was reported as 0.1 

dpm/sample for 232Th (Tracerlab analysis), which provides an MDA value of 0.2 dpm/sample. 

Thorium Exposures

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.

5.2.4

The bioassay records indicate urine bioassay analysis was 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 being provided by Controls for Radiation. If bioassay records are found 

in case files with results provided by Controls for Environmental Pollution, the results should not be 

used to estimate intake of mixed fission products.

Mixed Fission Products

Exposure to FPs was most likely to have occurred in the Laundry facility as part of the Apollo site 

commercial decontamination of clothing by laundering. The radionuclides representing mixed FPs 

could have included both fission and activation products representative of reactor operations. 

Possible radionuclides include 60Co, 90Sr, 99Tc (RU contaminant), 137Cs, 106Ru/Rh, and possibly others. 

No information is available on the analytical methods used to analyze for mixed FPs in urine. Urine 

bioassay data for mixed FPs should be used, when included in the case files, to estimate intakes of 

FPs. The technical information bulletin Fission & Activation Product Assignment for Internal DoseRelated Gross Beta & Gamma Analyses (ORAUT 2007c) can be used to determine the radionuclide 

appropriate for the dose calculation. This technical information bulletin requires specification of the 

decay time for the fission product mixture. Because little is known about the fission product material

likely to be present at the Apollo site, a decay time of 1 year should be assumed to provide a 

favorable to claimant assessment of dose (ORAUT 2007c).

The urine bioassay results do not indicate if the measurements are based on beta or gamma analysis, 

so the intake should be based on both methods according to guidance in the technical information 

bulletin. 
Document No. ORAUT-TKBS-0041 Revision No. 00 Effective Date: 09/26/2008 Page 35 of 67

5.2.5

The uranium processed at the Apollo site may have included recycled uranium. This material would 

contain contamination radionuclides formed during fission and activation processes when the material 

was irradiated in production or test reactors. The spent fuel elements were reprocessed to recover 

the uranium which was returned to the DOE inventories along with trace contaminants that included 

99Tc, 228Th, 237Np, 238Pu, and 239Pu. The intake of recycled uranium contaminant radionuclides can be 

estimated using the contaminant fraction values presented in Table 5-12 of the Fernald Technical 

Basis Document (ORAUT 2004). The estimated intake of each radionuclide is obtained by multiplying 

the estimated uranium intake by the appropriate contaminant fraction. The assignment of material 

solubility type is made based on the guidance provided in Table 5-2 of ORAUT-OTIB-0060 (ORAUT 

2007e). Because the plutonium contamination is a minor contaminant in the recycled uranium matrix, 

the consideration of type Super S plutonium is not necessary for evaluation of internal dose from this 

source of plutonium.

Unmonitored Radionuclides from Recycled Uranium

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 1966a, Caldwell 1968b). The counting performed in 1966 was provided 

by the University of Pittsburg Low Level Radioactivity Monitoring Facility at the PresbyterianUniversity Hospital, using a thin NaI crystal system (Caldwell 1966b). The bioassay records indicate 

this is the facility where the majority of routine whole body counts were performed for NUMEC 

workers starting in 1969. 

In 1968 and 1971 Helgeson performed whole body counts on individuals for fission products, 235U, 

241Am, with 239Pu estimated from the 241Am results based on expected activity ratios for 239Pu/241Am 

(Caldwell 1968b). The minimum detectable amount for 235U was listed as 0.08 mg for this system. 

The minimum detectable amount for 241Am ranged from 0.13 to 0.38 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 Pittsburg 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 in close proximity to the anterior 

chest region of the subject (BWTX 2006i, pp 92-97). The calibration was for 0.5 keV per channel and 

the count time was 40 min for both gross counts and background. Background correction was made 

using spectra obtained from unexposed individuals. Minor differences in the potassium and cesium 

body burdens were corrected by normalizing the spectra at an energy region from 90 to 125 keV. 

Activity calibrations were obtained from data published by Los Alamos Laboratory using a detector 

configuration identical to the one used by the Laboratory. The calibration factor was adjusted for 

attenuation due to variation in the subject’s chest wall thickness as measured with an 

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

present and all 17 keV X-rays were from 239Pu. The difficulty in measuring the low energy X-rays 

results in MDA values representing significant lung burdens (ORAUT 2007a).

Lung counts were performed from about 1966 to1992, possibly later. Uranium lung counting started 

regularly in December of 1971. Plutonium and americium counting started in 1966 (Caldwell 1966a) 

and on a regular basis in 1968. FPs were also counted intermittently. Lung counts are in general not 

as reliable as urinalysis (or fecal analysis – Caldwell 1966a) 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 
Document No. ORAUT-TKBS-0041 Revision No. 00 Effective Date: 09/26/2008 Page 36 of 67

uranium lung-counting minimum detectable activities (MDAs) for common enrichments that may have 

been processed at NUMEC. Actual MDAs from workers’ records should be used when available 

because the MDA for a measurement is dependent on the chest wall thickness, which varies by 

individual. The MDA for 235U was about 63 μg as indicated from the cursory review of worker 

dosimetry records in 1971 and later years, which is a reasonable default MDA value.

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

a

Uranium source term Total uranium MDA in μg Total uranium MDA in pCi

Natural uranium 8.75E+03 5.98E+03

93.00% 6.77E+01 4.61E+03

3.50% 1.80E+03 3.96E+03

2% 3.15E+03 5.09E+03

Typical DU 3.17E+04 1.27E+04

RU (1% 235U) 6.30E+03 5.73E+03

a. Based on U-235 MDA of 63 μg.

Table 5-9 is a summary of in vivo MDAs for 239Pu, 241Am based on a review of claimant files. The 

results are generally reported as whole-body counts in the dosimetry records. Data after 1985 are 

sparse in the bioassay records.

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

a

Year Pu-239 MDA (nCi) Am-241 MDA (nCi)

Minimum Maximum Counts Minimum Maximum Counts

1968 NRb NRb NRb 0.13 0.38 17

1969 NRb NRb NRb NRb NRb NRb

1970 NRb NRb NRb NRb NRb NRb

1971 NRb NRb NRb NRb NRb NRb

1972 9.0 11.5 3 0.13 0.13 1

1973 5.6 15.6 46 0.11 0.21 28

1974 5.44 21.3 122 0.09 0.22 96

1975 4.8 19.9 133 0.11 0.21 104

1976 5.0 20.3 109 0.11 0.19 91

1977 4.4 19.6 113 0.09 0.19 88

1978 4.7 19.0 132 0.10 0.19 100

1979 5.16 24.3 168 0.08 0.26 132

1980 5.03 28.2 132 0.09 0.21 94

1981 7.21 27.8 55 0.12 0.20 31

1982 7.12 34.3 77 0.12 0.21 44

1983 9.41 15.6 6 0.12 0.16 4

1984 8.67 22.32 9 0.12 0.15 5

1985 8.84 31.07 31 0.11 0.22 29

a. From a review of worker dosimetry records (BWXT 2006a,c,d,g,i,j). Values for 1968 through 1971 are based on the

Helgeson system, with remaining values for the University of Pittsburg system.

b. NR = none reported.

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.
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5.4 APOLLO PROCESS URANIUM AIR SAMPLING STUDIES

This discussion focuses on documented air-sampling data from five separate HASL reports by the 

AEC for the Apollo site. The information in the HASL reports can be useful in providing an estimate of 

the likely intake for a worker when job description and location are known. A general description of 

processes for various enrichments of uranium at Apollo is outlined in HASL Survey Reports 82, 92, 

103, 106, and 114 (Occupational Exposure to Radioactive Dusts reports), which cover the period from 

December 1959 to January 1961 and the Procedure for Recovery of Scrap Uranium from about 1963 

(AEC 1960a,b,c 1961a,b; NUMEC ca. 1963). HASL survey reports contain results for loose and total 

alpha samples, GA samples, fixed-station and weighted BZA samples, and some personnel protective 

equipment, ventilation description, and general observations of activities. Attachment A lists the 

results of all BZA surveys for HASL reports 82, 92, 103, 106, and 114. A summary of the HASL 

reports is listed in Table 5-10 based on information from Attachment A.

The air samples consisted of collection of radioactive particulates on filters from breathing zones and 

general areas during processing. The alpha activity measured on the filter was used to determine the 

airborne alpha activity concentrations. When multiple samples at a location were collected, the AEC 

used the mean air concentration in subsequent calculations. The AEC matched air concentration 

determinations with information about worker categories, locations, tasks, and time at each location or 

task. 

Table 5-10. HASL report summary.

HASL Report # Description of Report Date

82 Source: (AEC 1960a) HASL 82- Production of UO2 from UF6, UO2 pellet 

formation, U-graphite pellet production, Recovery of U3O8 HEU from U-Zr 

Scrap, R&D for coating U particles. Natural U was used in coating studies. 

Otherwise fully enriched (93%) HEU was used in the other processes.

12/1959

92 Source: (AEC 1960b) HASL 92- Production of UO2 from UF6, UO2 pellet 

formation, Uranium graphite pellet production, recovery of U3O8 HEU from UZr Scrap, R&D for coating Uranium particles. Natural Uranium was used in 

coating studies. Otherwise fully enriched (93%) HEU was used in the other 

processes.

6/1960

103 Source: (AEC 1960c) HASL 103- Production of UO2 from UF6, UO2 pellet 

formation, U-graphite pellet production, Recovery of HEU from U-Zr Scrap, 

R&D for coating Uranium particles. Natural U was used in coating studies. 

Otherwise (1.8 to 93%) Uranium was used in the other processes.

10/1960

106 Source: (AEC 1961a) HASL 106- Processing (93%) UO3 from UF6 and 93% 

U3O8 from U-Zr, powder handling of (93%) HEU in the Ceramics Laboratory. 

Pressing (3.5%) EU, centerless grinding (3.5%) EU, and sintering (3.5% and 

93%) EU and HEU were performed in the Ceramics Fabrication Area. 

Chemical reprocessing of (1.8%) EU and coating of U particles (93%) HEU 

were in operation.

12/1960

114 Source: (AEC 1961b) HASL 114- CP-2 chemical processing (3.4%) UO3 from 

UF6, CRP-2 chemical reprocessing (5.7%) U3O8 from U-Zr, Ceramics 

Fabrication (5.7%) EU. Uranium particle coating involved natural Uranium. 

No powder handling activities were evaluated.

5/1961

When estimating the intake for a specific worker, the dose reconstructor should look for all available 

information related to intakes of uranium. The information may include:

• Bioassay monitoring results

• Workplace breathing zone sampling results
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• General area monitoring results 

• Work location and job classification by time period

• Air concentration information from HASL reports (summarized in this section and 

Attachment A)

• Reports to the AEC/NRC of overexposures to airborne activity

• Reports to the worker of overexposures and work restrictions.

The bioassay monitoring results and workplace breathing zone results for the individual provide the 

best information because the data relate to the exposed individual. The reports to the AEC/NRC and 

work restriction letters also relate directly to the individual. The information related to work location 

and job classification is useful to establish potential for intakes, and for correlation to the HASL air 

sampling results. The HASL results have been analyzed to determine statistical information on air 

concentrations, as presented in tables in Attachment A. The analysis of the reported doses for all 

individuals from all HASL reports has been based on the assumption of the data being represented by 

a lognormal distribution. The highest value (6,300 dpm/m3

) is assumed to represent the 95% value 

and the lowest value (7 dpm/m3

) is assumed to represent the 5% value. The resulting distribution has 

a median value of 210 dpm/m3 and a geometric standard deviation of 7.91. This representation 

provides a more favorable to claimant estimate of air concentration than a strict numerical evaluation 

of the data. The reported values have several very high values that are not well captured in a 

standard statistical analysis.

Case files might contain information on breathing-zone air monitoring for individuals. This information 

can be used to establish potential for exposure and estimates of intakes for specific work tasks. In 

general, use of bioassay monitoring data should be used to estimate intakes when available. 

The results presented in the HASL reports represent estimates of daily exposures to airborne 

uranium. These include specific work tasks that may have involved much higher air concentrations, 

but for a short period of time. Although it is not possible to identify in specific detail an individual 

worker’s occupancy versus uranium airborne exposures to estimate uranium intakes during the entire 

operational period for NUMEC Apollo (1957-1983), it is believed that the results of the HASL BZA 

results can give a reasonable upper bound intake value. The HASL studies resulted in improvements 

to working conditions, reducing the overall exposure of workers to airborne uranium.

The median value from the HASL reported average daily exposures (applied as a lognormal 

distribution with a GSD of 7.91) would provide a reasonable estimate of the distribution of uranium air 

concentrations during the 1960 through 1983 period for individuals whose specific work location are 

not known. This intake should be limited to periods when the individual was likely to be involved with 

uranium work. It should not be applied to periods when the worker was on work restriction because of 

previous high exposures. This intake would represent a very favorable to claimant intake for 

individuals who did not work routinely in the Apollo uranium facility processes, such as health physics 

staff, inspectors, and maintenance workers.

The inhalation intake of uranium can be estimated from the air concentration by multiplying by the 

breathing rate of 1.2 m3

/hour and the annual period of exposure (hours).
Document No. ORAUT-TKBS-0041 Revision No. 00 Effective Date: 09/26/2008 Page 39 of 67

In the case where inhalation intakes are calculated from air concentrations, ingestion intakes are also 

to be considered. NIOSH (2004) states that the daily ingestion rate in picocuries can be estimated by 

multiplying the daily concentration in picocuries per cubic meter by a factor of 0.2 for an 8-hr workday. 

5.4.1

NUMEC Health Physics reported to the AEC anytime a radiation worker exceeded 40 MPC-hr in a 

workday or in a workweek. The individual dosimetry records should indicate if an individual exceeded 

the 40 MPC-hr limit. The dosimetry 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 when intakes may have occurred, and when 

intakes were unlikely. The individual dosimetry records should be used to reconstruct intakes on an 

individual basis, whenever possible. 

Apollo 40 MPC-hr Reports to AEC of Overexposures

Overexposures were required to be reported to the AEC/NRC. Overexposures were measured in 

terms of MPC-hours. If calculated MPC-hrs exceeded 40 for a week, then it was considered an 

overexposure. MPC-hrs 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 and it was assumed that no protection was afforded by the respirator 

and a bioassay was conducted. When a high nasal smear coincided with an impermissible air 

sample, then it was assumed that an overexposure had occurred and NUMEC reported the 

overexposure according to 10 CFR Part 20 requirements. If a high nasal smear could not be 

corroborated by a high air sample, then NUMEC only reported if the bioassay data indicated an 

overexposure (Shapiro 1969).

5.4.2

The following incidents occurred at the NUMEC facility in Apollo, Pennsylvania and may have 

contributed to employee exposures, but no exposure information was provided in the incident report 

or was reported as permissible. These incidents are in addition to those incidents/overexposures 

shown in Table 5-14. Information regarding individual involvement in incidents is likely to be 

contained in the workers’ dosimetry records, and should be consulted for evaluation of intakes of 

radionuclides during such events.

Other Incidents

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. A total of five bottles were damaged, containing approximately 

8.8 kilograms of HEU. NUMEC approximated about 0.5 kg of HEU may have been lost. No 

information is provided regarding 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. deep of material. 

The material was found to be approximately 400 kilograms of dry 10% uranium by weight and 3.3-

percent enriched in 235U; therefore, approximately 1.32 kilograms of 235U were present in the duct. 

The material was removed from the duct. Routine inspections of the ductwork were put in place and a 

HEPA filter installed (Reitler 1973).
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On April 20, 1974, a maximum of 6 kilograms of low-enriched UF6 was released to the in-plant 

atmosphere. A pipe and valve on the suction side of a hydrolysis column re-circulating pump failed 

and blew out from the penton pipe, releasing the water from the bottom of the hydrolysis column, thus 

releasing the UF6. Nasal smears were taken from all personnel involved, and all were within 

permissible limits (Fink 1974). 

5.5 APOLLO URANIUM AND PLUTONIUM AIR SAMPLING PROGRAMS

NUMEC uranium and plutonium workers wore lapel samplers starting in 1965. According to Caldwell, 

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 report surveys during the period from December 1959 to 

January 1961, and BZA sampling was observed in the 1963 Health Protection Program Review 

conducted by Oak Ridge Operations Office R&D Division. NUMEC used a Rochester Imaging 

Detector Laboratory gas flow proportional counter.

Before 1965, the BZA samples were likely fixed-station BZA samplers, and later NUMEC studies 

conducted in the 1966 -1967 time frame 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 samplers and general area air 

sampling usually underestimated airborne concentrations. Fifty percent of the lapel air sample results 

at Apollo showed concentrations seven times greater than the stationary air samples. The median of 

the ratio of lapel BZA to GA concentrations results was found to be ~7 at the NUMEC uranium and 

plutonium facilities (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 conducted in 1963, 

75 short-term breathing-zone air samples were obtained in Apollo process buildings every week but 

no routine general area air samples were taken which would indicate an average air concentration 

over an 8- to 24-hour period (AEC 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-

1967), whenever a BZA sample indicated that an exposure had occurred, the worker was removed 

from radiation work and both fecal and urine samples were collected (Caldwell and Schnell 1968). 

The MPCs in the NUMEC Health and Safety Manual (NUMEC 1963) were 1 × 10-10 μCi/ml or 

220 dpm/m3 for in-plant airborne uranium.

6.0

6.1 EXTERNAL EXPOSURE SOURCES AT APOLLO

OCCUPATIONAL EXTERNAL DOSE

The NUMEC Apollo Nuclear Fuel Facility started operations in 1957 with the small-scale production of 

HEU and LEU fuel. Between 1958 and 1983 the Apollo facility produced LEU dioxide fuel for use in 

commercial nuclear reactors. The process consisted of conversion of uranium hexafluoride (UF6) to 

uranium dioxide (UO2). In 1963 an additional production line was added to produce HEU fuel for US
Document No. ORAUT-TKBS-0041 Revision No. 00 Effective Date: 09/26/2008 Page 41 of 67

Naval propulsion reactors. From 1958 through the 1960s, NUMEC processed unirradiated EU scrap 

under license from the AEC (NIOSH 2007c). Smaller operations consisted of analytical laboratories, 

UO2 pellet production, and R&D into coating techniques for uranium particles (B&WNES 1997). HEU 

operations at NUMEC Apollo were discontinued in 1978, and LEU and all other processing operations 

involving radioactive materials had ceased by the end of 1983. In the mid-1960s, NUMEC was 

involved in production of thorium oxide (ThO2) pellets for use in nuclear fuel. 

6.2 RADIOLOGICAL EXPOSURE SOURCES

During NUMEC operations, occupational exposure was received primarily from handling or being in 

the vicinity of fissile material (either as feed or product) and from airborne radioactivity with the 

resultant buildup of surface contamination. Enriched UF6 arrived at NUMEC in approved shipping and 

storage cylinders, and was present in various forms (liquid, powder, or metal) to be converted for use 

in nuclear fuel. Little information has been obtained regarding potential exposures from handling 

thorium. Available information indicates PuBe neutron source production was performed at the Parks 

Township Site. Since there is considerable uncertainty as to whether neutron sources were 

manufactured at the Apollo facility and direct evidence (data) has not been found, the radiological 

exposures from neutron source production is not considered for the Apollo site. The primary sources 

of radiological external radiation exposure from operations performed at NUMEC Apollo are 

summarized as follows:

Source Exposure potential

HEU production: 1957–1978 Beta radiation primarily, possibly photon dose from U 

progeny such as radium, etc.

LEU production: 1957–1984 As above

HEU and LEU scrap recovery As above

UO2 pellet production, started in 1961 As above

R&D for coating U particles: start in 1961 As above

Thorium Operations, pellet production, started in 

1963

Beta radiation and more significant photon radiation

Laundry operations Uranium and thorium residues

6.2.1

The beta dose rate on the surface of yellowcake, a natural uranium compound, just after separation is 

negligible, but rises steadily thereafter due to the build-up 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 approximately 150 mrad/hr. 

Beta Radiation

6.2.2

Photon exposure rates are about 1.2 mrad/hr in contact with fresh yellowcake, but during the build-up 

of the uranium daughters 234Th and 234Pa in fresh yellowcake, the radiation levels increase somewhat 

for several months after yellowcake production. Photon exposure rates are estimated to be 

approximately 1.0 mrad/hr at 30 cm from a drum of aged yellowcake (NIOSH 2007d, Table 4). Little 

site-specific information has been obtained about thorium, which does emit significant photon 

radiation. 

Photon Radiation

6.2.3

Uranium compounds can be a source of neutrons from both spontaneous fission occurring in the 

isotopes of uranium and from alpha-neutron reactions with low atomic number materials, such as 

oxides and impurities. Neutron exposures from yellowcake, a natural uranium compound, are 

Neutron Exposures
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considerably lower than the photon exposures and are, therefore, not generally considered significant

based on analyses presented in ORAU-OTIB-0024 titled “Technical Information Bulletin: Estimation 

of Neutron Dose Rates from Alpha-Neutron Reactions in Uranium and Thorium Compounds” (ORAUT

2005b). This 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 feet is about 0.003 

mrem/hr for natural uranium, 0.016 mrem/hr for 5% enriched uranium, and 0.45 mrem/hr for 97+% 

enriched uranium. Little site-specific information has been obtained regarding thorium, which can also 

be a source of neutron radiation from alpha-neutron interactions. Similarly, little site-specific 

information has been obtained about PuBe neutron source production. 

6.3 PERSONNEL MONITORING 

From a cursory review of claimant records not all personnel were required to wear dosimeters but the 

work areas were monitored based on the available claimant information. For example, Nuclear 

Science and Engineering Corporation performed the personnel film dosimeter service for several 

years. In September 1963 Controls for Radiation Company provided a monthly service for 11 area 

film dosimeters that were positioned in process areas.

6.3.1

Records of beta/gamma radiation doses to individual workers from personnel dosimeters are available 

for NUMEC operations beginning in 1957 for Apollo as observed from a review of claimant records. 

Doses received by these dosimeters were recorded at the time of measurement and the results 

routinely reviewed by NUMEC operations and radiation safety staff to assess compliance with 

radiation control limits. The External Dose Reconstruction Implementation Guide (NIOSH 2007b)

indicates that these records represent the highest quality records for retrospective dose assessments.

Beta/Gamma Radiation

Film dosimetry was used from 1959 until about 1969, and then was replaced with thermoluminescent 

dosimeters (TLDs). There is evidence of NUMEC concern about film dosimetry over response to the 

low energy photons from plutonium (Caldwell 1966b). This issue seemingly resulted in a change in 

vendor to Landauer 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 

subsequently provided by Landauer beginning in 1976. Table 6-1 summarizes the respective venders 

and period of service; although the B&W records indicate the data work sheets and data for the TLD 

program, the actual calibration and dose calculation algorithm is not readily apparent. 

6.3.2

The source of neutron radiation at the NUMEC Apollo facility would be limited to work with HEU, and 

recycled uranium. Workers were monitored for neutron exposures likely with nuclear track emulsion, 

Type A (NTA) film from the respective commercial venders until about 1968, and then using TLDs 

thereafter. 

Neutron Radiation

Criticality dosimetry monitoring with an array of area critical assemblies that fed into a central system 

existed from at least 1963. As of at least September 1963 each visitor and employee was issued an 

indium foil criticality dosimeter as part of each security badge (NUMEC 1963).

6.3.3

External dosimetry technology minimum detectable levels (MDLs) used by NUMEC are expected to 

be similar to contemporary commercial vendor capabilities. Landauer film badges were reported in 

Limits of Detection
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January 1961 with a lower reporting limit of 10 mrem for gamma, 40 mrem from beta, 15 mrem from 

neutrons and 20 mrem from fast neutrons, based on information in claimant files. However, other 

documentation indicates that film dosimeter MDLs in the workplace are higher. This document 

recommends using MDLs of 40 mrem for gamma and beta radiation, and 50 mrem for neutron 

radiation (ORAUT 2006a, 2007d). Table 6-1 summarizes the venders, frequency, MDLs, and the 

annual missed dose.

Table 6-1. MDLs and potential missed photon, beta, or neutron dose.

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

Annual missed dosec (rem)

(frequency in parenthesis)

Nuclear Science & Engineering or 

Controls for Radiation film and NTA film

whole body

1957–1963 0.04 photons

0.04 beta

0.24 beta -photons (monthly)

0.05 neutrons 0.30 neutrons [fast](monthly)

Landauer film and NTA film

whole body

1964–1965a 0.04 photons

0.04 beta

0.24 beta photons (monthly)

0.05 neutrons 0.30 neutrons [fast](monthly)

Eberline film and NTA film 

whole body

1966–6/1968 0.04 photons

0.04 beta

0.24 beta photons (monthly)

0.05 neutrons 0.30 neutrons [fast](monthly)

NUMEC or Eberline (Film-Apollo) until 

about 1970 and TLD for neutron

whole body

7/1968–1975 0.02 photonsd

0.02 betad

0.12 beta photons (monthly)

0.02 neutronsd 0.12 neutrons (monthly)

Landauer TLD

whole body

1976–1995 0.01 photons

0.01 beta

0.06 beta photons (monthly)

0.01 neutrons 0.06 neutrons (monthly)

a. Landauer (1964).

b. Estimated MDLs for each dosimetry technology. Dose levels were recorded at values of less than the MDLs.

c. Annual missed dose calculated based on the MDL/2 method from NIOSH (2007a).

d. MDL during this period is likely twice the recording levels. 

Capabilities of dosimetry technology used by commercial or DOE Laboratory vendors were 

comparable in performance since similar films and calibration procedures were used. Characteristics 

of Landauer dosimetry capabilities are presented in Table 6-2 for beta, gamma, and neutron radiation 

monitoring (ORAUT 2006b).

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

Period Monitoring technique Dosimeter description

Beta/photon dosimeters – Whole Body

1957–5/1968 Photographic film badge Film badges contained single film packet. Three filters (front and 

back) were incorporated into film badge for energy dependence: 

cadmium, aluminum, and lead. 

The film badges likely had a gamma and X-ray: 30 keV to 20 MeV; 

beta: over 1.5 MeV, typical of vendor offered dosimetry.

6/1968–1975 NUMEC or Eberline 

TLD

Comprised of 2 TLD-700 chips, 2 TLD 600 chips, and 1 CaF2 for 

monitoring beta, X-ray, and gamma exposure.

The TLD badges likely had a gamma and X-ray: 30 keV to 20 

MeV; beta: over 1.5 MeV, typical of vendor offered dosimetry.

1976–present Landauer or Equivalent 

TLD (Z1 dosimeter -

1990)

Comprised of 3 TLD-700 chips for monitoring beta, X-ray, and 

gamma exposure. Insensitive to neutron radiation. Replaced 

Landauer G1 - Film.

The TLD badges likely had a gamma and X-ray: 30 keV to 20 
Document No. ORAUT-TKBS-0041 Revision No. 00 Effective Date: 09/26/2008 Page 44 of 67

Period Monitoring technique Dosimeter description

MeV; beta: over 1.5 MeV, typical of vendor offered dosimetry.

Beta/photon dosimeters – Wrist and Ring

July 1963–about 

5/1968

Film 

Landauer Type M (wrist 

beta-gamma) badges or 

equivalent.

Type M dosimetry was a film badge. Gamma and X-ray: 30 keV to 

20 MeV; beta: over 1.5 MeV.

The film badges likely had a gamma and X-ray: 30 keV to 20 MeV; 

beta: over 1.5 MeV, typical of vendor offered dosimetry.

About 6/1968–

1983

K5 wrist badge TLD Comprised of 3 TLD-100 chips.

The TLD badges likely had a gamma and X-ray: 30 keV to 20 

MeV; beta: over 1.5 MeV, typical of vendor offered dosimetry.

Neutron dosimeters – Whole Body

1957–5/1968 NTA film badge Film badges using NTA films: Fast neutrons undergoing elastic 

collision with content of emulsion or cellulose acetate base material 

produce recoil protons, which are recorded as photographic tracks 

in emulsion. Track density is a linear function of dose. Developed 

image exhibits tracks caused by neutrons, which can be viewed 

using appropriate imaging method (i.e., oil immersion) and 1000X 

power microscope or projection capability.

6/1968–1995 Landauer I8, I1 or RI 

Neutrak Extended 

Range dosimeter.

Combined TLD albedo neutron monitor with track recoil device 

(CR-39 [allyl diglycol carbonate]) that responds to proton recoil 

events. Neutron energy range was approximately 1 x 10-6 to 10 

MeV. Albedo response to thermal neutron radiation was subtracted 

to yield fast neutron dose. The “Neutrak ER” has an albedo 

element with above-described elements. Qualitative relationship 

was derived to determine ratios of neutrons of various energies. 

This badge was combined with G1 to make dosimeter, known as 

R1 that monitored beta, X-ray, gamma, and neutrons. 

Source: ORAUT (2006b).

6.3.4

In compliance with 10 CFR 20.407, NUMEC was required to submit reports of personnel exposure 

information to the AEC/NRC. Only two records from 1976 and 1977 were found in publicly available 

databases at the NRC that provide personnel exposure information (Bruer 1977, 1978). The annual 

whole-body doses were segregated into dose bins with the number of workers in each bin. The 

results of the research of NRC records are shown in Table 6-3.

Radiological Records

Table 6-3. Annual occupational radiation exposures at the Apollo facility (Bruer 1977, 1978). 

Year

Total 

number 

monitored

No. with 

measured 

dose

Number of individuals with whole-body doses in the ranges (rem)

<0.1

0.1 –

0.25

0.25 –

0.5

0.5 –

0.75

0.74 –

1.0 1.0 – 12 >12.0

1976 42 42 27 14 1 0 0 0 0

1977 39 39 15 16 6 0 2 0 0

In addition, an AEC inspection report showed the following highest beta-gamma and neutron 

exposures by quarter in 1973. Table 6-4 shows these findings.
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Table 6-4. Highest beta-gamma and neutron 

exposures measured for 1973 (Nelson 1974).

Quarter

Highest betagamma (mrem)Highest neutron 

(mrem)

1 141 169

2 153 215

3 94 773

4 111 352

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

Facilities like the Apollo site handling primarily uranium would not have significant exposure to low 

energy photons (ORAU 2006e). If there is an indication that the worker was exposed to low-energy 

photons from plutonium work, then the guidance in this section should be followed. 

Low-Energy Photons

Experience at NUMEC since the mid 1960s indicated the limitations of film external dosimetry for the 

monitoring of plutonium low-energy photons. Caldwell (1966b) 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

• Effective energy of 400 keV. 

There is no information on use of plutonium at the Apollo site. However, a study performed at the 

Parks Township site evaluated photon fields from plutonium work. A photon spectroscopy survey was 

conducted at the plutonium laboratory to try to ascertain the energy spectrum of the NUMEC Parks 

Township plutonium work environment. Even though it was suspected that the 17- and 60-keV

gamma radiation groups would dominate, uncertainties in geometry, self absorption and incidental 

shielding material would make theoretical or model calculations difficult (Caldwell 1966b).

Surveys were conducted of the plutonium chemical processing line and ceramics line. The 60-keV 

peak from 241Am was found to predominate. The 17-keV X-rays did not produce a peak and must 

have been absorbed in great part by the glovebox walls (Caldwell 1966b, p. 4). Higher energy peaks 

at 208, 267 and 333 keV were produced by 237U. The photon energy spectrum is presented in 

Figure 6-1.

A survey of the ceramics line and plutonium-uranium Mo alloy melt box line indicates a predominance 

of 60-keV 241Am gamma radiation. The gamma energy spectrum is presented 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 because the plutonium in the chemical processing line had 

aged an additional 2 years whereas the plutonium in the melt box had just been received (Caldwell 

1966b).
Document No. ORAUT-TKBS-0041 Revision No. 00 Effective Date: 09/26/2008 Page 46 of 67

Figure 6-1. Gamma spectrum external to 

plutonium chemical processing line and plutonium 

ceramics line (Caldwell 1966b, figure 3).

Figure 6-2. Gamma spectrum external to 

plutonium ceramics line and plutonium melt 

glovebox (Caldwell 1966b, figure 4).

An analysis of the Eberline film dosimeter response for OW (open window) versus AL (aluminum), PL 

(plastic), and CD (cadmium) filters was made. This is presented in Figure 6-3. 

Figure 6-3. Dosimeter filter ratios (Caldwell 1966b).

The following interpretation was made by Caldwell (1966b):
Document No. ORAUT-TKBS-0041 Revision No. 00 Effective Date: 09/26/2008 Page 47 of 67

1. Hot cell workers are exposed to 60Co and FPs. The energy response of the film badge is 

constant above 200 keV. The dose was taken directly from a 60Co calibration curve. If the 

OW/CD ratio was close to 1.0 the reported dose was accepted.

2. Plutonium workers were exposed to a wide range of gamma energies. The upper end of the 

spectrum would produce OW/CD ratios close to 1.0. The lower gamma energies would 

produce an OW/CD ratio of 3.0 or greater. One plutonium worker may 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, NUMEC would use a plutonium spectrum calibration 

curve that represented a typical plutonium gamma spectrum.

3. The OW/PL (plastic) ratio was about 1 and was within a 6% standard deviation. This meant 

that the large OW/CD ratio is not due to beta radiation. 

4. The OW/AL ratio was sensitive to X-ray exposures, but NUMEC did not incorporate this into 

their analysis.

Caldwell (1966b) presented an assessment of the energy dose fraction for personnel exposure due to

typical plutonium fuel fabrication from 1,000 MWD/T plutonium. Sixty five percent of the dose was 

from the 241Am 60-keV gamma. Less than 7% was from the highest energy groups (Caldwell 1966b). 

The summary of the energy dose fraction is presented in Figure 6-4. It was noted that the gonadal 

dose was 50% of the whole-body or trunk dose due to the effect of the steel bottom of the plutonium 

gloveboxes (Caldwell 1966b). Table 6-5 summarizes the gamma energy distribution for NUMEC 

plutonium in comparison with Hanford plutonium. Beta energies are included as well as 233U and 

241Am which have similar overall photon and beta properties.

0

20

40

60

80

Fraction of 

Dose

20 kev60 kev 115

kev

195

kev

335

kev

Energy in kev

 

 

 

Figure 6-4. Relative contribution by energy group to 

personnel exposure during plutonium fuel fabrication from 

1000 MWD/T plutonium (Caldwell 1966b, p. 18, Figure 8). 

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

Energy–photon NUMEC plutonium, (241Am & 233U) Hanford plutonium

<30 keV 15 25

30 – 250 keV 82 75

>250 keV 3 0

Energy–beta NUMEC plutonium, (241Am & 233U) Hanford plutonium

>15 keV 100 100
Document No. ORAUT-TKBS-0041 Revision No. 00 Effective Date: 09/26/2008 Page 48 of 67

6.4.2

NTA film is a poor detector of neutron energies below 500 to 800 keV (ORAUT 2006c). Table 6-6 

contains neutron dose correction factors for NTA film developed for dose reconstruction at Argonne 

National Laboratory-East (ANL-E) to account for missed low-energy neutrons. The facilities and 

activities at the Laboratory appear to encompass the activities and facilities at NUMEC. An NTA film

energy threshold correction factor of 1.5 is recommended for NUMEC due to the presence of uranium 

fuel and critical assemblies. If the information on a worker indicates that work with plutonium 

occurred, then the energy threshold correction factor of 4 would be appropriate.

Neutron Radiation

Table 6-6. NTA energy threshold correction factors.

Facility type

NTA film neutron dose 

correction factorsa

U Fuel and critical assemblies (and U-Th MOX) 1.5

Pu handling facilities (and U-Pu MOX) 4

a. Adapted from ORAUT (2006c).

6.5 DOSE RECONSTRUCTION RECOMMENDATIONS

6.5.1

Recorded and reported dose practices are summarized in Tables 6-7 and 6-8. 

Recorded Dose Practices

Table 6-7. Recorded dose practices. 

Year Dosimeter measured quantities Compliance dose quantities

Photon/electron film dosimeter + NTA neutron dosimeter 

1957–1971 Gamma (G) 

Neutron (N) 

Beta (B)

WB or total = gamma (photon) + neutron

Beta separate

Extremity = gamma (+ neutron)

Photon/electron film dosimeter + TLD neutron dosimeter

1972–1983 Deep = gamma and neutron (DBG)

Shallow beta gamma (SBG)

WB = gamma + neutron

Skin = beta 

Extremity = gamma + neutron

Photon/electron/neutron–Panasonic TLD + CR-39 neutron dosimeter 

1983–present Deep

Shallow

Skin = beta + soft gamma & neutron 

WB = photon + neutron 

Extremity = gamma + neutron

Table 6-8. Interpretation of reported data. 

Period

Reported 

quantity Description

Interpretation of 

zeroes

Interpretation of 

blanks (no data)

Rollup of 

individual and 

annual data

Monitored/ 

unmonitored

1957–

1971

R or rem Reported 

WB doses 

include 

gamma and 

neutron 

doses

Zeroes were generally 

not reported. Reported 

zeroes should be 

interpreted as meaning 

less than MDL.

The absence of data 

should be interpreted 

as individual was 

monitored with zero 

result.

Photon WB dose 

Neutron WB 

dose Shallow 

skin dose Total 

deep WB dose

All employees 

with significant 

exposure 

potential were 

monitored

1972–

1982

rem Reported 

WB doses 

qualified as 

either 

photon or 

neutron

Zeroes were generally 

not reported. Reported 

zero should be 

interpreted as meaning 

less than MDL.

The absence of data 

should be interpreted 

as individual was 

monitored with zero 

result.

Photon WB dose 

Neutron WB 

dose Shallow 

skin dose Total 

deep WB dose

All employees 

with significant 

exposure 

potential were 

monitored
Document No. ORAUT-TKBS-0041 Revision No. 00 Effective Date: 09/26/2008 Page 49 of 67

Period

Reported 

quantity Description

Interpretation of 

zeroes

Interpretation of 

blanks (no data)

Rollup of 

individual and 

annual data

Monitored/ 

unmonitored

1983–

present

rem Photon 

deep, 

neutron 

deep, and 

skin dose 

reported.

Zeroes were typically 

reported. Reported 

zero should be 

interpreted as meaning 

less than MDL.

No data or blanks 

should be interpreted 

as individual was 

monitored with zero 

result.

Photon WB dose 

Neutron WB 

dose Shallow 

skin dose Total 

deep WB dose

All employees 

with significant 

exposure 

potential were 

monitored

6.5.2

6.5.2.1 Beta Dose Adjustments

Adjustments to Recorded Dose

Landauer documentation also indicates that the Landauer or other badges used in the 1960s at 

NUMEC had a beta energy threshold on the order of 1.5 MeV resulting in an under reporting of 

shallow dose equivalent. Consequently, the recorded nonpenetrating film results should be multiplied

by 2.0 to determine the corrected beta dose (ORAUT 2005c). Nevertheless, if a NUMEC employee 

worked with a source term that included primarily beta emitting radionuclides with a maximum beta 

energy less than 1.5 MeV, the personal beta dose would be essentially unmonitored. 

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 ORAU 

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, as 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, as this minimizes the POC.

3. If the compensability decision may hinge on this issue, and if the partitioning of the 

nonpenetrating dose cannot be decided based on the available information, additional 

research may 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 are considered to provide a realistic estimates of the actual doses.

6.5.2.3 Neutron Dose Adjustments

Because very little neutron spectral information is available, this analysis assumed that the neutron 

doses reported were calculated using the quality factors (QFs) from National Council on Radiation 

Protection and Measurements Report 38 (NCRP 1971). The Report 38 QFs were averaged over the 

ICRP Publication 60 (ICRP 1991) energy groups. A dose multiplier of 1.91 was used for conversion 

of the Report 38 dose equivalent to the Publication 60 equivalent dose for the 0.1 to 2 MeV energy 

range (ORAUT 2006c), which includes 100% for HEU, EU, natural uranium, and others. 

Consequently, the neutron doses recorded on film for personnel at NUMEC should be multiplied by a 

factor of 3 to account for changes in QFs and NTA film energy threshold as shown in Table 6-9. 
Document No. ORAUT-TKBS-0041 Revision No. 00 Effective Date: 09/26/2008 Page 50 of 67

Table 6-9. Neutron dose correction factors.

a

Facility type

NTA film neutron 

dose correction 

factorsa

ICRP 38 to ICRP 

60 adjustment 

factor

Total neutron 

dose correction 

factor

U fuel and critical assemblies (and U-Th MOX) 1.5 1.91 3

a. Adapted from ORAUT (2006c).

6.5.3

The potential for missed dose exists when workers are exposed to radiation at levels below the 

detection limit of their personnel dosimeters.

Missed and Unmonitored Dose

6.5.3.1 Shallow Dose and Deep Dose

The assignment of missed dose based on dosimetry records is performed using guidance in the 

External Dose Reconstruction Implementation Guide (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 a zero or that is below the LOD/2. The LOD values for NUMEC dosimeters are provided 

in Table 6-1. 

For cases involving the skin as the target organ, guidance in the Technical Information Bulletin

Interpretation of Dosimetry Data For Assignment of Shallow Dose (ORAUT 2005c) should also be 

followed for assignment of missed shallow and deep doses. 

6.5.3.2 Neutron Dose

The potential missed neutron dose can be estimated from LOD values in Table 6-1 for monitored 

workers. The missed dose should be adjusted based on the LOD/2 method described above for 

shallow and deep dose. 

If monitoring records do not included neutron dosimetry information, but the worker is suspected of 

being exposed to neutrons, the neutron dose can be estimated using a neutron-to-photon dose ratio. 

Measurements or a study for a neutron to gamma ratio for uranium work has not been located. A 

neutron to gamma ratio of 0.2 has been determined for uranium processing in gaseous diffusion 

plants from measurements made by Cardarelli (1997). Exposure to uranium hexafluoride cylinders is 

a possible source of neutron exposure at the Apollo site. 

6.5.4

Dose reconstructors may incorporate consideration of uncertainty in the dose calculation for

measured and missed doses as follows:

Uncertainty

• The technology used to measure worker dose at NUMEC is similar to the technology used by 

commercial and AEC laboratory facilities. The errors in the penetrating dose are anticipated to 

be approximately ±30% and normally distributed. For non-compensable 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 (ORAUT 

2006d). A constant distribution is applied.
Document No. ORAUT-TKBS-0041 Revision No. 00 Effective Date: 09/26/2008 Page 51 of 67

• 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 (ORAUT 2006d). A lognormal distribution is applied.

6.5.5

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 approximately 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 2000). As naturally occurring uranium is enriched the photon dose is lowered but 

the spectra becomes correspondingly more energetic. The average energy of the spectra can 

increase from solid or liquid uranium sources since these may 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 30-250 keV category to result in a higher calculated organ dose under 

most situations.

Radiation Dose Fraction

NUMEC Apollo workers had limited potential for some exposure to a variety of radioactive sources in 

addition to uranium. This includes thorium, plutonium, and photon sources such as radium, 192Ir, 

137Cs and 60Co. Mixed fission product exposure could occur at the laundry facility that provided 

commercial laundering of contaminated clothing. Generally recommended categories for IREP input 

for the measured and assigned components of radiation dose are presented in Table 6-10 unless 

there is specific claim specific information regarding the source of radiation exposure.

Table 6-10. Beta, photon, and neutron radiation energies and percentages for IREP input. 

Description

Dates Radiation 

type

Energy 

Begin End selection Percent

Uranium facilities 1/1/1957 12/31/1983 Beta > 15 keV 100

1/1/1957 12/31/1983 Photon 30-250 keV 100

Thorium handling 1/1/1957 12/31/1983 Beta > 15 keV 100

1/1/1957 12/31/1983 Photon 30-250 keV 75

1/1/1957 12/31/1983 >250 keV 25

Photon and neutron sources 1/1/1957 12/31/1983 Beta > 15 keV 100

1/1/1957 12/31/1983 Photon 30-250 keV 50

1/1/1957 12/31/1983 >250 keV 50

1/1/1957 12/31/1983 Neutron 0.1 – 2 MeV 100

7.0

The B&W Apollo facility ceased manufacturing nuclear fuel in 1983. Final decommissioning of the 

facilities was completed in 1995. During the period of residual contamination, employees of 

subsequent owners and operators of this facility are also covered under the Energy Employees 

Occupational Illness Compensation Program Act. The residual period for the NUMEC Apollo plant 

covers the period from 1984 through 1995.

ESTIMATION OF EXPOSURE TO 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

If dosimeter readings are available for the residual period, then the dose should be based on the 

recorded and missed dose, as described in Section 6. The external dose from residual activity would 

be included in measured dose based on dosimeter readings for the workers. However, the recorded 
Document No. ORAUT-TKBS-0041 Revision No. 00 Effective Date: 09/26/2008 Page 52 of 67

dose may also include dose from current activities that are not related to DOE work, and, therefore, 

represent an overestimate of external dose. 

If external dosimetry information is not available for the residual period, then the following method is 

provided to estimate a favorable to claimant external dose.

An estimate of the mean surface concentration at the end of the operational period is described in 

Section 7.4, derived in support of internal dose estimates during the residual period. The analysis 

resulted in a mean surface concentration of 4.97 x 106 dpm/m2 (GSD = 7.91). This value can be used 

to estimate the annual external dose to workers exposed to the residual activity. The annual dose is 

evaluated as follows, using a dose conversion factor (DCF) for exposure to uniform activity on a 

ground plane.

Dose (rem/yr) = residual level (dpm/m2

) x DCF (rem/dpm/m2/hr) x exposure time (hr/yr)

The external dose conversion factor for exposure to isotopes of uranium and short-lived progeny is 

provided in Federal Guidance Report 12 (Eckerman and Ryman 1993). The median annual external 

dose from exposure to residual surface contamination is provided in Table 7-1 for organs considered 

in the Federal Guidance Report. These values are based on an exposure time of 2000 hours per 

year. The dose values are for exposure to natural uranium as this provides a higher external dose 

than other enrichments (except depleted uranium). This provides a favorable to claimant dose 

estimate because most uranium at the Apollo uranium facility was natural or enriched uranium. 

Natural uranium provides a higher external dose, per unit activity, because significant contributions 

come from the short-lived progeny of 238U (

234Th and 234mPa).

Table 7-1. External annual dose from residual 

surface contamination.

Organ Annual dosea rem

Adrenals 1.09E-03

Bladder wall 1.19E-03

Bone surface 2.99E-03

Brain 1.14E-03

Breast 1.56E-03

Esophagus 1.01E-03

Stomach wall 1.19E-03

Small intestine wall 1.12E-03

Upper large intestine wall 1.15E-03

Lower large intestine wall 1.16E-03

Kidneys 1.20E-03

Liver 1.19E-03

Lungs 1.26E-03

Muscle 1.48E-03

Ovaries 1.11E-03

Pancreas 1.06E-03

Red marrow 1.24E-03

Skin 0.378

Spleen 1.20E-03

Testes 1.55E-03

Thymus 1.17E-03

Thyroid 1.31E-03

Uterus 1.11E-03
Document No. ORAUT-TKBS-0041 Revision No. 00 Effective Date: 09/26/2008 Page 53 of 67

a. Values are input into IREP as a lognormal 

distribution, value listed is the geometric mean, the 

GSD is 4.1.

The dose should be entered into the IREP input as a lognormal distribution with a geometric standard 

deviation of 7.91 (parameter 2) as photons in energy range 30 – 250 keV as indicated in Table 6-10. 

This provides a favorable estimate of probability of causation for all organs, even though some of the 

photon energy is likely to be of higher energy.

7.2 EXTERNAL AMBIENT DOSE FROM RESIDUAL ACTIVITY 

All unmonitored workers are assigned external dose as described in Section 7.1. If the worker is 

monitored (the assigned external dose based on monitoring records), or unmonitored and 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 EXTERNAL MEDICAL X-RAY 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. Therefore, medical X-ray doses should 

not be assigned after 1983.

7.4 INTERNAL DOSE FROM RESIDUAL ACTIVITY IN THE WORKPLACE

If bioassay data is available during the residual period, then that data should be used to estimate and 

assign internal dose for the worker. Internal doses using bioassay data for this period should be 

evaluated using the guidance in Section 5. However, bioassay data is unlikely to be available after 

1985.

When bioassay data is not available, and the worker may have worked in areas containing residual 

activity, the following method may be used to assign internal dose. 

Guidance in the Technical Information Bulletin Dose Reconstruction during Residual Radioactivity 

Periods at Atomic Weapons Employer Facilities (ORAUT 2008) describes methods to estimate intake 

of radionuclides during residual periods. The method relates residual surface contamination to 

workplace air concentration during periods after completion of DOE work.

The majority of the DOE related work at the Apollo site was involved with scrap recovery and fuel 

fabrication operations in the Apollo uranium facility (East Bay of the Main Building). The HASL reports 

documented the average workplace air concentration that workers were exposed to during 1960 and 

1961. The results of these studies are described in Section 5. The resulting mean average daily air 

concentration was estimated to be 210 dpm/m3 with a geometric standard deviation of 7.91 (assuming 

a lognormal distribution). 

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

using guidance from the OTIB (ORAUT 2008). The annual deposition amount is estimated using a 

deposition velocity of 0.00075 m/s, with deposition assumed to occur for one year. Using this 

approach a surface concentration of uranium is estimated as follows.

210 dpm/m3 × 31,536,000 s/yr x 0.00075 m/s = 4.97 × 106 dpm/m2
Document No. ORAUT-TKBS-0041 Revision No. 00 Effective Date: 09/26/2008 Page 54 of 67

This mean surface concentration is described as a favorable to claimant level at the end of the 

operating period. 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 also due to depletion (ORAUT 2008). The depletion factors applied to each 

year are described in Table 3-1 of ORAUT (2008a). The depletion factors indicate the residual 

concentration at the end of the operational period is to be used for the first year, the second year is 

reduced by a factor of 0.03, and the third and remaining years are reduced by a factor of 0.0007 

(representing a constant concentration after the third year).

The air concentration for each year is estimated using a resuspension factor (ORAUT 2008) of 1x10-6

m-1

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

contamination level of 1.605x106 dpm/m2

, results in the air concentration and annual intakes given in 

Table 7-2. The intake evaluation is based on exposure for 2000 hours per year and an inhalation rate 

of 1.2 m3

/hr.

Table 7-2. Uranium air concentration and 

annual intake in the residual period.

Year

Air concentration, 

dpm/m3

Intake, 

dpm/year

1984 4.97 1.19E+04

1985 0.149 3.58E+02

1986–1995 0.00348 8.34E+00

The intakes in Table 7-2 can be used to estimate the internal dose to the target organ for the years of 

employment for the worker. The estimated internal doses are assigned as a lognormal distribution 

with a geometric standard deviation of 7.91, corresponding to the distribution of the average daily air 

concentrations used to estimate the annual uranium intake. The uranium intake is represented as 

234U in the dose estimate. The dose should be evaluated for the three uranium material solubility 

types of F, M, and S, with the dose from the highest type used in the IREP input. 

Although uranium-aluminum alloy was present at the Apollo scrap recovery facility, the form of the 

material was not likely to be an inhalation hazard and modeling an intake of uranium aluminide is not 

necessary at the NUMEC Apollo facility. 

The internal dose analysis should include the potential inadvertent ingestion of uranium activity, based 

on guidance in OCAS (2004). The daily intake rate (dpm/day) is estimated as 0.2 times the average 

daily air concentration expressed in units of dpm/m3

. Using the air concentration for the first year after 

the end of operations (1984) from Table 7-2, an intake rate of 0.99 dpm/day is obtained. This value 

should be applied for all years of the residual period, without depletion, as the transfer of material from 

the contaminated surfaces may not be reduced with time to the same extent that resuspension is 

reduced. This provides a favorable to claimant assessment of ingestion intake.

8.0

All information requiring identification was addressed via references integrated into the reference 

section of this document.

ATTRIBUTIONS AND ANNOTATIONS
Document No. ORAUT-TKBS-0041 Revision No. 00 Effective Date: 09/26/2008 Page 55 of 67

REFERENCES

Adams, W. C., 1993, “Interim Radiological Survey Letter Report for the Main Building Area and a 

Portion of the South Parking Lot Area – Babcock & Wilcox – Apollo Pennsylvania, Incinerator 

Amendment,” memorandum to J. Roth (U.S. Nuclear Regulatory Commission), July. [SRDB 

Ref ID: 19919] 

AEC (U. S. Atomic Energy Commission), 1960a, Occupational Exposure to Radioactive Dusts; Safety 

and Fire Protection, Nuclear Materials and Equipment Corp., Apollo, PA., HASL-82, New York 

Operations Office, New York, New York, February 17. [SRDB Ref ID: 41066].

AEC (U.S. Atomic Energy Commission), 1960b, Occupational Exposure to Radioactive Dusts; Safety 

and Fire Protection, Nuclear Materials and Equipment Corp., Apollo, PA., HASL-92; New York 

Operations Office, New York, New York, August 24. [SRDB Ref ID: 41068].

AEC (U.S. Atomic Energy Commission), 1960c, Occupational Exposure to Radioactive Dusts, Nuclear 

Materials and Equipment Corp., Apollo, PA., HASL-103; New York Operations Office, New 

York, New York, November 8. [SRDB Ref ID: 41070].

AEC (U. S. Atomic Energy Commission), 1961a, Occupational Exposure to Radioactive Dust, Nuclear 

Materials and Equipment Corporation, Apollo, PA., HASL-106; New York Operations Office, 

New York, New York, January 31. [SRDB Ref ID: 41072].

AEC (U.S. Atomic Energy Commission), 1961b, Occupational Exposure to Radioactive Dust, Nuclear 

Materials and Equipment Corp., Apollo, PA., HASL-114, New York Operations Office, New 

York, New York, June 15. [SRDB Ref ID: 41075] 

AEC (U.S. Atomic Energy Commission), 1963, Health Protection Program Review of Nuclear 

Materials and Equipment Corporation, Apollo Pennsylvania, September 16-17, ORO Research 

and Development Division, Oak Ridge, Tennessee. [SRDB Ref ID: 11963]

AEC (U.S. Atomic Energy Commission), 1964, Health Protection Program Review of Nuclear 

Materials and Equipment Corporation, Apollo Pennsylvania, October 6-7, ORO Research and 

Development Division, Oak Ridge, Tennessee. [SRDB Ref ID: 11961]

DOE (US Department of Energy), 2000, Guide to Good Practices for Occupational Protection in 

Uranium Facilities, DOE-STD-1136-2000, US Department of Energy, Washington, D.C., 

(p 6-8) 

Author unknown, 2004, Summary of NUMEC Apollo and Parks Township Plants, Oak Ridge, 

Tennessee, October. [SRDB Ref ID: 30958]

B&W (Babcock & Wilcox Company),1976a, “Liquid and Gaseous Effluent Summary, July through 

December 1975,” memorandum to J. P. O’Reilly, (U.S. Nuclear Regulatory Commission, 

Region I), Apollo, Pennsylvania, February. [SRDB Ref ID: 20081]

B&W (Babcock & Wilcox Company), 1976b, “Liquid and Gaseous Effluent Summary, January through 

June 1976,” memorandum to J. P. O’Reilly (U.S. Nuclear Regulatory Commission, Region I), 

Apollo, Pennsylvania, August. [SRDB Ref ID: 30953]
Document No. ORAUT-TKBS-0041 Revision No. 00 Effective Date: 09/26/2008 Page 56 of 67

B&W (Babcock & Wilcox Company), 1977a, “Liquid and Gaseous Effluent Summary, July through 

December 1976,” memorandum to J. P. O’Reilly (U.S. Nuclear Regulatory Commission, 

Region I), Apollo, Pennsylvania, February. [SRDB Ref ID: 20080]

B&W (Babcock & Wilcox Company), 1977b, “Liquid and Gaseous Effluent Summary, January through 

June 1977,” memorandum to B. H. Grier (U.S. Nuclear Regulatory Commission, Region I), 

Apollo, Pennsylvania, August. [SRDB Ref ID: 30951]

B&W (Babcock & Wilcox Company), 1978a, “Liquid and Gaseous Effluent Summary, July through 

December 1977,” memorandum to B. H. Grier (U.S. Nuclear Regulatory Commission, Region 

I), Apollo, Pennsylvania, February. [SRDB Ref ID: 20079]

B&W (Babcock & Wilcox Company), 1978b, “Liquid and Gaseous Effluent Summary, January through 

June 1978,” memorandum to B. H. Grier (U.S. Nuclear Regulatory Commission, Region I), 

Apollo, Pennsylvania, August. [SRDB Ref ID: 20078]

B&W (Babcock & Wilcox Company), 1979a, “Liquid and Gaseous Effluent Summary, July through 

December 1978,” memorandum to B. H. Grier (U.S. Nuclear Regulatory Commission, Region 

I), Apollo, Pennsylvania, February. [SRDB Ref ID: 30950]

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Document No. ORAUT-TKBS-0041 Revision No. 00 Effective Date: 09/26/2008 Page 63 of 67

ATTACHMENT A

BREATHING-ZONE AIR SAMPLE SURVEY DATA

Page 1 of 5

LIST OF TABLES

TABLE TITLE

A-1 HASL 103 weighted BZA survey, October 1960.....................................................................

PAGE

63

A-2 HASL 92 weighted BZA survey, June 1960............................................................................ 64

A-3 HASL 82 weighted BZA survey, December 1959 ................................................................... 65

A-4 HASL 106 weighted BZA survey, December 1960 ................................................................. 66

A-5 HASL 114 weighted BZA survey, May 1961........................................................................... 67

Table A-1. HASL 103 weighted BZA survey, October 1960.

a

Operator

Number of 

persons

Average daily weighted 

exposures in dpm/m3

Ceramics Laboratory 3 36

Ceramic Laboratory–Group Leader 3 24

Ceramics Fabrication–Reduction kiln Operator 3 65

Ceramics Fabrication–Pellet Press 1 19

Sinter Furnace Operator 3 16

Ceramics Reduction Tube Operator 1 590

Center less Grinder 2 23

Coated Sphere Operator 6 31

CRP-3 Leach Operator 4 26

CRP-3 Filter and Feed Prep Operator 4 39

CRP-3 Extraction Operator 3 28

CRP-3 ADU Filter and Reduction Kiln Operator 3 27

a. Data from AEC (1960c).
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ATTACHMENT A

BREATHING-ZONE AIR SAMPLE SURVEY DATA

Page 2 of 5

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

a

Operator

Number of 

persons

Average daily weighted 

exposures in dpm/m3

UF6 to UO3 Operator (Kiln and Filter) 3 370

UF6 to UF3 Operator (Hydrolysis and Waste Filter 6 73

Ceramic Reduction–Kiln Operator 3 96

Powder Preparation–Ceramics Fabrication 1 820

Ceramics Laboratory Operator 1 5,500

Sinter Furnace Operator 3 94

Center less Grinder 2 100

Ceramic Fabrication–Prepress Operator (Graphite) 2 6,300

Ceramic Fabrication–Press Operator (Graphite) 2 490

Ceramic Fabrication–Pellet Press Operator (Uranium) 2 73

Ceramics Laboratory–Group Leader 3 600

Inspection – Graphite 6 57

Inspection – Uranium 2 57

U-Zr Recovery Operator–Extraction 2 160

U-Zr Recovery Operator–Powder 1 820

Wet Chemistry Laboratory (Group Leader, Chemists, 

Technicians) 10 9

Gas Analysis–(Chemist, technician) 2 39

Spectrographic Laboratory–(Group Leader, Technician) 6 19

Metallographic (Group Leader and Technician) 3 10

Coated Sphere Operator 6 11

Maintenance 21 110

Sweepers 2 150

Sampler 3 110

a. Data from AEC (1960b).
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ATTACHMENT A

BREATHING-ZONE AIR SAMPLE SURVEY DATA

Page 3 of 5

Table A-3. HASL 82 weighted BZA survey, December 1959.

a

Operator

Number of 

persons

Average daily weighted 

exposures in dpm/m3

UF6 to UO3 Operator (Kiln and Filter) 4 790

Ceramics Reduction Furnace Operator 1 560

Ceramics Laboratory Operator 3 6,300

Sinter Furnace Operator 3 130

Ceramics laboratory–Group Leader 3 940

Co-precipitator Operator 2 950

U-Zr Recovery Operator-Chip 1 43

U-Zr Recovery Operator-Dissolving 1 49

U-Zr Recovery Operator-Reduction 1 39

Wet Chemistry Laboratory-Group Leader 1 37

Wet Chemistry Laboratory-Chemist 2 37

Wet Chemistry Laboratory-Technician 3 37

Gas Analysis–Chemist 1 7.5

Gas Analysis–Technician 1 7.5

Spectrographic Laboratory–Group Leader 1 17

Spectrographic Laboratory–Technician 3 17

Metallographic Group Leader 1 7

Grinding and Polishing Technician 1 7

Coated Sphere Operator 3 12

a. Data from AEC (1960a).
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ATTACHMENT A

BREATHING-ZONE AIR SAMPLE SURVEY DATA

Page 4 of 5

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

a

Operator

Number of 

persons

Average daily weighted 

exposures in dpm/m3 % U

Ceramics Laboratory Technician 2 680 93

Ceramics Laboratory Powder Production 6 910 93

Ceramics Laboratory Clerk 1 310 93

Ceramics Laboratory Group Leader 1 310 1.8-93

Ceramics Fabrication-Center less Grinder 2 190 3.5

Ceramics Fabrication-Pellet Press 2 120 3.5

Ceramics Fabrication-Sinter Furnace 3 120 3.5-93

Ceramics Fabrication-Quality Control 4 61 3.5

CP-2 (UF6 to UO2) Filter and Drying 3 190 93

CP-2 (UF6 to UO2) Hexdrolysis 3 140 93

U-Zr powder 3 220 93

U-Zr Extraction Cascade 3 180 93

CRP-3 Extractor 3 110 1.8

CRP-3 Precipitate and Filter 1 82 1.8

CRP-3 Helper 1 88 1.8

Coated Sphere Operator 6 46 93

a. Data from AEC (1961a).
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ATTACHMENT A

BREATHING-ZONE AIR SAMPLE SURVEY DATA

Page 5 of 5

Table A-5. HASL 114 weighted BZA survey, May 1961.

a

Operator

Number of 

persons

Average daily weighted 

exposures in dpm/m3 % U

Ceramics Laboratory Reduction Tubes 1 130 5.7

Ceramics Fabrication-Pellet Press 2 35 5.7

Ceramics Fabrication-Center less Grinder 2 200 5.7

Ceramics Fabrication-Sinter Furnace 3 27 3.4-5.7

Ceramics Fabrication-Quality Control 2 33 5.7

Ceramics Fabrication-Group Leader 2 31 5.7

CP-2 (UF6 to UO3) Hydrolysis 3 24 3.4

CP-2 Ammoniation and Filtering ADU 3 27 3.4

CP-2 filtrate 3 20 3.4

CP-2 calciner 3 57 3.4

CP-2 Reduction kiln Operator 3 31 3.4

CRP-2 (U-Zr) Dissolving and filtering 3 39 5.7

CRP-2 Ammoniation, ADU Filter Conversion 3 27 5.7

CRP-2 Helper 3 33 5.7

CRP-2 Extraction Cascade 3 27 5.7

Coated Sphere 6 22 Normal

a. Data from AEC (1961b).