Page 1 of 65

DOE Review Release 12/11/2015

Feed Materials Production Center – 

Occupational Environmental Dose

ORAUT-TKBS-0017-4 Rev. 03

Effective Date: 12/03/2015

Supersedes: Revision 02

Subject Expert(s): Robert L. Morris, Billy P. Smith, Eugene Potter

Document Owner 

Approval: Signature on File

Karen S. Kent, Document Owner

Approval Date: 12/02/2015

Concurrence: Signature on File

James P. Griffin, Deputy Project Director

Concurrence Date: 12/02/2015

Concurrence: Signature on File

Edward F. Maher, Objective 3 Manager

Concurrence Date: 12/02/2015

Concurrence: Vickie S. Short Signature on File for

Kate Kimpan, Project Director

Concurrence Date: 12/02/2015

Approval: Signature on File

David S. Sundin, Deputy Director, DCAS

Approval Date: 12/03/2015

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

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

New Total Rewrite Revision Page Change
Document No. ORAUT-TKBS-0017-4 Revision No. 03 Effective Date: 12/03/2015 Page 2 of 65

 

PUBLICATION RECORD

EFFECTIVE 

DATE

REVISION 

NUMBER DESCRIPTION

04/06/2004 00 New Technical Basis Document for the Fernald Environmental 

Management Project – Occupational Environmental Dose. First 

approved issue. Initiated by Samuel L.T. Chu.

02/07/2006 01 Approved issue of Revision 01 to add NIOSH “boilerplate” language. 

Added totals to tables 4.2a and b, revised tables 4.9a and b to 

include 222Rn, table 4-10a to include intakes uranium and 

nonuranium radionuclides and table 4-10b to include site wide 

intakes of 222Rn, added 222Rn concentrations in Table 4A in Appendix 

A. Constitutes a total rewrite of the document. Initiated by Melton H. 

Chew.

03/13/2014 02 Revision initiated to update information. Changed site name to Feed 

Materials Production Center. Incorporated Sanford Cohen & 

Associates matrix issues. Added Special Exposure Cohort 

information and includes additional references, updates, and 

personnel radon exposure data information. Incorporated an 

auditable set of calculations in a referenced spreadsheet. Combined 

all intake rate calculations into a single spreadsheet available in the 

Site Research Database and deleted Attachment E, which contained 

sample exposure calculations and the detailed calculation results, 

and Attachment F, which contained the exposure rate calculation 

results. After 1988, operations ceased, environmental monitoring 

was increased, and so the Exposure Areas (EA) used to define 

intakes prior to 1989 is replaced by the standard approach defined in 

ORAUT-PROC-0031. Incorporates formal internal and NIOSH 

review comments. Constitutes a total rewrite of the document. 

Training required: As determined by the Objective Manager. 

Initiated by Karen S. Kent.

12/03/2015 03 Revision initiated to modify the K-65 radon emissions from the 

median value to the 95th-percentile value, to include Am-241 in the 

list of radionuclides that contaminate recycled uranium, and to 

address S. Cohen & Associates comments. These revisions were 

made to an auditable set of calculations in an updated referenced 

spreadsheet. Incorporates minor changes made for clarification 

purposes. Incorporates formal internal and NIOSH review 

comments. Constitutes a total rewrite of the document. Training 

required: As determined by the Objective Manager. Initiated by 

Karen S. Kent.
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TABLE OF CONTENTS

SECTION TITLE PAGE

Acronyms and Abbreviations ............................................................................................................... 6

4.1 Introduction .............................................................................................................................. 8

4.1.1 Purpose........................................................................................................................ 9

4.1.2 Scope........................................................................................................................... 9

4.1.3 Special Exposure Cohort.............................................................................................. 9

4.2 Site Background and History.................................................................................................. 11

4.3 Sitewide Occupational Environmental Dose Summary........................................................... 11

4.4 Internal Doses from Atmospheric Radionuclide Concentrations ............................................. 17

4.4.1 Radioactive Materials Air Concentrations and Intakes, 1951 to 1988.......................... 17

4.4.1.1 Radionuclide Discharges and Sources of Emissions.................................... 17

4.4.1.2 Exposure Site Locations............................................................................... 23

4.4.1.3 Airborne Radionuclide Concentration Derivation .......................................... 24

4.4.1.3.1 Radionuclide Emission Rates, Ri × RF ...................................... 24

4.4.1.3.2 Far-Field Atmospheric Dilution Factor, χ/Q................................ 25

4.4.1.3.3 Receptor Downwind Distances.................................................. 27

4.4.1.3.4 Effects of Wind Direction, fi ....................................................... 28

4.4.1.3.5 Concentration Fields Near Buildings, Modified χ/Q for 

Building Wake ........................................................................... 29

4.4.1.3.6 Episodic Releases..................................................................... 30

4.4.1.4 Radionuclide Intake Derivation..................................................................... 31

4.4.1.4.1 Radon ....................................................................................... 33

4.4.1.4.2 Radon Exposure Records ......................................................... 33

4.4.1.5 Results......................................................................................................... 34

4.4.2 Nonradon Radionuclide Air Concentrations and Intakes, 1989 to 2012 ...................... 34

4.4.2.1 Annual Air Emissions for Radionuclides Other than Radon.......................... 34

4.4.2.1.1 Monitored Stacks....................................................................... 35

4.4.2.1.2 Unmonitored Stacks .................................................................. 35

4.4.2.1.3 Scrubbers.................................................................................. 35

4.4.2.1.4 Building Vents ........................................................................... 35

4.4.2.1.5 Fugitive Emissions .................................................................... 35

4.4.2.2 Results......................................................................................................... 36

4.4.3 Radon Air Concentrations and Intake Estimates, 1989 to 2012 .................................. 36

4.5 Onsite External Doses from Ambient Radiation...................................................................... 37

4.5.1 K-65 Silos and Production Plants External Dose Contribution .................................... 37

4.5.2 External Dose Contribution from Waste Pits............................................................... 45

4.5.3 Onsite Ambient Dose Rate Estimates, 1976 to 2005 .................................................. 46

4.5.4 Onsite Ambient Dose Rate Estimates, 1952 to 1975 .................................................. 47

4.5.4.1 Dose Rate Due to K-65 Silos, Before 1976 .................................................. 47

4.5.4.2 Dose Rate Due to Production and Storage Facilities, Before 1976............... 51

4.5.5 Calculated Dose Rates and Recorded Measurements Comparisons.......................... 51

4.6 Attributions and Annotations .................................................................................................. 51

References ........................................................................................................................................ 53
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Glossary ............................................................................................................................................ 59

ATTACHMENT A INCORPORATION OF Q-11 SILOS AS A RADON RELEASE SOURCE, 

1952 TO 1958................................................................................................. 61

ATTACHMENT B 95TH-PERCENTILE WIND ROSE DATA FOR 16 COMPASS POINTS.......... 63

ATTACHMENT C ANNUAL AVERAGE RADON BACKGROUND CONCENTRATIONS AND 

CONCENTRATION AT SELECTED RADON LOCATIONS, 1989 TO 

2005................................................................................................................65

LIST OF TABLES

TABLE TITLE PAGE

4-1 Facilities and buildings in each EA......................................................................................... 12

4-2 Maximum sitewide average (1951 to 1988) or median (1989 to 2012) annual intake rates 

via inhalation..........................................................................................................................13

4-3 External gamma radiation ...................................................................................................... 15

4-4 Historical information and data available for environmental dose reconstruction .................... 17

4-5 Uranium emissions summary ................................................................................................. 19

4-6 Summary of six episodic releases of uranium from incident reports and air monitoring .......... 21

4-7 Fugitive thorium emissions from wind erosion of Waste Pits 2, 3, 4, and 5............................. 21

4-8 Thorium emissions data by year............................................................................................. 22

4-9 Releases of 222Rn used in this analysis for 1952 through 1988 .............................................. 23

4-10 Receptor downwind estimated distances from contributing emission sources........................ 27

4-11 95th-percentile values from wind rose data ............................................................................ 29

4-12 Building and K-65 silo dimensions for building wake χ/Q calculations .................................... 30

4-13 Uranium enrichment isotopes................................................................................................. 31

4-14 RU contaminant radionuclides ............................................................................................... 32

4-15 Ratio of 226Ra and 230Th to uranium........................................................................................ 33

4-16 Alphabetical list of fields......................................................................................................... 34

4-17 Medians of the maximum values of all uranium air concentration samples by year and 

maximum fence line values of 230Th during waste pit remediation activities............................ 36

4-18 Medians of the maximum values of all on-site and fence line radon air concentrations 

samples by year.....................................................................................................................40

4-19 Radiation dose rates at site boundary .................................................................................... 42

4-20 TLD measurement summary.................................................................................................. 44

4-21 Waste pit dose rate data ........................................................................................................ 46

4-22 Comparison of estimated dose rates in EAs and aerial survey results.................................... 50

4-23 Projected onsite dose rates based on systems capacity......................................................... 52

B-1 Wind rose data and derived 95th-percentile values for environmental radionuclide 

concentration estimates ......................................................................................................... 63

LIST OF FIGURES

FIGURE TITLE PAGE

4-1 EAs for 1951 to 1988 as used in this analysis ........................................................................ 12

4-2 Measured emission stack particle size from 15 different FMPC stacks................................... 25

4-3 Percentage of stack emision less than approximately 10 μm AMAD. ..................................... 25
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4-4 Horizontal diffusion σy versus downwind distance from point source for Pasquill’s stability 

categories ..............................................................................................................................26

4-5 Vertical diffusion σz versus downwind distance from continuously emitting point source 

for Pasquill’s stability categories............................................................................................. 27

4-6 95th-percentile wind rose data, 10-m height, 1991 to 1997 and 2000 to 2001........................ 28

4-7 Offsite and fence line radon monitoring locations ................................................................... 38

4-8 Onsite radon monitoring locations .......................................................................................... 39

4-9 Fence line TLD locations, 1976 to 1995 ................................................................................. 43

4-10 TLD monitoring locations, 1996 to 2005................................................................................. 44

4-11 Beta/gamma dose rate contours in Waste Pit Area ................................................................ 45

4-12 Waste pit dose rate profiles.................................................................................................... 46

4-13 1976 exposure rate contour map of FMPC survey area ......................................................... 48

4-14 1985 exposure rate contour map of FMPC survey area ......................................................... 49

4-15 Projected K-65 silos fence line dose rates histogram ............................................................. 50

A-1 Mean and highest yearly worker radon exposure estimates due to Q-11 silos ....................... 61

C-1 Annual average radon background concentrations and concentration at selected radon 

locations, 1989 to 2005 .......................................................................................................... 65
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ACRONYMS AND ABBREVIATIONS 

AMAD activity median aerodynamic diameter

AMS air monitoring station

AWE Atomic Weapons Employer

Bq becquerel

Ci curie

d day

DOE U.S. Department of Energy

DOL U.S. Department of Labor

DCF dose conversion factor

DR dose reconstructor

EA Exposure Area

EEOICPA Energy Employees Occupational Illness Compensation Program Act of 2000

FEMP Fernald Environmental Management Project

FMPC Feed Materials Production Center

ft foot

g gram

hr hour

ICRP International Commission on Radiological Protection

IREP Interactive RadioEpidemiological Program

kg kilogram

L liter

m meter

mg milligram

min minute

mR milliroentgen

mrad millirad

mrem millirem

MTU metric ton of uranium

NA not available

NIOSH National Institute for Occupational Safety and Health

ORAU Oak Ridge Associated Universities

pCi picocurie

ppb parts per billion

RU recycled uranium

s second

SRDB Ref ID Site Research Database Reference Identification (number)
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SEC Special Exposure Cohort

TBD technical basis document

TLD thermoluminescent dosimeter

U.S.C. United States Code

WL working level

WLM working level month

yr year

μBq microbecquerel

μm micrometer

μR microroentgen

§ section or sections
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4.1 INTRODUCTION

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

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

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

dose reconstructions at particular Department of Energy (DOE) or Atomic Weapons Employer (AWE) 

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

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

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

of the individual work required for each dose reconstruction.

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

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

“AWE facility” or a “DOE facility.” The terms AWE and DOE facility are defined in sections 7348l(5) 

and (12) of the Illness Compensation Program Act of 2000 (EEOICPA), respectively. An AWE facility 

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

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

atomic weapon, excluding uranium mining or milling.” 42 U.S.C. § 7384l(5). On the other hand, a 

DOE facility is defined as “any building, structure, or premise, including the grounds upon which such 

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

on behalf of, the [DOE] (except for buildings, structures, premises, grounds, or operations ... 

pertaining to the Naval Nuclear Propulsion Program),” and with regard to which DOE has or had a 

proprietary interest; or “entered into a contract with an entity to provide management and operation, 

management and integration, environmental remediation services, construction, or maintenance 

services.” 42 U.S.C. § 7384l(12). The Department of Energy (DOE) determines whether a site meets 

the statutory definition of an AWE facility and the Department of Labor (DOL) determines if a site is a 

DOE facility and, if it is, designates it as such.

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

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

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

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

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

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

at a DOE facility includes individuals who are or were employed by DOE and its predecessor 

agencies, as well as their contractors and subcontractors at the facility. Unlike the abovementioned 

statutory provisions on DOE facility definitions that contain specific descriptions or exclusions on 

facility designation, the statutory provision governing types of exposure to be included in dose 

reconstructions for DOE covered employees only requires that such exposures be incurred in the 

performance of duty. As such, NIOSH broadly construes radiation exposures incurred in the 

performance of duty to include all radiation exposures received as a condition of employment at 

covered DOE facilities in its dose reconstructions for covered employees. For covered employees at 

DOE facilities, individual dose reconstructions may also include radiation exposures related to the 

Naval Nuclear Propulsion Program at DOE facilities, if applicable. No efforts are made to determine 

the eligibility of any fraction of total measured exposure for inclusion in dose reconstruction.

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

duty as a condition of employment at a DOE facility. Therefore these exposures are not included in 

dose reconstructions for covered employees (NIOSH 2010):

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

conventional structures

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

reasons
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4.1.1 PURPOSE

This technical basis document (TBD) is part of the overall Feed Materials Production Center (FMPC) 

site profile, which describes plant facilities and processes, historical information, and environmental 

data in relation to dose reconstruction for FMPC workers. This TBD provides a technical basis for 

evaluating the occupational environmental dose for EEOICPA claimants who were employed at 

FMPC. Since operations ended in 1988, the Fernald site has been variously named the Fernald 

Environmental Management Project (FEMP), the Fernald Closure Project, and is now the Fernald 

Preserve. Occupational environmental dose refers to radiation exposures workers received from 

radioactive material discharges to the atmosphere from the facilities, while outside the facilities. In 

addition, environmental dose includes ambient external radiation that originated in the facilities but is 

outside the buildings or facilities. FMPC employees who did not wear external dosimetry or who were 

not fully monitored for external exposures are assigned environmental doses.

4.1.2 SCOPE

This TBD describes the estimated annual intakes for inhalation exposure and estimated radiation 

doses from ambient external exposures at FMPC to provide the basis for estimating the 

environmental dose to unmonitored workers or for years and radionuclides that have not already been 

assessed for dose based on positive bioassay results or missed dose, or whose records were 

incomplete or unavailable.

For the operational years (1951 to 1988) when stack monitoring efforts were extensive and when 

environmental monitoring systems were not comprehensive, air concentrations were derived from 

facility radionuclide emissions. Environmental measurements were used to establish intake rates for 

years after 1988.

This TBD provides annual environmental intakes and ambient external doses for 1951 to 2012. 

Intake rates and exposure after 2012 should be assigned using the values established for 2012.

Exposures to the skin, including localized doses from direct deposition should be reconstructed in 

accordance with ORAUT-OTIB-0017, Interpretation of Dosimetry Data for the Assignment of Shallow 

Dose (ORAUT 2005), and other applicable Project guidance.

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

justification, or clarification of the associated information, are presented in Section 4.6. 

4.1.3 SPECIAL EXPOSURE COHORT

The Secretary of the U.S. Department of Health and Human Services has designated three classes of 

employees at Fernald as additions to the SEC:

Thorium, 1968 through 1978

All employees of DOE, its predecessor agencies, and their contractors, or 

subcontractors who worked at the Feed Materials Production Center in Fernald, Ohio, 

from January 1, 1968 through December 31, 1978, for a number of work days 

aggregating at least 250 work days, occurring either solely under this employment, or 

in combination with work days within the parameters established for one or more other 

classes of employees included in the SEC (Sebelius 2012). 
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It was determined that NIOSH lacked the sufficient information to allow it to estimate with sufficient 

accuracy the potential internal doses from exposure to thorium, to which employees at this facility may 

have been subjected.

Thorium, 1954 through 1967

All employees of the DOE, its predecessor agencies, and their contractors and 

subcontractors who worked at the Feed Materials Production Center in Fernald, Ohio, 

from January 1, 1954, through December 31, 1967, for a number of work days 

aggregating at least 250 work days, occurring either solely under this employment, or 

in combination with work days within the parameters established for one or more other 

classes of employees included in the Special Exposure Cohort (Sebelius 2013a).

It was determined that NIOSH lacked sufficient information to allow it to estimate with sufficient 

accuracy the potential internal doses from exposure to thorium, to which employees working at this 

facility may have been subjected.

Uranium, 1951 through 1983 [applies only to subcontractors]

All employees of the Feed Materials Production Center in Fernald, Ohio, who were not 

employed by National Lead of Ohio, NLO, or the Department of Energy or its 

predecessor agencies, who worked at FMPC from January 1, 1951, through 

December 31, 1983, for a number of work days aggregating at least 250 work days, 

occurring either solely under this employment, or in combination with work days within 

the parameters established for one or more other classes of employees included in the 

Special Exposure Cohort (Sebelius 2013b). 

It was determined that NIOSH lacked sufficient information to allow it to estimate with sufficient 

accuracy the potential internal doses from exposure to uranium, to which employees of FMPC 

working at this facility may have been subjected.

This dose reconstruction infeasibility for the period from 1951 through 1983 applies only to 

subcontractors. NIOSH has access to an electronic dataset that contains the results of the uranium 

urinalysis bioassay program for all years of FMPC operations, and the overwhelming majority of 

employees of the prime contractor (National Lead of Ohio, later named NLO, Inc.) have results in the 

bioassay dataset. However, the dataset does not contain bioassay results for employees of 

companies other than the prime contractor (i.e., non-prime contractor employees). NIOSH has 

obtained a limited number of bioassay samples from non-prime contractor employees through data 

captures, but cannot be certain that all non-prime contractor employee bioassay data were retained 

by the site or captured by data capture efforts. In addition, there are some reasons to conclude that 

the prime contractor did not consistently evaluate whether non-prime contractor employees should be 

monitored for radiation exposure because of the transitory nature of their work (Sebelius 2013b). 

Dose reconstruction guidance in this document for the period from January 1, 1954, to December 31, 

1983, is presented to provide a technical basis for partial dose reconstructions for claims not 

compensated under the SEC (i.e., nonpresumptive cancers and SEC employment less than 250 

days). Although it is not possible to completely reconstruct internal radiation doses for all workers for 

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

concurred, that it is feasible to reconstruct external radiation doses for all FMPC workers for this 

period (Sebelius 2013b).
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4.2 SITE BACKGROUND AND HISTORY

FMPC began limited operations in 1951 and was fully operational by the end of 1954. Its primary 

function was to convert uranium ore concentrates and recycled materials to either uranium oxides or 

highly purified uranium ingots and billets for machining or extrusion into tubular forms of standard 

isotopic assays. These products were for use as production reactor fuel cores and target fuel element 

fabrication. In addition, FMPC processed relatively small amounts of thorium.

Uranium metal deliveries peaked in 1960 at approximately 10,000 metric tons of uranium (MTU) 

(Mead, Savage, and Fugate ca. 1985, p. 19). Deliveries began to decline in 1964 and reached a low 

of approximately 1,230 MTU in 1975. During the 1970s, DOE predecessor agencies considered 

closing FMPC. The staffing level, which peaked at 2,891 in 1956, slowly declined from 662 in 1972 to 

538 in 1979. Starting in 1981, production output increased to three times the 1979 level and 

employment increased from 538 to more than 1,000 (ASI ca. 1986, p. 4).

The production of uranium metal at FMPC ended in July 1988, and site personnel began to 

concentrate on cleanup efforts. Current operations are limited to long-term surveillance and 

maintenance.

Production area facilities included nine separate plants, the Pilot Plant, ancillary buildings, and 

administrative buildings that were connected by a network of roadways. These facilities, along with 

concrete storage pads, gravel ground cover, railroad access, sanitary landfill, and metal scrap piles, 

were surrounded by security fencing (as are the remaining facilities). The Waste Storage Area 

outside the 136-acre fenced production area included six low-level radioactive waste storage pits, 

two earthen-bermed concrete silos that contained K-65 (high-specific-activity, low-level, radiumbearing) residues, one concrete silo that contained metal oxides, and all affected adjoining areas. 

This area included two fly-ash piles about 3,000 ft south-southwest of the waste storage area and a 

burn pit between Waste Pits 3 and 4. Production operations occurred in Plants 1 to 9 and the Pilot 

Plant.

In 2005, after remedial actions were completed, responsibility for the site was assigned to DOE 

Legacy Management. As a result of remedial actions, occupational environmental dose at the site is 

minimal and is expected to continue to be so in the future.

4.3 SITEWIDE OCCUPATIONAL ENVIRONMENTAL DOSE SUMMARY

Airborne emissions and external gamma radiation data indicate that the radioactive aerosol 

concentration and gamma dose rates have varied significantly from facility to facility and in various 

sections of the site. The location where maximum exposure to thorium occurred was not the same 

location where maximum exposure to uranium occurred. The same is true for radium and radon 

progeny. This difficulty is addressed by dividing FMPC into smaller areas, which are designated 

Exposure Areas (EAs), on a grid that provides a realistic estimated representation of radiological 

conditions where employees might have worked. Figure 4-1 shows the FMPC site plan with the grid. 

Table 4-1 lists major facilities in each EA. Airborne radionuclide intake rates are presented in 

Table 4-2. Section 4.4 describes how Table 4-2 was derived. A spreadsheet (ORAUT 2015) 

documents the various calculations that were used to develop annual exposures and intake rates for 

each source, EA, and year combination. The maximum sitewide annual intakes are assigned in the 

Interactive RadioEpidemiological Program (IREP) as a lognormal distribution with a geometric 

standard deviation of 3.0.
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EA-1 EA-2

EA-4 EA-3

EA-8

EA-5

EA-6

EA-11 EA-7

EA-9

EA-10

Figure 4-1. EAs for 1951 to 1988 as used in this analysis. 

Table 4-1. Facilities and buildings in each EA. 

EA Facilities and buildings

EA-1 Plant 1, Buildings 67 and 68 

EA-2 Plant 9, Buildings 64 and 65

EA-3 Plants 4, 5, 6, and 7 

EA-4 Plant 2/3, Plant 8, and Incinerator Building

EA-5 Waste Pit Area

EA-6 K-65 silos

EA-7 Pilot Plant, Building 69, Laboratory Building Waste Water Treatment Facility, Storm Water 

Retention Basins, Administrative Area (expanded into this EA in the 1990s)

EA-8 Administrative Area, Industrial Relations, and Security Buildings

EA-9 Onsite Disposal Facility and Sewage Treatment Plant

EA-10 Railroad

EA-11 Waste Haul Road

The external gamma dose rate sitewide exposure rates by the EAs are presented in ORAUT (2015) 

and summarized in Table 4-3.

These doses should be assigned based on work location as defined by an EA. If the work location is 

unknown or an individual worked in various locations throughout the site in a given year, the annual 

maximum gamma doses should be assigned for overestimate claims with a constant distribution in 

IREP. For best-estimate claims or those with a probability of causation greater than 50%, the annual 

site average gamma doses should be used and assigned as a normal distribution in IREP with a 

standard deviation of 30%.
 

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3/2015 Page 13 of 65

Table 4-2. Maximum sitewide average (1951 to 1988) or median (1989 to 2012) annual intake rates via inhalation. Rates shown are for all 

except 222Rn. The 222Rn intake rate for 1952 through 1988 is based on the 95th-percentile value of the K-65 silo emission rates. Intake 

rates are based on inhalation and in units of WLM/yr for 222Rn and Bq/yr for all other radionuclides.

a,b 

Year

Total Ud

(Bq/yr)

Th-228 

(Bq/yr)

Th230/Ra226c,

e

(Bq/yr)

Th-232 

(Bq/yr)

Pu-239 

(Bq/yr)

Np-237 

(Bq/yr)

Tc-99 

(Bq/yr)

Ru-103/ 

Ru-106f

(Bq/yr)

Zr/Nb-95g

(Bq/yr)

Sr-90 

(Bq/yr)

Am-241 

(Bq/yr)

Rn-222 

(WLM/yr)

1951 5.87E-01 0.00E+00 0 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 6.11E-01

1952 3.25E+01 0.00E+00 0 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 2.25E+01

1953 1.57E+03 0.00E+00 0 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 3.98E+01

1954 1.14E+03 4.83E+00 0 4.83E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 3.98E+01

1955 2.84E+03 5.52E+00 0 5.52E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 5.73E+01

1956 1.18E+03 5.18E+00 0 5.18E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 5.73E+01

1957 8.35E+02 2.55E-03 0 2.55E-03 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 5.52E+01

1958 9.19E+02 5.83E-03 0 5.83E-03 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 4.77E+01

1959 1.06E+03 5.83E-03 0 5.83E-03 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 3.31E+00

1960 5.56E+03 7.51E-02 0 7.51E-02 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 3.31E+00

1961 1.32E+03 1.46E-01 0 1.46E-01 1.064E+01 4.21E+00 2.59E+02 1.85E+02 5.56E+01 7.41E+00 1.88E+00 3.31E+00

1962 9.39E+02 2.86E-01 0 2.86E-01 7.54E+00 3.00E+00 1.85E+02 1.32E+02 3.96E+01 5.28E+00 1.34E+00 3.31E+00

1963 1.01E+03 3.20E-01 0 3.21E-01 8.07E+00 3.21E+00 1.98E+02 1.41E+02 4.24E+01 5.66E+00 1.44E+00 3.31E+00

1964 7.46E+02 2.46E+00 0 2.46E+00 5.99E+00 2.38E+00 1.47E+02 1.05E+02 3.15E+01 4.20E+00 1.06E+00 3.31E+00

1965 3.34E+02 2.46E+00 0 2.46E+00 2.10E+00 8.36E-01 5.15E+01 3.68E+01 1.10E+01 1.47E+00 5.97E-01 3.31E+00

1966 3.12E+04 3.12E+00 0 3.12E+00 1.96E+02 7.81E+01 4.81E+03 3.44E+03 1.03E+03 1.38E+02 5.57E+01 3.31E+00

1967 2.80E+02 2.46E+00 0 2.46E+00 1.76E+00 7.00E-01 4.31E+01 3.09E+01 9.25E+00 1.23E+00 5.00E-01 3.31E+00

1968 4.92E+02 2.47E+00 0 2.47E+00 3.09E+00 1.23E+00 7.57E+01 5.41E+01 1.62E+01 2.17E+00 8.77E-01 3.31E+00

1969 1.95E+02 1.43E+01 0 1.43E+01 1.23E+00 4.88E-01 3.00E+01 2.15E+01 6.44E+00 8.59E-01 3.48E-01 3.31E+00

1970 1.78E+02 1.05E+01 0 1.05E+01 1.12E+00 4.44E-01 2.74E+01 1.96E+01 5.87E+00 7.83E-01 3.17E-01 3.31E+00

1971 1.07E+02 5.56E+00 0 5.56E+00 6.75E-01 2.68E-01 1.65E+01 1.18E+01 3.54E+00 4.73E-01 1.91E-01 3.31E+00

1972 2.00E+02 1.03E+00 0 1.03E+00 1.24E+00 4.99E-01 3.07E+01 2.20E+01 6.59E+00 8.80E-01 3.56E-01 3.31E+00

1973 2.80E+02 5.69E-01 0 5.69E-01 6.94E+00 2.20E+00 1.51E+02 3.08E+01 9.23E+00 1.23E+00 2.44E+00 3.31E+00

1974 3.12E+02 7.44E-01 0 7.44E-01 7.74E+00 2.45E+00 1.69E+02 3.44E+01 1.03E+01 1.37E+00 2.72E+00 3.31E+00

1975 3.65E+02 5.41E-01 0 5.41E-01 9.06E+00 2.87E+00 1.97E+02 4.02E+01 1.21E+01 1.61E+00 3.19E+00 3.31E+00

1976 3.04E+02 6.63E-01 0 6.63E-01 7.55E+00 2.39E+00 1.65E+02 3.35E+01 1.01E+01 1.34E+00 2.66E+00 3.31E+00

1977 1.31E+02 2.94E+00 0 2.94E+00 3.26E+00 1.03E+00 7.10E+01 1.45E+01 4.34E+00 5.78E-01 1.15E+00 3.31E+00

1978 6.78E+01 2.94E+00 0 2.94E+00 1.68E+00 5.33E-01 3.67E+01 7.47E+00 2.24E+00 2.99E-01 5.92E-01 3.31E+00

1979 1.26E+02 2.94E+00 0 2.94E+00 3.12E+00 9.89E-01 6.80E+01 1.39E+01 4.15E+00 5.54E-01 1.10E+00 3.31E+00

1980 4.82E+01 7.51E-01 0 7.51E-01 1.20E+00 3.79E-01 2.60E+01 5.31E+00 1.59E+00 2.12E-01 4.21E-01 6.42E-01

1981 5.74E+01 7.51E-01 0 7.51E-01 1.42E+00 4.51E-01 3.10E+01 6.32E+00 1.90E+00 2.53E-01 5.01E-01 6.42E-01

1982 4.94E+01 7.51E-01 0 7.51E-01 1.23E+00 3.88E-01 2.67E+01 5.44E+00 1.63E+00 2.18E-01 4.31E-01 3.67E-01

1983 8.31E+01 7.51E-01 0 7.51E-01 2.06E+00 6.54E-01 4.49E+01 9.15E+00 2.74E+00 3.66E-01 7.25E-01 3.67E-01

1984 5.85E+01 7.51E-01 0 7.51E-01 1.45E+00 4.60E-01 3.16E+01 6.44E+00 1.93E+00 2.58E-01 5.11E-01 3.67E-01

1985 4.65E+01 7.51E-01 0 7.51E-01 1.15E+00 3.66E-01 2.51E+01 5.12E+00 1.54E+00 2.05E-01 4.06E-01 3.67E-01

1986 4.74E+01 7.51E-01 0 7.51E-01 1.18E+00 3.73E-01 2.56E+01 5.22E+00 1.57E+00 2.09E-01 4.14E-01 3.67E-01

1987 4.53E+01 7.85E-01 0 7.85E-01 1.12E+00 3.56E-01 2.45E+01 4.98E+00 1.49E+00 1.99E-01 3.95E-01 3.67E-01

1988 4.40E+01 8.14E-01 0 8.14E-01 1.09E+00 3.46E-01 2.38E+01 4.84E+00 1.45E+00 1.94E-01 3.84E-01 2.17E-01
 

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Year

Total Ud

(Bq/yr)

Th-228 

(Bq/yr)

Th230/Ra226c,

e

(Bq/yr)

Th-232 

(Bq/yr)

Pu-239 

(Bq/yr)

Np-237 

(Bq/yr)

Tc-99 

(Bq/yr)

Ru-103/ 

Ru-106f

(Bq/yr)

Zr/Nb-95g

(Bq/yr)

Sr-90 

(Bq/yr)

Am-241 

(Bq/yr)

Rn-222 

(WLM/yr)

1989 6.71E+01 5.08E-05 0 7.26E-05 1.67E+00 5.28E-01 3.63E+01 7.39E+00 2.22E+00 2.96E-01 5.86E-01 9.64E-02

1990 2.38E+01 1.80E-05 0 2.58E-05 5.91E-01 1.87E-01 1.29E+01 2.63E+00 7.87E-01 1.05E-01 2.08E-01 9.64E-02

1991 3.06E+01 2.32E-05 0 3.31E-05 7.60E-01 2.41E-01 1.66E+01 3.38E+00 1.01E+00 1.35E-01 2.68E-01 1.05E-01

1992 3.99E+01 3.02E-05 0 4.31E-05 9.90E-01 3.14E-01 2.16E+01 4.39E+00 1.32E+00 1.76E-01 3.48E-01 7.99E-02

1993 4.45E+01 3.37E-05 0 4.81E-05 1.10E+00 3.50E-01 2.41E+01 4.90E+00 1.47E+00 1.96E-01 3.89E-01 1.25E-01

1994 1.56E+01 1.18E-05 0 1.68E-05 3.86E-01 1.22E-01 8.41E+00 1.71E+00 5.14E-01 6.86E-02 1.36E-01 1.29E-01

1995 1.12E+01 8.47E-06 0 1.21E-05 2.78E-01 8.80E-02 6.05E+00 1.23E+00 3.70E-01 4.93E-02 9.77E-02 1.21E-01

1996 2.53E+01 1.91E-05 0 2.73E-05 6.28E-01 1.99E-01 1.37E+01 2.79E+00 8.35E-01 1.11E-01 2.21E-01 1.58E-01

1997 4.99E+01 3.77E-05 0 5.39E-05 1.24E+00 3.92E-01 2.70E+01 5.49E+00 1.65E+00 2.20E-01 4.35E-01 1.70E-01

1998 3.70E+01 2.80E-05 0 4.00E-05 9.17E-01 2.91E-01 2.00E+01 4.07E+00 1.22E+00 1.63E-01 3.23E-01 1.13E-01

1999 5.35E+01 4.05E-05 1.22E+00 5.79E-05 1.33E+00 4.21E-01 2.89E+01 5.89E+00 1.77E+00 2.36E-01 4.67E-01 1.21E-01

2000 4.82E+01 3.64E-05 9.24E+00 5.21E-05 1.19E+00 3.79E-01 2.60E+01 5.30E+00 1.59E+00 2.12E-01 4.20E-01 1.05E-01

2001 4.82E+01 3.64E-05 3.60E+01 5.21E-05 1.19E+00 3.79E-01 2.60E+01 5.30E+00 1.59E+00 2.12E-01 4.20E-01 1.05E-01

2002 1.35E+01 1.02E-05 2.82E+01 1.46E-05 3.34E-01 1.06E-01 7.28E+00 1.48E+00 4.45E-01 5.94E-02 1.18E-01 1.21E-01

2003 1.12E+02 8.47E-05 1.02E+01 1.21E-04 2.78E+00 8.80E-01 6.05E+01 1.23E+01 3.70E+00 4.93E-01 9.77E-01 1.05E-01

2004 6.32E+02 4.79E-04 2.04E+01 6.84E-04 1.57E+01 4.97E+00 3.42E+02 6.97E+01 2.09E+01 2.79E+00 5.52E+00 1.05E-01

2005 1.70E+01 1.29E-05 3.79E+00 1.84E-05 4.22E-01 1.34E-01 9.20E+00 1.88E+00 5.62E-01 7.50E-02 1.49E-01 1.00E-01

2006 3.19E+00 2.41E-06 0 3.44E-06 7.91E-02 2.51E-02 1.72E+00 3.51E-01 1.05E-01 1.40E-02 2.78E-02 1.13E-01

2007 1.26E+00 9.57E-07 0 1.37E-06 3.14E-02 9.95E-03 6.84E-01 1.39E-01 4.18E-02 5.57E-03 1.10E-02 9.64E-02

2008 8.76E-01 6.63E-07 0 9.47E-07 2.17E-02 6.89E-03 4.73E-01 9.64E-02 2.89E-02 3.86E-03 7.64E-03 1.13E-01

2009–

presenth 6.81E-01 5.15E-07 0 7.36E-07 1.69E-02 5.36E-03 3.68E-01 7.50E-02 2.25E-02 3.00E-03 5.95E-03 1.09E-01

a. The material type (F, M, and S) to be used for dose reconstruction should be that which is favorable to the claimant and consistent with the specific claim and 

guidance in ICRP Publication 68 (ICRP 1995).

b. Total uranium is interpreted as U-234 for dose reconstruction purposes.

c. Intake rates for Th-230 and Ra-226 are equal. 

d. Three short-duration episodic emissions (Table 4-6) were modeled using Equation 4-7 and averaged into the annual intake rates.

e. These intake rates are associated with the remediation of waste pits that occurred from 1999 through 2005. Intake rates for Th-230 and Ra-226 for other years are 

negligible. 

f. Ru-103/106 is assumed to be 100% Ru-106, due to its longer half-life.

g. Zr/Nb-95 pair is assumed to be in equilibrium, with equal activities totaling the intake rate indicated in the table.

h. Results for 2010 to present based on 2009, the last year of monitoring.
 

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Table 4-3. External gamma radiation (mrem/yr).a,b,c

Year EA-1 EA-2 EA-3 EA-4 EA-5 EA-6 EA-7 EA-8 EA-9 EA-10 EA-11

Site 

average

Standard 

deviation Max.

1952 241.2 169.3 169.2 241.1 14.7 34.9 12.6 20.7 15.4 13.5 9.8 85.7 97.7 241

1953 294.4 206.5 206.4 294.1 19.6 60.0 15.4 25.3 18.8 16.5 12.0 106.3 118.2 294

1954 294.6 206.6 206.5 294.3 21.7 82.3 15.5 25.4 18.9 16.6 12.1 108.6 117.4 295

1955 490.8 344.2 344.1 490.3 34.0 114.8 25.8 42.2 31.4 27.6 20.1 178.7 196.4 491

1956 980.6 688.1 687.9 980.1 61.3 140.5 51.2 84.3 62.7 55.0 39.7 348.3 397.0 981

1957 980.3 688.0 687.8 980.0 56.8 140.0 51.2 84.2 62.6 54.9 39.7 347.8 397.3 980

1958 980.3 688.0 687.8 980.0 57.2 140.0 51.2 84.2 62.7 55.0 39.7 347.8 397.3 980

1959 980.4 688.0 687.9 980.0 58.3 140.1 51.2 84.2 62.7 55.0 39.7 348.0 397.2 980

1960 980.6 688.0 687.9 980.1 60.5 140.4 51.2 84.2 62.7 55.0 39.7 348.2 397.1 981

1961 980.7 688.1 687.9 980.1 62.5 140.6 51.2 84.3 62.7 55.1 39.7 348.4 396.9 981

1962 980.8 688.1 688.0 980.2 64.5 140.8 51.3 84.3 62.7 55.1 39.8 348.7 396.8 981

1963 980.9 688.2 688.0 980.2 66.5 141.0 51.3 84.3 62.7 55.1 39.8 348.9 396.7 981

1964 981.4 688.4 688.1 980.5 69.4 202.0 51.5 84.4 62.8 55.2 40.1 354.9 393.8 981

1965 981.9 688.6 688.2 980.8 72.0 263.0 51.7 84.5 62.9 55.3 40.3 360.8 391.7 982

1966 982.4 688.8 688.4 981.1 74.7 324.0 51.9 84.6 62.9 55.4 40.6 366.8 390.5 982

1967 982.9 689.0 688.5 981.4 77.1 384.9 52.1 84.7 63.0 55.5 40.9 372.7 390.2 983

1968 983.2 689.1 688.6 981.6 76.6 445.6 52.2 84.8 63.0 55.5 41.1 378.3 390.9 983

1969 983.7 689.3 688.7 981.9 79.1 506.5 52.4 84.9 63.1 55.6 41.4 384.2 392.3 984

1970 984.2 689.5 688.8 982.2 81.6 567.5 52.6 85.0 63.2 55.7 41.6 390.2 394.5 984

1971 847.5 593.5 592.6 845.4 76.9 621.3 45.7 73.4 54.5 48.1 36.4 348.7 347.2 848

1972 710.9 497.4 496.5 708.5 72.4 675.2 38.7 61.7 45.8 40.5 31.2 307.2 305.7 711

1973 564.4 394.5 393.5 561.9 67.3 728.5 31.3 49.2 36.5 32.4 25.6 262.3 270.1 729

1974 427.8 298.4 297.3 425.1 62.7 782.3 24.4 37.6 27.8 24.8 20.3 220.8 250.0 782

1975 310.7 216.1 214.9 307.8 59.1 837.2 18.5 27.6 20.4 18.3 15.9 186.0 247.0 837

1976 302.8 214.3 213.0 299.7 73.5 910.0 30.5 38.8 32.0 30.2 28.3 197.6 261.2 910

1977 351.4 247.9 247.1 349.0 68.5 558.1 29.0 39.6 31.9 29.6 25.7 179.8 182.6 558

1978 330.8 233.3 232.3 328.0 72.1 733.4 28.9 38.6 31.2 29.1 26.1 189.4 218.8 733

1979 443.3 310.1 308.0 438.5 86.5 1,615.0 34.4 46.5 36.2 33.2 31.3 307.5 465.0 1,615

1980 419.0 293.6 290.8 413.0 99.9 2,145.6 39.0 49.2 39.3 36.6 37.0 351.2 615.0 2,146

1981 740.5 515.5 512.8 734.4 106.5 2,150.8 44.1 65.6 48.4 43.0 38.1 454.5 631.9 2,151

1982 754.1 526.4 524.3 749.2 101.7 1,626.1 45.0 67.8 50.5 45.1 38.2 411.7 498.5 1,626

1983 766.2 534.5 532.2 760.9 106.4 1,802.3 45.8 68.7 51.1 45.6 39.1 432.1 543.5 1,802

1984 361.0 253.9 252.1 356.7 91.5 1,281.9 33.2 42.9 34.7 32.4 30.8 251.9 366.8 1,282

1985 294.3 208.6 206.8 289.9 94.9 1,318.8 35.0 42.2 35.5 33.9 33.4 235.8 374.2 1,319

1986 516.1 361.2 359.7 512.2 92.0 1,091.1 35.1 50.6 38.8 35.4 30.5 283.9 332.9 1,091

1987 60.6 46.1 44.8 57.0 79.9 923.3 25.1 24.9 23.6 23.8 25.9 121.4 266.6 923
 

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Year EA-1 EA-2 EA-3 EA-4 EA-5 EA-6 EA-7 EA-8 EA-9 EA-10 EA-11

Site 

average

Standard 

deviation Max.

1988 252.2 181.0 179.0 247.3 103.4 1,517.4 40.8 45.9 40.2 39.0 39.5 244.2 430.7 1,517

1989 196.6 141.4 139.7 192.3 93.1 1,247.7 34.0 37.9 33.4 32.6 32.9 198.3 354.1 1,248

1990 87.4 62.7 61.3 83.7 75.7 988.4 20.0 20.9 18.8 18.7 20.2 132.5 285.2 988

1991 279.9 195.9 194.3 289.8 81.6 545.7 24.5 31.9 25.4 23.8 22.4 155.9 167.0 546

1992 184.7 132.2 131.9 145.4 65.6 66.8 21.8 27.4 23.5 22.6 19.4 76.5 61.0 185

1993 269.4 190.0 189.6 277.1 66.3 73.2 21.6 30.1 24.2 22.7 18.3 107.5 103.4 277

1994 323.0 227.2 226.7 325.2 68.4 74.7 23.2 33.5 26.4 24.4 19.3 124.7 124.7 325

1995 306.0 215.7 215.3 334.0 69.1 111.6 23.9 33.6 26.9 25.0 20.2 125.6 120.6 334

1996 21.2 18.9 18.5 19.4 61.5 180.0 18.6 18.3 18.1 18.6 18.8 37.4 49.0 180

1997 21.3 18.8 18.4 19.4 62.0 224.0 18.5 18.2 18.0 18.4 18.7 41.4 61.9 224

1998 19.8 17.3 16.8 17.8 60.7 236.4 16.9 16.6 16.4 16.8 17.2 41.1 66.1 236

1999 19.7 17.2 16.8 17.8 60.5 226.4 16.9 16.6 16.4 16.8 17.2 40.2 63.1 226

2000 20.0 17.3 16.8 18.0 61.2 268.4 17.0 16.7 16.4 16.9 17.3 44.2 75.5 268

2001 20.9 18.2 17.7 18.9 62.4 295.2 17.9 17.5 17.2 17.7 18.2 47.4 83.2 295

2002 21.8 19.0 18.5 19.7 63.3 300.0 18.7 18.3 18.0 18.5 19.0 48.6 84.4 300

2003 18.2 16.1 15.8 16.6 58.0 121.2 15.8 15.7 15.5 15.9 15.9 29.5 32.9 121

2004 18.7 16.8 16.5 17.1 57.9 69.8 16.5 16.4 16.2 16.7 16.5 25.4 19.2 70

2005 26.1 23.8 23.3 24.3 66.6 205.4 23.4 23.2 22.9 23.4 23.6 44.2 55.0 205

2006 26.1 23.8 23.3 24.3 66.6 205.4 23.4 23.2 22.9 23.4 23.6 44.2 55.0 205

2007 26.1 23.8 23.3 24.3 66.6 205.4 23.4 23.2 22.9 23.4 23.6 44.2 55.0 205

2008 26.1 23.8 23.3 24.3 66.6 205.4 23.4 23.2 22.9 23.4 23.6 44.2 55.0 205

2009–

presentd

26.1 23.8 23.3 24.3 66.6 205.4 23.4 23.2 22.9 23.4 23.6 44.2 55.0 205

a. External exposures include ambient doses because background doses were not available for all of the dose components.

b. Dose reconstructors should use ambient (H*10) dose conversion factors (DCFs) for 1952 to 1976 (based on instrument readings) and exposure (R) DCFs for 1977 to 

the present (based on readings from a TLD on a fence line) (NIOSH 2007a, Section 4.1.2). 

c. Ambient gamma radiation doses are based on 2,000 hr/yr.

d. Results for 2010 to the present based on 2009, the last year of monitoring.
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4.4 INTERNAL DOSES FROM ATMOSPHERIC RADIONUCLIDE CONCENTRATIONS

This section describes the derivation of radionuclide concentrations in the general areas for each year 

for the unmonitored worker dose reconstruction effort using available documented technical 

information and data from the system of air monitoring stations (AMSs). Table 4-4 indicates the types 

of information and data available from the beginning of operations at FMPC.

Table 4-4. Historical information and data available for environmental dose reconstruction. 

Period

Available historic information and data 

description Source document

1951–1988 Airborne uranium, thorium, and radon emissions 

including unmonitored and episodic releases.

Radiation Doses and Risk to Residents 

from FMPC Operations from 1951–1988, 

Volumes I and II (RAC 1998a, 1998b).

History of FEMP Radionuclide 

Discharges (Boback et al. 1987)

Addendum to FMPC-2082, “History of 

FEMP Radionuclide Discharges” (Dolan 

and Hill 1988)

The Fernald Dosimetry Reconstruction 

Project, Tasks 2 and 3, Radionuclide 

Source Terms and Uncertainties

(Voillequé et al. 1995). 

1989–2005 Estimated airborne emissions for FMPC for 

radionuclides other than radon reported in the 

annual site environmental reports.

Environmental reports for 1989 to 2005

(Dugan et al. 1990; Byrne et al. 1991; 

WEMCO 1992; FERMCO 1993 to 1996; 

FDF 1997 to 1999; Fluor Fernald 2000 to 

2006.)

1989–1996 Radon in ambient air from continuous, highvolume AMSs at the site fence line locations. 

Results for onsite locations not reported until 

1997.

Environmental reports for 1989 to 1996

(Dugan et al. 1990; Byrne et al. 1991; 

WEMCO 1992; FERMCO 1993 to 1996; 

FDF 1997)

1997–2005 Radon in ambient air data from the Radon 

Monitoring Program that included AMSs at onsite 

locations.

Environmental reports for 1997 to 2005

(FDF 1998, 1999; Fluor Fernald 2000 to 

2006.)

2006–

presenta

Ambient particulate and radon air monitoring data 

collected by DOE Legacy Management. 

Particulate through 2009 and radon through 2008.

Environmental reports for 2006 to 2012

(Stoller 2007; DOE 2008 to 2013)

a. For dose reconstruction purposes, the last year that ambient air was monitored and reported at the site was 2009.

It is necessary to develop appropriate methodologies for determining the atmospheric radionuclide 

concentrations specific to the type of information available for the various periods in Table 4-4. 

Therefore, this TBD presents the derivation of the internal occupational environmental dose in 

separate periods as listed in the first column of Table 4-4.

4.4.1 RADIOACTIVE MATERIALS AIR CONCENTRATIONS AND INTAKES, 1951 TO 1988

4.4.1.1 RADIONUCLIDE DISCHARGES AND SOURCES OF EMISSIONS

During FMPC production operations, approximately 308,000 kg of uranium and an estimated 

14,500 kg of thorium were released to the atmosphere from various emission sources during routine 

operations. History of FEMP Radionuclide Discharges (Boback et al. 1987) and The Fernald 

Dosimetry Reconstruction Project, Tasks 2 and 3, Radionuclide Source Terms and Uncertainties

(Voillequé et al. 1995) provide the most recent evaluations of radioactive emissions from the FMPC 

operating years (1951 through 1988). Dolan and Hill (1988), an addendum to Boback et al. (1987), 

provides the initial assessments of the emissions from site operations. For the purpose of 

environmental dose reconstruction, this TBD uses the larger of the emissions amounts from either 
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Voillequé et al. (1995) or Boback et al. (1987). The values for routine and nonroutine uranium 

discharges from operations and episodic releases during the operating years are from Voillequé et al

(1995). The thorium discharges are from Dolan and Hill (1988). The values for releases of 222Rn and 

its progeny are from Voillequé et al. (1995) for the K-65 silo releases and drummed K-65 materials at 

the Plant 1 Pad. Radon releases from the material in Q-11 silos were derived from Pinney et al. 

(2004).

Table 4-5 lists annual airborne uranium emissions due to FMPC operations throughout the operating 

history (1951 through 1988) from each production plant and processing facility from Voillequé et al. 

(1995, p. 643).

The values in Table 4-5 include:

• Monitored source emissions from dust collectors,

• Emissions from Plant 2/3 UO3 gulping,

• Emissions from Plant 8 scrubbers,

• Building exhaust emissions due to normal operating conditions,

• Fugitive emissions from waste pits,

• Nonroutine uranium fires and UF6 gas releases,

• Incinerator and burner emissions, and

• Six episodic releases of uranium documented in incident reports.

Table 4-6 reproduces Table K-43 of Voillequé et al. (1995, p. 643) to show a summary of six episodic 

releases of uranium that were identified from the incident reports and air monitoring.

Table 4-7 summarizes annual fugitive thorium emissions due to wind erosion from four of the six 

waste pits in the waste storage area (Dolan and Hill 1988). Only Waste Pits 2, 3, 4, and 5 released 

thorium. Data on uranium and radon emissions from these waste pits are also published in Dolan and 

Hill (1988). The uranium data are not used in this analysis because the values from this source are 

lower than the values Voillequé et al. (1995) subsequently estimated (Table 4-5). Radon effluents 

from the waste pits are not included because they contributed a negligible fraction of the total site 

annual radon emission rate, which in many years was in the kilocurie range.

Table 4-8 lists annual airborne emissions from thorium operations at Plants 8 and 9 and the Pilot 

Plant from 1951 through 1988 from Dolan and Hill (1988).

RAC (1998b) provides estimates of the annual emission of 222Rn from the K-65 silos for the period 

from 1952 through 1988. The 95th-percentile values of these estimates, which define upper 

estimates of emission rates, have been used in this analysis. Median estimates of the emission rates 

(Voillequé et al. 1995, p. 435) were applied to materials stored at Plant 1. Those materials stored at 

Plant 1 were eventually placed in the K-65 silos.

Pinney et al. (2004) identified another large radon emission source, the Q-11 ore silos near Plant 2/3, 

which provided a relatively larger exposure to workers in its vicinity from 1952 to 1958 when the silos 

were filled with highly radioactive ore. Hornung et al. (2008, p. 9) also provided information on this 

topic. Interviews with former employees revealed that six smaller silos in the production area had 

been used in the 1950s to store high-grade ore from the African Metals Corporation before 

processing. This ore had high 226Ra content, which decays to 222Rn. Although the amount of material 

was considerably less than that stored in the K-65 silos, the Q-11 silos were much closer to the 

production area and were not sealed.
 

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Table 4-5. Uranium emissions summary (kg) (Voillequé et al. 1995).

Year Plant 1 Plant 2/3a Plant 4 Plant 5 Plant 6b Plant 7c Plant 8d Plant 9e

Pilot

Plantf

Building

exhaustg

Waste

pits

UF6 

releaseh Year total

1951 0.0 0.0 0.00.0 0.0 2.9 0.0 0.0 0.0 11 0.0 0.0 0.0 14

1952 0.0 0.0 0.0 0.0 19 0.0 0.0 0.0 300 16 0.0 44 379

1953 5.5 245 2,500 2,000 62 0 245 0 500 16 0.0 44 5,617

1954 60 1,547 7,000 7,000 121 5,500 3,192 0 520 171 0.0 44 25,155

1955 59 3,429 14,000 14,000 153 25,000 6,015 0 800 236 0.0 44 63,736

1956 55 4,311 6,400 4,500 174 4,200 8,082 0 35 206 0.0 44 28,007

1957 80 5,911 1,200 4,800 174 0.0 6,062 265 20 256 125 0.0 18,893

1958 300 5,360 700 850 217 0.0 7,531 712 30 291 230 44 16,265

1959 55 5,534 2,000 600 164 0.0 9,019 458 40 326 230 45 18,470

1960 35 5,601 220 250 346 0.0 9,864 254 1,470 381 285 45 18,750

1961 45 4,898 550 150 174 0.0 12,568 81 120 366 340 45 19,337

1962 40 2,096 1,000 490 105 0.0 9,157 163 130 323 345 45 13,894

1963 85 2.0 2,000 900 217 0.0 10,447 183 55 306 320 45 14,560

1964 28 193 650 400 38 0.0 7,794 275 15 308 310 45 10,056

1965 5.2 685 370 300 62 0.0 2,134 71 11 282 165 45 4,130

1966 21 1,196 300 90 16 0.0 1,073 50 21 167 125 45 3,104

1967 30 1,831 300 180 6.2 0.0 1,521 81 15 181 20 44 4,210

1968 0.8 1,927 300 110 46 0.0 3,481 122 4.0 146 20 0.0 6,157

1969 45 1,057 53 130 5.3 0.0 1,260 13 4.0 92 20 0.0 2,679

1970 7.0 660 42 89 8.8 0.0 1,268 15 1.0 108 20 0.0 2,219

1971 15 1,251 1.3 1.0 5.0 0.0 276 1.0 1.0 23 20 0.0 1,594

1972 70 2,719 20 45 4.0 0.0 100 25 1.0 23 20 0.0 3,027

1973 2.0 3,860 75 102 4.0 0.0 120 15 1.0 24 20 0.0 4,223

1974 1.0 4,344 30 40 4.0 0.0 146 41 1.0 25 20 0.0 4,652

1975 7.3 4,733 150 26 4.0 0.0 468 1.0 1.0 31 23 0.0 5,445

1976 3.0 4,114 30 17 5.3 0.0 276 3.1 1.0 20 24 0.0 4,493

1977 0.8 1,227 16 75 4.0 0.0 624 1.0 12 16 24 0.0 2,000

1978 2.0 2.0 16 40 4.0 0.0 544 79 2.0 11 24 0.0 724

1979 1.0 2.0 100 17 4.0 0.0 1,077 2.0 1.0 10 25 0.0 1,239

1980 18 14 200 130 4.0 0.0 125 1.0 2.2 16 25 0.0 535

1981 1.0 56 530 160 4.0 0.0 153 1.0 1.0 21 25 0.0 952

1982 2.0 89 27 140 5.0 0.0 279 5.1 1.0 25 25 0.0 599

1983 8.0 172 60 60 6.5 0.0 628 1.0 1.0 28 25 0.0 990

1984 13 505 42 100 4.0 0.0 204 204 3.2 23 25 0.0 1,123

1985 2.0 48 55 40 3.9 0.0 160 18 0.6 23 25 1.5 378

1986 1.0 19 32 20 3.3 0.0 266 9.2e 0.3 20 25 7.5 403

1987 0.0 65 20 15 3.2 0.0 153 6.1 0.2 2.6 25 1.5 292

1988 0.0 32e 0.6 0 3.0 0.0 133 0 1.0 0.5 25e 1.5 197

Total 1,104 69,736 40,990 37,867 2,184 34,700 106,444 3,158 4,134 4,519 2,980 680 308,495

a. Emissions from Plant 2/3 include dust collectors, scrubbers, and unmonitored releases.

b. Emissions from Plant 6 include dust collectors, unmonitored releases, and cooling tower releases due to their proximity to Plant 6.

c. Emissions from Plant 7 include dust collectors and episodic releases.
 

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d. Emissions from Plant 8 include dust collectors, scrubbers, and unknown episodic and incinerator releases due to their proximity to Plant 8.

e. Emissions from Plant 9 include dust collectors and unmonitored releases.

f. Emissions from the Pilot Plant include dust collectors and episodic releases.

g. Particulate emissions from the building exhaust and nonroutine fire releases are combined and assumed to be released at the center of the FMPC operational areas. 

Plant 4 is near the center of the operational area, so the Plant 4 stack location is assumed, however, the release height is assumed to be ground level.

h. Emissions from nonroutine UF6 releases are assumed to be ground-level gaseous releases from the location of the Pilot Plant.
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Table 4-6. Summary of six episodic releases of uranium from incident reports and air monitoring

(Voillequé et al. 1995, p. 643).

Event Start Description

Uranium 

release (kg)

Duration of 

release

1 11/07/1953 Release of UF6 from defective cylinder in Pilot Plant 45 15 min

2 11/12/1960 Dust loss from dust collector bags in Pilot Plant 310 1 or 3 d

3 02/14/1966 Release of UF6 from cylinder in Pilot Plant due to 

operator error 750 1 hr

4 09/21/1978 Unknown 20–370a 7 d

5 02/01/1979 Unknown 60–680a 7 d

6 09/13/1983 Unknown 290–360a 7 d

a. Range of values is based on results from several ambient AMSs.

Table 4-7. Fugitive thorium emissions (kg) from wind erosion of Waste 

Pits 2, 3, 4, and 5 (Dolan and Hill 1988). 

Year Pit 2 Pit 3 Pit 4 Pit 5 Total

1957 0.014 0.0 0.0 0.0 0.014

1958 0.032 0.0 0.0 0.0 0.032

1959 0.032 0.0 0.0 0.0 0.032

1960 0.032 0.0 0.38 0.0 0.412

1961 0.036 0.0 0.76 0.0 0.796

1962 0.036 0.0 1.53 0.0 1.566

1963 0.036 0.0 1.72 0.0 1.756

1964 0.036 0.01 1.91 0.0 1.956

1965 0.018 0.014 1.91 0.0 1.942

1966 0.01 0.014 2.1 0.0 2.124

1967 0.0 0.0 2.1 0.0 2.1

1968 0.0 0.0 2.3 0.0 2.3

1969 0.0 0.005 2.3 0.0 2.305

1970 0.0 0.01 2.3 0.0 2.31

1971 0.0 0.01 2.3 0.0 2.31

1972 0.0 0.005 2.49 0.0 2.495

1973 0.0 0.005 2.49 0.0 2.495

1974 0.0 0.0 2.49 0.1 2.59

1975 0.0 0.01 2.68 0.24 2.93

1976 0.0 0.01 2.87 0.76 3.64

1977 0.0 0.01 3.25 0.29 3.55

1978 0.0 0.01 3.64 0.29 3.94

1979 0.0 0.01 3.83 0.29 4.13

1980 0.0 0.005 3.83 0.29 4.125

1981 0.0 0.005 3.83 0.29 4.125

1982 0.0 0.0 3.83 0.29 4.12

1983 0.0 0.0 3.83 0.29 4.12

1984 0.0 0.0 3.83 0.29 4.12

1985 0.0 0.0 3.83 0.29 4.12

1986 0.0 0.0 3.83 0.29 4.12

1987 0.0 0.0 3.83 0.48 4.31

1988 0.0 0.0 3.83 0.64 4.47

Total 0.282 0.133 79.8 5.1 85.4
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Table 4-8. Thorium emissions data by year (kg) (Dolan and Hill 1988).

Year Plant 8 Plant 9 Pilot Plant Yearly total

1953 0 0 0 0

1954 0 1,028 0 1,028

1955 0 1,176 0 1,176

1956 0 0 0 0

1957 0 0 0 0

1958 0 0 0 0

1959 0 0 0 0

1960 0 0 0 0

1961 0 0 0 0

1962 0 0 0 0

1963 0 0 0 0

1964 0 0 344 344

1965 0 0 344 344

1966 118 0 344 462

1967 0 0 344 344

1968 0 0 344 344

1969 1,040 0 394 1,434

1970 699 0 499 1,198

1971 430 0 62 492

1972 0 0 141 141

1973 0 0 50 50

1974 0 0 100 100

1975 0 0 3 3

1976 0 0 0 0

1977 0 0 408 408

1978 0 0 408 408

1979 0 0 408 408

1980 0 0 0 0

1981 0 0 0 0

1982 0 0 0 0

1983 0 0 0 0

1984 0 0 0 0

1985 0 0 0 0

1986 0 0 0 0

1987 0 0 0 0

1988 0 0 0 0

Total 2,287 2,204 4,193 8,684

Figure 13 in Pinney et al. (2004) reported the yearly mean and maximum exposure in working level 

months (WLMs) to workers due to the Q-11 emissions during the years when these silos served as 

radon sources (1952 to 1958). Figure 13 from Pinney et al. (2004) is reproduced in Figure A-1 in 

Attachment A. The yearly radon emissions from the Q-11 silos were estimated based upon a 

rearrangement of the atmospheric dilution factor (χ/Q) equation and the maximum exposures (in 

WLM) from Figure 13. A detailed description of the treatment of adding the Q-11 silos to the overall 

radon release source terms for those years the silos were active is provided in Attachment A.

Each of these radon emission sources are combined and summarized in Table 4-9. Waste pits were 

also a source of radon at FMPC, but the released amounts were negligible in comparison with those 

in Table 4-9. To ensure the radon intake rate is not underestimated, the estimated naturally occurring 

background radon levels are included in this evaluation. An ambient background radon concentration 

of 0.47 pCi/L is assumed. This value is based on the average of the measured background radon 

concentrations from 1989 to 2000 (Environmental Report Fluor Fernald 2001). Attachment C shows 

background concentration measurements.
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Table 4-9. Releases (Bq) of 222Rn used in this analysis for 1952 through 

1988 (RAC 1998b; Voillequé et al. 1995, p. 435; Pinney et al. 2004). 

Year K-65 silos Q-11 silo

Plant 1 pad, 

drum storage Annual total

1951 0.00E+00 0.00E+00 1.30E+12 1.30E+12

1952 1.96E+14 3.05E+14 2.15E+13 5.23E+14

1953 1.96E+14 6.37E+14 3.52E+12 8.37E+14

1954 3.70E+14 6.31E+14 0.00E+00 1.00E+15

1955 3.70E+14 9.15E+14 0.00E+00 1.29E+15

1956 3.70E+14 9.15E+14 0.00E+00 1.29E+15

1957 3.70E+14 8.81E+14 0.00E+00 1.25E+15

1958 3.70E+14 7.59E+14 0.00E+00 1.13E+15

1959 4.81E+14 0.00E+00 0.00E+00 4.81E+14

1960 4.81E+14 0.00E+00 0.00E+00 4.81E+14

1961 4.81E+14 0.00E+00 0.00E+00 4.81E+14

1962 4.81E+14 0.00E+00 0.00E+00 4.81E+14

1963 4.81E+14 0.00E+00 0.00E+00 4.81E+14

1964 4.81E+14 0.00E+00 0.00E+00 4.81E+14

1965 4.81E+14 0.00E+00 0.00E+00 4.81E+14

1966 4.81E+14 0.00E+00 0.00E+00 4.81E+14

1967 4.81E+14 0.00E+00 0.00E+00 4.81E+14

1968 4.81E+14 0.00E+00 0.00E+00 4.81E+14

1969 4.81E+14 0.00E+00 0.00E+00 4.81E+14

1970 4.81E+14 0.00E+00 0.00E+00 4.81E+14

1971 4.81E+14 0.00E+00 0.00E+00 4.81E+14

1972 4.81E+14 0.00E+00 0.00E+00 4.81E+14

1973 4.81E+14 0.00E+00 0.00E+00 4.81E+14

1974 4.81E+14 0.00E+00 0.00E+00 4.81E+14

1975 4.81E+14 0.00E+00 0.00E+00 4.81E+14

1976 4.81E+14 0.00E+00 0.00E+00 4.81E+14

1977 4.81E+14 0.00E+00 0.00E+00 4.81E+14

1978 4.81E+14 0.00E+00 0.00E+00 4.81E+14

1979 4.81E+14 0.00E+00 0.00E+00 4.81E+14

1980 8.88E+13 0.00E+00 0.00E+00 8.88E+13

1981 8.88E+13 0.00E+00 0.00E+00 8.88E+13

1982 8.88E+13 0.00E+00 0.00E+00 8.88E+13

1983 8.88E+13 0.00E+00 0.00E+00 8.88E+13

1984 8.88E+13 0.00E+00 0.00E+00 8.88E+13

1985 8.88E+13 0.00E+00 0.00E+00 8.88E+13

1986 8.88E+13 0.00E+00 0.00E+00 8.88E+13

1987 8.88E+13 0.00E+00 0.00E+00 8.88E+13

1988 4.81E+13 0.00E+00 0.00E+00 4.81E+13

4.4.1.2 EXPOSURE SITE LOCATIONS

Airborne emissions data indicate that the amount of normal releases at FMPC has varied significantly 

from facility to facility. This TBD analysis assumed that the radioactive aerosol concentration varied 

significantly in various sections of the site. The dose a worker received is highly dependent on the 

amount of time spent in specific work areas. To provide supportive information for dose 

reconstruction, this TBD has divided FMPC into smaller areas (designated as EAs) on a grid that 

provides a realistic estimated representation of radiological conditions where employees might have 

worked. Figure 4-1 and Table 4-1 above show the FMPC site plan with the grid and the major 

facilities in each EA, respectively. This TBD presents airborne radionuclide intakes from operations 

for the 11 EAs. These calculations are documented in ORAUT (2015).
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4.4.1.3 AIRBORNE RADIONUCLIDE CONCENTRATION DERIVATION

The radioactive aerosol concentrations at various distances from a continuous source in each EA can 

be obtained by:

EA i i

i di

χ C R f (RF ) Q

  =     ∑ (4-1)

where

CEA = radionuclide concentration in EA (g/m3

) 

Ri = radionuclide emission rate from source i (g/s) where i can be Plant 1, 2/3, 4, etc.

(χ/Q)di = atmospheric dilution factor at distance d from source i (s/m3

) 

fi = fraction of time that the wind blew toward the EA from source i 

RF = respirable fraction of the source aerosol [<10 μm (DOE 1994b)]

Airborne radionuclide concentrations for each EA were calculated as the sum of the contributions of 

all emission sources to the most densely occupied area in the EA. In EAs that contained multiple 

emission sources (i.e., EA-3 and EA-4), the radionuclide concentrations were determined at the 

vicinity of the highest emission source in the EA.

4.4.1.3.1 RADIONUCLIDE EMISSION RATES, Ri × RF

The average annualized emission rates were derived by dividing emission amounts in Tables 4-5, 4-8, 

and 4-9 by 3.156 × 107 seconds (an average year). In 1951 and 1953, 222Rn and radon progeny were 

released from stored drums of K-65 materials at the Plant 1 pad over 3 months in one case and 6 

months in the other (Table 4-9). However, their release rates were annualized as if the total emission 

amount were released over a 1-year period.

For internal dose assessment, the particle sizes of interest are those below 10 μm, which are 

respirable and sufficiently small to remain airborne during transport. Spenceley (1985) presents 

16 datasets on emission stack particle size distributions for 15 different stacks in three different plants 

(Plants 4, 5, and 9). The activity median aerodynamic diameter (AMAD) and the 1-sigma uncertainty 

were reported for each determination. The range of the measured AMAD values was 0.72 to 9.8 μm. 

The stack with the lowest value was apparently resampled with a result of 9.0 μm. Figure 4-2 shows 

the measured particle size distributions from the stacks. The average value of the AMAD was 6.6 μm.

The measurements were made using a multi-stage particle size separator, so a cumulative 

distribution of the “percent greater than stated size” was reported. The scale of the “stated size” was 

not consistent with each determination, but a value of about 10 μm (the nominal cutoff value 

associated with respirable particles) was available in each case. Typically (in 13 out of 16 instances) 

the cutoff value was greater than 11 μm, so the percentage of respirable particles tends to be slightly 

overestimated, which results in a bias that is favorable to the claimant for dose reconstruction 

purposes. Figure 4-3 shows the percentage of the emissions from the various stacks that were less 

than nominally 10 μm. The average of the values represented in Figure 4-3 is 74%.

Appendix E of The Fernald Dosimetry Reconstruction Project (Voillequé et al. 1995, p. 291) indicates 

that, in general, the size distribution of emissions from uranium processes involved particle sizes 

below 10 μm in the 60th to 70th percentile, weighted toward the 60th percentile. In particular, 

Table E-8 of Voillequé et al. (1995, p. 291) indicates that the composite of U3O8 releases from the 

dust collectors that showed more than 75% of the emission had a particle size less than 10 μm. 

Therefore, for particulate emissions, a value of 75% was used for RF in Equation 4-1. For gaseous 

releases – including UF6, radon, and radon progeny – a value of 100% was used.
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Figure 4-2. Measured emission stack particle size from 15 different FMPC 

stacks (Spenceley 1985).

Figure 4-3. Percentage of stack emission less than approximately 10 μm 

AMAD (Spenceley 1985).

4.4.1.3.2 FAR-FIELD ATMOSPHERIC DILUTION FACTOR, Χ/Q

The atmospheric dilution factor χ/Q can be calculated using the Pasquill-Gifford equation for the 

dispersion of gas from a continuous source (Cember 1983, p. 347):

2 2

2 2

2 2

1 1

y y 2

χ(x,y ) y H exp Q πσ σ u σ σ

   

= −+          

 (4-2)

where

χ(x,y) = ground-level concentration at point x,y (g/m3

) 

x = downwind distance on plume centerline (m)
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y = crosswind distance (m) (in all cases y = 0)

Q = emission rate (g/s)

σy, σz = horizontal and vertical standard deviations of contaminant concentration in the 

plume (m), Figures 4-4 and 4-5, respectively

u = mean wind speed at level of plume centerline (m/s)

H = effective release height (m)

For particulates, the χ/Q equation becomes:

2 2

2 2

2 2

(, ) 1 1 exp 2 2 y y

χxy y H

Q πσ σ u σ σ

   

= −+          

 (4-3)

Equation 4-2 yields the ground-level concentration of a gas that is continuously emitted from a point 

source and is based on total reflection of the gas by the ground and an elevated mixing lid (inversion) 

that keeps the contaminant within the bounds of the ground and lid. If the pollutant in the plume is 

retained on the ground, as in the case of particulates (Equation 4-3), the ground-level concentration is 

approximately half of that in Equation 4-2. Note that these equations do not assume depletion of the 

plume by such mechanisms as dry or wet deposition or gravitational settling, which are known to 

occur and would act to reduce plume concentrations.

Instead of manually interpreting the values of σy and σz from Figures 4-4 and 4-5, an alternate method 

(Napier and Ramsdell 2005) can be applied. This method results in the same values for σy and σz

when the limitations of graphical interpretation are considered.

Figure 4-4. Horizontal diffusion σy versus downwind distance from point source 

for Pasquill’s stability categories (Slade 1968, p. 115).
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4.4.1.3.3 RECEPTOR DOWNWIND DISTANCES

The horizontal and vertical standard deviations of radionuclide concentration in the plume (σy, and σz) 

are functions of the downwind receptor distances from the emission sources. Table 4-10 lists the 

EA receptor downwind distance estimates from each contributing emission source that were used for 

radioactivity concentration calculations. The estimated distances were based on the line-of-sight from 

the emission sources to the closest occupied location in each EA. These distances are the shortest 

distance between the emission sources and the occupied locations.

Figure 4-5. Vertical diffusion σz versus downwind distance from continuously 

emitting point source for Pasquill’s stability categories (Slade 1968, p. 116).

Table 4-10. Receptor downwind estimated distances from contributing emission sources (m).

EA Plant 1

Plant 

2/3 Plant 4 Plant 5 Plant 6 Plant 7 Plant 8 Plant 9

Pilot 

Plant

Waste 

pits

K-65 

silos

1 50 150 150 250 400 250 250 350 350 450 500

2 300 250 150 150 150 200 300 25 500 800 900

3a 250 150 50 100 100 100 150 150 300 750 750

4a 100 100 100 200 300 100 50 350 150 550 550

5 550 350 550 650 800 550 450 800 500 100 400

6 450 550 700 850 950 700 550 1,000 450 300 50

7 350 200 250 300 400 150 100 550 50 700 500

8 450 250 200 200 150 100 200 400 300 950 800

9 800 700 500 400 250 550 700 250 900 1,000 1,500

10 400 600 550 600 600 650 650 450 800 250 700

11 600 450 750 800 1,000 650 500 1,000 300 500 200

a. The receptor was located at the vicinity of the highest emission source.

For EAs that contained multiple emission sources (i.e., EA-3 and EA-4), the receptor locations were 

placed at the vicinity (within 50 m) of the highest emission source. Pasquill’s stability category F was 

used for the σy and σz determination because it is the most favorable to claimants. In accordance with 
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Figures 4-4 and 4 5, the F stability category represents a moderately stable condition that has the 

lowest σy and σz, which results in a higher material concentration downwind from the point of release. 

Ground releases and elevated releases were assumed as appropriate to the configuration of the 

emission sources. The elevated releases assumed an average stack height of 10 m from Table 2 of 

Boback et al. (1987, p. 54), which is also favorable to claimants because Table 2 shows that the 

overwhelming majority of the FMPC stacks have been well over 30 m above ground level (not 

accounting for stack exit velocity or plume buoyancy that in both cases would increase the effective 

stack height). The effects of building wake were taken into account in estimating close-in 

concentrations (below 200 m) from an elevated release. In 1963, lateral support berms were added to 

K-65 silos. As a consequence, the K-65 release height is estimated to be 10 m through 1962 and 0 m 

beginning in 1963. The method for modifying χ/Q to account for building wake effects is presented in 

Section 4.4.1.3.5. Because uranium and thorium emissions from the production plants were 

particulates that were released through the stacks, an elevated release χ/Q was used for uranium and 

thorium releases. A ground-level release χ/Q was used for uranium and thorium fugitive emissions 

from the waste pits. Radon-222 was treated as a ground-level gaseous release primarily from the 

headspace of the K-65 and Q-11 silos and from drummed material storage near Plant 1.

4.4.1.3.4 EFFECTS OF WIND DIRECTION, Fi

Wind direction plays a key role in predicting the amount of aerosol that arrived at an EA. In 

Equation 4-2, the mean wind speed u at the level of the plume centerline is dependent on the general 

wind direction and expressed as wind rose data. The wind rose data include the percentage of time 

that the wind blew in a particular direction. The site environmental reports for 1991 to 2001 (WEMCO 

1992; FERMCO 1993 to 1996; FDF 1997 to 1999; Fluor Fernald 2000 to 2002) provided FMPC 

meteorological conditions for climatic condition evaluations and atmospheric modeling. Only 9 years 

of the data (1991 to 1997 and 2000 to 2001) were sufficiently legible for use. This TBD used the 

95th-percentile values of this wind rose data (Figure 4-6) to determine the wind direction frequency for 

each of the compass points. Table 4-11 presents the summary wind rose data that were used in this 

analysis. Attachment B contains more detail.

Figure 4-6. 95th-percentile wind rose data, 10-m height, 1991 to 1997 

and 2000 to 2001 (FDF 1998, p. 35; Fluor Fernald 2002, p. 28). 
Document No. ORAUT-TKBS-0017-4 Revision No. 03 Effective Date: 12/03/2015 Page 29 of 65

 

Table 4-11. 95th-percentile values from wind rose data

Wind from the ... Frequency Speed (m/s)

N 0.0456 2.70

NNE 0.0468 2.80

NE 0.058 2.60

ENE 0.083 2.60

E 0.0472 1.77

ESE 0.028 1.42

SE 0.028 1.67

SSE 0.0426 2.12

S 0.0756 2.62

SSW 0.125 2.94

SW 0.1296 2.24

WSW 0.113 2.24

W 0.1156 2.36

WNW 0.103 2.41

NW 0.0892 2.20

NNW 0.0712 2.41

4.4.1.3.5 CONCENTRATION FIELDS NEAR BUILDINGS, MODIFIED Χ/Q FOR BUILDING WAKE

When the assumed height of the release is within 2.5 times the height of the surrounding structures, 

estimation of the aerosol concentrations near buildings needs to consider the effect of the building 

wake that entrains some of the effluent and brings it to ground level. The building wake effect can be 

accounted for by calculating the modified diffusion factors σy' and σz' using the method in Randerson 

(1984). The modified expressions for σy' and σz' for an elevated release are:

σ . (W / ) . (x H ) y b ′ = +− 0 7 2 0 0967 3 (4-4)

0.7 0.067( 3 ) σ H xH z b b ′ =+ − (4-5)

and

2

2

1 1

exp 2

e

y z z

χ h

Q πσ σ u σ

    =   −   ′ ′ ′      

 (4-6)

where

Hb = building height (m)

W = building width (m)

x = distance to the receptor (m)

he = effective release height (m)

u = mean wind speed (m/s)

Information on building dimensions (i.e., width and roof peak height) from Boback et al. (1987, p. 54) 

and from Figure 1 of ORAUT (2011, p. 15) for the K-65 silos were used for the near-building χ/Q

calculations. Table 4-12 lists the building or structure dimensions that were used to solve 

Equations 4-4 and 4-5.
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Table 4-12. Building and K-65 silo dimensions for building 

wake χ/Q calculations. 

Structures

Width

ft (m)

Roof peak height

ft (m)

Plant 1 100 (30.5) 60 (18.3)

Plant 2/3 60 (18.3) 67 (20.4)

Plant 4 165 (50.3) 94 (28.7)

Plant 5 100 (30.5) 52 (15.8)

Plant 6 200 (61.0) 39 (11.9)

Plant 7 80 (24.4) 114 (34.7)

Plant 8 60 (18.3) 48 (14.6)

Plant 9 225 (68.6) 40 (12.2)

Pilot Plant 210 (64.0) 54 (16.5)

K-65 silos 80 (24.4) 36 (11.0)

4.4.1.3.6 EPISODIC RELEASES

Table 4-6 summarizes six significant episodic releases of uranium that were identified from incident 

reports and air monitoring data with a total median release estimate of 1,700 kg of uranium. For 

comparison purposes, the total estimated uranium emissions to the atmosphere from nonroutine 

events between 1952 and 1988 was 2,784 kg (Voillequé et al. 1995, p. 40), which is 2% of the 

approximately 135,000 kg of uranium that was previously reported by Boback et al. (1987) for 1951 

through 1984 (Voillequé et al. 1995, p. 616). Releases that had occurred in a period of less than a 

day (i.e., Events 1, 2, and 3) were evaluated as quasi-instantaneous (puff) releases. The remaining 

episodic releases (i.e., Events 4, 5, and 6), which occurred over about a week, were treated as 

continuous releases. The exposures from the puff releases were considered as additions to the 

annual environmental doses, which also include all of the episodic uranium releases from Table 4-6.

The exposures from these events would have been received during the months when the releases 

occurred, but for dose reconstruction are averaged into the annual intake rate.

The atmospheric diffusion equation for puff release (Slade 1968) is:

2 2

2 2

1 1

exp 2 yl zl yl zl

ψ y h

Q πσ σ u σ σ

   

= −+          

(4-7)

where

ψ = consequence or exposure to a receptor expressed in terms of the integrated 

concentration (g/m3

)

Q = emission rate (g/s)

σyI, σzI = horizontal and vertical standard deviations of puff concentration (m)

u = mean wind speed at level of plume centerline (m/s)

y = crosswind distance (m) (in all cases y = 0)

h = effective release height (m)
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For particulates, multiply Equation 4-7 by 0.5. Slade (1968, p. 419) presented equations to calculate 

values of σyI and σzI as a function of downwind distance x in meters. To apply these equations, this 

analysis assumed that the default Class F atmospheric stability condition corresponds to “very stable” 

conditions. The value of σyI is calculated as:

0.93 σyl = × 0.14 x (4-8)

where

x = downwind distance (m)

The value of σzI is calculated as:

0.61 0.05 σ x zl = × (4-9)

The equation for the radionuclide intakes using ψ/Q can be written:

ψ l q f Br RF Q

  = × ×× ×    

 (4-10)

where

I = radionuclide intake (g)

q = episodic release amount (g)

(ψ/Q) = atmospheric dilution factor for puff release (s/m3

) (see Attachment C)

f = fraction of time that the wind blew toward the EA (Figure 4-7)

Br = breathing rate (m3

/s)

RF = respirable fraction of the source aerosol (<10 μm)

4.4.1.4 RADIONUCLIDE INTAKE DERIVATION

The intake calculation assumed a breathing rate of 2,400 m3

/yr. The isotopic content in the uranium 

emissions varied based on the type and form of uranium that was processed in various plants and 

during different periods.

For purposes of dose reconstruction, the assumption was made that, through 1964, all uranium 

processed at FMPC was enriched in 235U to a level of 1%. After 1964, all uranium was assumed to be 

2% enrichment (NIOSH 2007b, p. 48). Table 4-13 lists the primary assumptions for FMPC uranium 

enrichments and the isotopes that are associated with these enrichments.

Table 4-13. Uranium enrichment isotopes (ORAUT 2004, p.10).

Isotope Uranium enrichment

Specific activity 

(Bq/mg U) Activity fraction

U-234 1% enriched 29 0.637

U-235 1% enriched 29 0.021

U-238 1% enriched 29 0.342

U-234 2% enriched 37 0.769

U-235 2% enriched 37 0.0268

U-238 2% enriched 37 0.204

ORAUT-RPRT-0052, Feed Materials Production Center Internal Dose Topics (ORAUT 2011), 

documented the resolution of several internal dosimetry issues including some of the bases for 

recycled uranium (RU) default values in Table 4-14. RU was first processed in significant amounts at 
Document No. ORAUT-TKBS-0017-4 Revision No. 03 Effective Date: 12/03/2015 Page 32 of 65

 

FMPC beginning in 1961, so none of the RU-associated contaminants were present before 1961. 

After Revision 02 of this document, NIOSH determined that the concentration values that were 

assigned to plutonium would be higher in some periods and that 241Am, a decay product of one of the 

isotopes of plutonium in the weapons-grade plutonium mixture, was also present as a contaminant in 

RU. Table 4-14 presents the RU contaminants and their concentrations. Plutonium values have been 

adjusted upward at the second decimal place. 

Table 4-14. RU contaminant radionuclides (Bq/g U) (Carbaugh 2003 pg. 183 and ORAUT 2011). 

Period Pu-239a Np-237 Tc-99 Th-232 Th-228

Ru-103/ 

Ru-106b Zr/Nb-95c Sr-90 Am-241

1961–1964 2.327+02 9.25E+01 5.70E+03 4.00E–02 2.80E–02 4.08E+03 1.22E+03 1.63E+02 4.14E+01

1965–1972 2.327E+02 9.25E+01 5.70E+03 4.00E–02 2.80E–02 4.08E+03 1.22E+03 1.63E+02 6.60E+01

1973–

present 9.18E+02 2.91E+02 2.00E+04 4.00E–02 2.80E–02 4.08E+03 1.22E+03 1.63E+02 3.23E+02

a. Dose reconstructors (DRs) should assume that the plutonium mixture is 100% Pu-239, as an assumption that is 

favorable to claimants. 

b. Ru-103/106 is assumed to be 100% Ru-106, due to its longer half-life. 

c. Zr/Nb-95 pair is assumed to be in equilibrium, with equal activities totaling the intake rate indicated in the table.

The assumptions used for plutonium contamination in RU are as follows: 

• Before 1961: No plutonium contamination existed in the uranium.

• 1961 to 1964: Total alpha emitting plutonium isotopes are assumed to exist in RU at a mass 

concentration 100 ppb (i.e., 0.1 μg Pu/g U). All of the plutonium isotopes are represented for 

dosimetric purposes as 239Pu. The age of plutonium mixture (i.e., the time for 241Am in-growth

since chemical purification occurred) is 10 years. 

• 1965 to 1972: Total alpha emitting plutonium isotopes are assumed to exist in the mixture at 

100 ppb. The age of plutonium mixture is 20 years. 

• 1973 to present: Total alpha emitting plutonium isotopes are assumed to exist in the mixture 

at 400 ppb. The age of plutonium mixture is 30 years.

Plant 2/3, the Refinery, was operated to convert the U.S. government stockpile of natural uranium ore 

concentrates to UO3. Ore process operation in Plant 2/3 started in 1953 and continued until 1962 when 

UO3 production was transferred to the Mallinckrodt Chemical Company at the Weldon Spring Plant. In 

1966, the UO3 refinery at the Weldon Spring Plant was shut down; the refinery at Fernald was 

reactivated in April 1966 and continued operations through 1977 when it was placed in standby (Mead, 

Savage, and Fugate ca. 1985, pp. 20–22). In 1988, the refinery operated intermittently in a sequential 

campaign to convert enriched recycled materials to oxide. The natural uranium ore concentrates (often 

called yellowcake) include the decay products of uranium, 226Ra, and 230Th. To a large extent these 

impurities were removed during the milling process, but some of the decay products remained in the ore 

concentrates. These decay products could have been emitted from Plant 2/3 and therefore could have 

contributed nontrivial environmental doses. This TBD evaluates the environmental impacts of these 

uranium decay products in the Plant 2/3 uranium emissions from 1953 through 1988. A wide range of 

ratios for 226Ra and 230Th to uranium existed in Plant 2/3 depending on the origins of the uranium ore. 

This evaluation assumed none of the decay products were removed in the milling process and that the 

decay products were in equilibrium with the activity of natural uranium in the ore. This assumption is 

favorable to claimants. Based on this assumption, Table 4-15 provides the ratios of 226Ra and 230Th 

activity to uranium mass. The uranium effluent in kilograms from Plant 2/3 was multiplied by these 

ratios to estimate the activity of 226Ra and 230Th in the Plant 2/3 effluent. The Table 4-15 information is 

reflected in Table 4-2.
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Table 4-15. Ratio of 226Ra and 230Th to uranium. 

Radioisotopes Progeny/U ratio

Ra-226 3.47E-01 Bq Ra-226/kg U

Th-230 3.47E-01 Bq Th-230/kg U

a. See ORAUT (2015) for derivation.

4.4.1.4.1 RADON

Radon-222 intake levels are expressed in units of WLM. For 222Rn, 1 working level (WL) is equal to 

100 pCi/L of 222Rn in 100% equilibrium with its short-lived decay products. A month is defined as a 

period of 170 work hours. The derivation of the number of WLMs in a year was based on an intake 

period of 2,000 hours for the year and an assumed environmental radon progeny equilibrium factor of 

70%. This ratio for ambient outside air is in accordance with widespread sampling throughout the 

United States from National Council on Radiation Protection and Measurements Report 78 

(NCRP 1984).

Therefore,

2

0.7 2000 pCi

100 170 L

pCi 8.24 10

L

222Rn WLM in a year Concentration

Concentration −

×   = ×   ×  

 

= ××    

(4-11)

The average concentration of radon emissions from the waste pits (EA-5) was typically an order of 

magnitude lower than those from the K-65 silos. Therefore, the radon exposure results for the EAs do 

not include the radon emissions from the waste pits.

4.4.1.4.2 RADON EXPOSURE RECORDS

Some claimant files might contain radon exposure records in units of WLMs. These records typically 

appear in “Personnel Exposure Records,” separate from the “DOE Response” file. DRs should follow 

the guidance in DCAS-TIB-011, Lung Dose Conversion Factor for Thoron WLM, when evaluating 

these records (NIOSH 2013a). For dose reconstructions that require assignment of 222Rn exposures, 

and when individual radon exposure data exist, both the environmental 222Rn and the individual 222Rn 

data need to be evaluated. However, DRs should assign only the highest dose determined from this 

comparison, not both. The radon intakes from the Personnel Exposure Records should be assumed 

to be a chronic exposure and should be assigned as a constant distribution in IREP for overestimates. 

For best-estimate or compensable claims, DRs must prorate the radon doses to actual periods of 

employment.

Based on information from Pinney et al. (2004), the total radon exposure is the summation of the 

radon exposure from the K-65 silos and Q-11 silos. Worker radon data were obtained from two files, 

NIOSH (ca. 2013b) and NIOSH (ca. 2013c). The total K-65 radon exposure is equivalent to the sum 

of the individual radon results from (NIOSH ca. 2013b) in the YRLY_K65 field to the results in the 

K65_TOTAL field of the second file (NIOSH ca. 2013c). The total Q-11 radon exposure is equivalent 

to the sum of the individual radon results from (NIOSH ca. 2013b) in the YRLY_Q11 field to the 

results in the Q11_TOTAL field of the second file (NIOSH ca. 2013b). The fields in NIOSH (ca. 

2013b) are YR, YRLY_Q11, YRLY_K65, TOTAL_EXP, SURVEY_YRLY_Q11, SURVEY_YRLY_K65, 

SURVEY_TOTAL_EXP, and SSN. The fields in NIOSH (ca. 2013c) are SSN, K65_TOTAL, 
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Q11_TOTAL, and TOTAL_EXP. The alphabetical listing of fields and their content is shown in 

Table 4-16, which provides information to DRs on what the fields represent.

4.4.1.5 RESULTS

ORAUT (2015) documents each calculation and lists estimated radionuclide intake rates in 

becquerels per year or WLM per year for unmonitored workers who spent a year in an EA between 

1951 and 1988. These values are summarized and, for each radionuclide of interest, the intake rate 

in the EA with the highest value is assigned for sitewide application. Table 4-2 presents these results 

in summary form.

Table 4-16. Alphabetical list of fields (Pinney et al. 2004, p. 67). 

Field Content

SSN Social Security Number

SURVEY_TOT_EXP Yearly K-65 + Q-11 radon exposure (WLM) from survey data

SURVEY_YRLY_K65 Yearly K-65 radon exposure (WLM) from survey data

SURVEY_YRLY_Q11 Yearly Q-11 radon exposure (WLM) from survey data

TOTAL_EXP Yearly K-65 + Q-11 radon exposure (WLM) from estimated shift

YR Start date year

YRLY_K65 Yearly K-65 radon exposure (WLM) from estimated shift

YRLY_Q11 Yearly Q-11 radon exposure (WLM) from estimated shift

4.4.2 NONRADON RADIONUCLIDE AIR CONCENTRATIONS AND INTAKES, 1989 TO 2012

Starting in 1989, the operational and effluent data needed to calculate environmental intake rates in 

each Exposure Area were no longer available. Instead air sample data at fence line and on-site 

locations published in the annual environmental reports were used as input to the intake rate 

calculations. Citations to the individual annual reports are organized by year in the reference section 

of this document. When full data sets were published, the maximum air sample result for each 

sample location was tabulated and the median of each of these maximum values was assumed as the 

concentration for that year. When full data sets were not available, the maximum published value was 

assumed. No ambient offsite background air concentration was subtracted. The annual airborne 

uranium concentrations from the onsite and perimeter AMSs were used with the model in ORAUTPROC-0031, Site Profile and Technical Basis Document Development (ORAUT 2012), to determine 

the onsite air concentrations and intakes for environmental dose to workers for 1989 to 2012. 

Recycled uranium contaminant levels shown in Table 4-14 were also applied. The date range for this 

section suggests that the period of time of analyses of data occurred through 2012. The empirical 

uranium air sampling data from the environmental reports only extend through 2009 because DOE 

ended the collection programs because the data from previous years indicated levels had reached de 

minimis levels. For years after 2008 (for uranium) and 2009 (for radon) when no air monitoring 

occurred, the data from the last year that monitoring occurred were used to extend the intakes to 

2012.

4.4.2.1 ANNUAL AIR EMISSIONS FOR RADIONUCLIDES OTHER THAN RADON

The uranium air emissions in the FMPC environmental reports contain measured air concentrations 

from onsite and perimeter AMSs. The air concentrations are attributed to uranium losses from:

• Monitored stacks,

• Unmonitored stacks,

• Scrubbers,

• Building vents, and

• Fugitive emissions from the waste pits.
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4.4.2.1.1 MONITORED STACKS

Stack discharges were monitored by taking continuous isokinetic particulate samples, which were 

routinely analyzed for uranium and other radionuclides.

4.4.2.1.2 UNMONITORED STACKS

Facilities that operated on a limited basis were unmonitored. This included losses from the cooling 

towers. Estimates were derived by engineering calculations, periodic stack emission measurements, 

or a combination of both. In general, a detailed analysis was made of the processes, the 

characteristics of emitted materials, and the operating time for each system. This information was 

used to derive emission factors that, when multiplied by amounts of material processed over a period 

of time, provided estimates for emissions of unmonitored processes.

4.4.2.1.3 SCRUBBERS

A scrubber is a pollution control device that uses a liquid to remove contaminants from air. Losses 

were based on an emission factor (per hour) times the actual number of hours these sources 

operated. The emission factors were derived from stack emission tests that were conducted in 

accordance with Environmental Protection Agency methods.

4.4.2.1.4 BUILDING VENTS

Building ventilation emissions were calculated based on data from continuous air monitors in each 

plant near processes that were thought to produce radionuclide emissions. The yearly emissions for 

each plant were estimated from the average annual concentrations from the plant continuous air 

monitors, the estimated hours each fan operated, and a 0.1 dilution factor. Because the emissions 

were measured only in the areas of highest suspected concentrations, the actual average uranium 

concentration in the air in the plants would have been significantly lower than the calculated average 

of the continuous air monitor results.

4.4.2.1.5 FUGITIVE EMISSIONS

Estimates of fugitive uranium were based on the average number of samples that were processed per 

year in each hood, the probable loss per sample, and the estimated uranium concentration of the 

sample.

Waste pits 1 through 6, the Clearwell, and the Burn Pit are described in ORAUT-TBKS-0017-2. 

These structures were remediated over the period 1999 through 2005. Maximum fence line 230Th air 

concentrations observed during each of those years (Fluor-Fernald, 2000, 2001, 2002, 2003, 2004, 

2005, 2006) were used to calculate the annual intake rate assuming full-time work year occupancy. 

Results are presented in Table 4-2.

Table 4-17 presents the median of maximum values of all uranium air samples that were collected 

after 1988. It also show the maximum fence line concentrations of 230Th measured during waste pit 

remediation activities.
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Table 4-17. Medians of the maximum values of all uranium air concentration 

(Bq/m3

) samples by year (ORAUT 2015) and maximum fence line values of 230Th 

during waste pit remediation activities. 

Year Uranium air concentration Th-230 Emissions from Waste Pits

1989 2.03E–02 NA

1990 1.09E–02 NA

1991 1.70E–02 NA

1992 9.19E–03 NA

1993 9.46E–03 NA

1994 5.27E–03 NA

1995 8.11E–05 NA

1996 1.14E–02 NA

1997 1.30E–02 NA

1998 1.18E–02 NA

1999 1.58E–02 6.76E-04

2000 1.53E–02 5.14E-03

2001 1.41E–02 2.00E-02

2002 7.49E–03 1.57E-02

2003 3.16E–02 5.68E-03

2004 1.77E–01 1.14E-02

2005 5.04E–03 2.11E-03

2006 1.76E–03 NA

2007 6.76E–04 NA

2008 4.46E–04 NA

2009–present 3.65E–04a NA

a. Results for 2010 to present results based on 2009 air sampling data when 

particulate uranium site sampling ended.

4.4.2.2 RESULTS

Intake rates for uranium and intakes of associated trace contaminants and 230Th are reported in Table 

4-2. Calculation of the intake rates is presented in ORAUT (2015).

4.4.3 RADON AIR CONCENTRATIONS AND INTAKE ESTIMATES, 1989 TO 2012 

Radon was monitored as a separate component of the air pathway at FMPC. FMPC used alpha-track 

radon detectors in weatherproof housings to determine radon concentrations in the environment 

(Environmental Report Dugan et al. 1990, p. 87). From 1989 to 1998, there were alpha-track radon 

detectors at each of the locations. Alpha-track monitoring ended in 1998. An alpha-track radon 

detector is a passive, long-term device for integrating radon concentrations in air by permanently 

recording the tracks of alpha particles from radon and its decay products. The detectors were 

changed quarterly and sent to the supplier for analysis. At the end of the year, the averages of the 

four quarterly concentrations were computed and documented in the site’s annual environmental 

reports. Beginning in 1999, the radon monitoring continued through the radon continuous monitoring 

program using alpha scintillation monitoring cells. These cells allowed the air to continuously diffuse 

into the detector cell through a foam barrier, without the aid of a pump, and radon and its progeny 

would alpha-decay in the cell and be measured by the scintillator. The system was used until radon 

monitoring at the site was discontinued in 2008. These data were analyzed using the method in 

ORAUT-PROC-0031 (ORAUT 2012) to develop intakes to workers, which are presented in Table 4-2.
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The monitoring locations changed over time as remedial actions occurred. Figures 4-7 and 4-8 show 

monitoring locations in the 1994 timeframe for 11 locations off site, 21 locations along the fence line,

four locations on site at various distances from the silos, and 16 locations immediately adjacent to the 

K-65 silos.

Table 4-18 presents the median of maximum values of all on-site and fenceline radon air samples that 

were collected after 1988 through 2008. The data in the table from 2009 to 2012 are an extension of 

the data from 2008. These data were used to estimate the radon and radon progeny intake rate, 

which are reported in Table 4-2.

4.5 ONSITE EXTERNAL DOSES FROM AMBIENT RADIATION

External environmental radiation dose at FMPC has primarily been the result of photon radiation 

(i.e., gamma and X-rays from radionuclides that were stored on the site). For unmonitored personnel 

who worked in the administration areas, the largest source of direct radiation was the waste in the 

K-65 silos, thorium residues in various warehouses throughout the site, and the radionuclides in the 

Production Plants and other onsite facilities. To a lesser extent, the radioactive Waste Pit Area was 

another direct radiation source for unmonitored personnel who worked in that area.

4.5.1 K-65 SILOS AND PRODUCTION PLANTS EXTERNAL DOSE CONTRIBUTION

There were three significant external radiation sources at FMPC, but by far the largest was the waste 

in the K-65 silos. The ambient dose rates started to increase to measurable levels in 1955 with the 

relocation of wastes from Niagara Falls, New York, to the silos. This elevation in ambient dose was 

due to the radium content of these wastes. Direct radiation occurred not only from radium but also 

from the radon progeny that were in secular equilibrium within the headspace of each of the silos. 

This condition prevailed until the silos became such a significant source of radiation that DOE found it 

necessary to construct reinforcement berms and coat the exterior surfaces in an attempt to strengthen 

the structures and reduce the radon emissions. While these actions provided some benefit, the radon 

dose contribution continued to increase as gaseous radionuclides emanated and accumulated in the 

void space under the domes and became an increasing source of external exposure both on and off 
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Figure 4-7. Offsite and fence line radon monitoring locations (Environmental Report FERMCO 

1995, p. 171).

AMS-1
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Figure 4-8. Onsite radon monitoring locations (Environmental Report FERMCO 1995, p. 172).

AMS-8a

AMS-9b
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Table 4-18. Medians of the maximum values of all on-site 

and fence line radon air concentrations (Bq/m3

) samples by 

year (ORAUT 2015). 

Year Radon air concentration

1989 2.03E+04

1990 1.76E+04

1991 2.16E+04

1992 2.16E+04

1993 2.70E+04

1994 2.70E+04

1995 2.43E+04

1996 3.78E+04

1997 4.32E+04

1998 5.41E+03

1999 2.70E+04

2000 2.16E+04

2001 2.16E+04

2002 2.70E+04

2003 2.16E+04

2004 2.16E+04

2005 1.89E+04

2006 2.30E+04

2007 1.89E+04

2008–present 2.30E+04a

a. Results for 2009 to present results based on 2008 air sampling data 

when radon site sampling ended.

the site. The source strength increased to the point that radiation levels required additional shielding 

to mitigate the direct radiation dose to personnel. In 1991, a bentonite layer was added to the surface 

of the stored material in an attempt to contain the radioactive gases (Environmental Report Fluor 

Fernald 2001, p. 13). This was effective, as indicated by an immediate reduction in direct radiation 

dose rates due to the shielding by the solid materials in the silos. An immediate dose rate reduction 

of approximately 1 order of magnitude occurred as a result of the bentonite. However, the bentonite 

provided only a temporary reduction. Dose rates slowly increased again and monitoring in 2002 

showed an increase of about a factor of 4 in external radiation.

The only data found from before 1975 (Klein 1963) consisted of contact radiation readings using a 

Juno survey meter for measurements of exposure rates in the vicinity of the North and South Tanks 

(i.e., the K-65 silos). The dose rates ranged from 10 to 20 mR/hr and 10 to 30 mR/hr, respectively. 

Later data (Boback 1978) provided readings from thermoluminescent dosimeters (TLDs) inside the 

South Tank that ranged from 590 mR/hr at the surface of the residue to 650 mR/hr at 10 ft and 

470 mR/hr at 13 ft above the surface of the residue. These data are indicative of the magnitude of the 

K-65 silos as sources of ambient radiation, and they were used to project the radiation environment of 

the site in its early operating years (1952 to 1975). ORAUT (2015) documents the details of this 

calculation.

Starting in 1976, gamma radiation at the site boundary (fence line) sampling stations was measured 

with TLDs that were changed and processed every 3 months (Environmental Report Boback et al.

1977, p. 62). Table 4-19 is a summary of the maximum measured dose rates at the monitoring 

stations for 1976 through 1995. Figure 4-9 shows the locations of the fence line monitoring stations.

In 1996, FMPC expanded its direct radiation monitoring system to continuously measure 32 locations, 

including several onsite monitoring locations. Table 4-20 provides summary level information about 
Document No. ORAUT-TKBS-0017-4 Revision No. 03 Effective Date: 12/03/2015 Page 41 of 65

 

direct radiation measurements for 1996 through 2005. Figure 4-10 shows the locations of the TLD 

stations.
 

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3/2015 Page 42 of 65

Table 4-19. Radiation dose rates at site boundary (mR/hr or mrem/hr).

a,b

Maximum measured dose at TLD stations

Year BS1/1A BS2 BS3 BS4 BS5 BS6 BS7 BS8 BS9 Onsite Background

1976 0.010 0.012 0.012 0.011 0.012 0.015 NA NA NA NA 0.010

1977 0.011 0.010 0.011 0.011 0.011 0.012 NA NA NA NA 0.009

1978 0.010 0.011 0.011 0.010 0.011 0.013 NA NA NA NA 0.009

1979 0.012 0.012 0.010 0.010 0.011 0.018 NA NA NA NA 0.009

1980 0.012 0.014 0.014 0.013 0.012 0.023 NA NA NA NA 0.011

1981 0.010 0.012 0.012 0.012 0.012 0.021 0.013 NA NA NA 0.009

1982 0.011 0.014 0.013 0.013 0.014 0.019 0.013 NA NA NA 0.010

1983 0.012 0.013 0.013 0.013 0.013 0.020 0.014 NA NA NA 0.010

1984 0.010 0.0113 0.0114 0.011 0.0115 0.0171 0.0123 NA NA NA 0.010

1985 0.0123 0.0129 0.0125 0.0122 0.0127 0.0191 0.0138 NA NA NA 0.0118

1986 0.0127 0.011 0.0111 0.0111 0.0116 0.015 0.0104 NA NA NA 0.009

1987 0.0096 0.0107 0.0103 0.0104 0.010 0.0157 0.0108 NA NA NA 0.0106

1988 0.0144 0.0164 0.0169 0.0155 0.015 0.0235 0.0158 0.0148 0.0186 NA 0.0151

1989 0.0135 0.0131 0.0161 0.0126 0.0127 0.0197 0.0132 0.0116 0.0175 NA 0.0128

1990 0.0079 0.0079 0.0076 0.0078 0.0073 0.013 0.0075 0.0074 0.010 NA 0.0075

1991 0.0081 0.0085 0.0077 0.0082 0.0077 0.013 0.008 0.0078 0.010 0.258 0.0069

1992 0.0080 0.0094 0.009 0.009 0.0091 0.0087 0.0083 0.0084 0.0115 0.0205 0.0082

1993 0.0089 0.0084 0.0077 0.0078 0.0077 0.0077 0.0075 0.0079 0.0107 0.0239 0.0068

1994 0.0128 0.0082 0.0071 0.0074 0.0081 0.0079 0.0072 0.0079 0.0107 0.0239 0.0068

1995 0.0157 0.0085 0.0075 0.0074 0.0079 0.0085 0.0077 0.0077 0.0108 0.0420 0.0074

a. Sources: 1976 to 1995 environmental reports (Boback et al. 1977, 1978; Boback and Ross 1980, 1981; Fleming et al. 1982 to 1984; Facemire et al. 1985; Aas et al.

1986, 1987; Gels and Lojeck 1988; Gels et al. 1989; Dugan et al. 1990; Byrne et al. 1991; WEMCO 1992; FERMCO 1993 to 1996). 

b. NA = not available.
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Figure 4-9. Fence line TLD locations, 1976 to 1995 (Environmental Reports FERMCO 1993 to 1996).

Tables 4-19 and 4-20 list the external radiation information for 1976 to 2005 with source information. 

The collection of these TLD measurement data became possible with advances in the development of 

the TLD as a useful dose measurement device. These technology developments made it possible to 

measure radiation doses over long periods, in the exposure environment, and without power. TLDs 

became a popular and accurate tool for measuring doses in the environment over extended periods.
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Figure 4-10. TLD monitoring locations, 1996 to 2005 (Environmental Report Fluor Fernald 2004).

Table 4-20. TLD measurement summary (mrem/hr).

a,b

Year

Fence line 

min.

Fence 

line max.

Onsite min. 

(Health and 

Safety 

Bldg.)

Onsite max. 

(K-65 silos 

fence line)

Offsite 

min.

Offsite 

max.

Background 

average

1996 0.008 0.0104 0.0069 0.0789 0.0069 0.0084 0.0088

1997 0.0082 0.0103 0.0070 0.101 0.0068 0.0085 0.0087

1998 0.0072 0.0096 0.0063 0.108 NA NA 0.0079

1999 0.0072 0.0093 0.0063 0.103 NA NA 0.0079

2000 0.0074 0.0097 0.0066 0.124 NA NA 0.0079

2001 0.0079 0.0103 0.0066 0.137 NA NA 0.0083

2002 0.0081 0.0111 0.0064 0.139 NA NA 0.0087

2003 0.0073 0.0087 0.0064 0.0508 NA NA 0.0075

2004 0.0073 0.0093 0.0059 0.0247 NA NA 0.0079

2005 0.0104 0.0126 0.0097 0.0892 NA NA 0.0112

a. 1996 to 2005 environmental reports (FDF 1997 to 1999; Fluor Fernald 2000 to 2006).

b. NA = not available.
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4.5.2 EXTERNAL DOSE CONTRIBUTION FROM WASTE PITS

Figure 4-11 shows the beta/gamma dose rate contours in the Waste Pit Area, which is in EA-5 

(DOE 1994a). Table 4-21 lists operational periods and the direct dose rate for each pit (DOE 1994a). 

Figure 4-12 shows the radiation profile of the Waste Pit Area, which accounts for the development of 

each pit (i.e., increasing dose rates as the pits were filled with wastes).

Figure 4-11. Beta/gamma dose rate (mrad/hr) contours in Waste Pit Area (DOE 1994a, p. 1213).

The beta and gamma radiation at the Waste Pit Area was emitted from radionuclides that were buried 

in the ground. Gamma radiation is emitted at ground level into the atmosphere, which then scatters 

the radiation to receptor locations some distance from the area; this is the skyshine phenomenon. 

Radiation from skyshine at the receptor location is typically about 1% of the dose rate at the source.
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Table 4-21. Waste pit dose rate data (DOE 1994a).

Waste Pit Operation period

Average ambient dose rate 

(mrad/hr)

1 1952–1959 1.0

2 1957–1964 1.2

3 1958–1977 2.0

4 1960–1986 16.0a

5 1968–1987 1.0

6 1979–1985 1.3

Clearwell 1960s–1970s 1.0

Burn pit 1957–1984 1.3

a. Waste Pit 4 dose rates are much higher than others due to having the highest 

uranium waste inventory. Calculations using National Council on Radiation Protection 

and Measurements methodology show that releases of uranium by far are the most 

important contributors to the potential doses from releases to the atmosphere at the 

FMPC (DOE 1994a).

Average Ambient Dose Rates

(mRad/hr)

Years

1952 1954 1958 1962 1966 1970 1974 1978

0

5

10

15

20

1982 1986 1990

25

Pit 1 Pit 2

Pit 3

Pit 4

Pit 5 Pit 6

Clearwell

Burn Pit

Waste Pit

Area Total

Figure 4-12. Waste pit dose rate profiles. 

4.5.3 ONSITE AMBIENT DOSE RATE ESTIMATES, 1976 TO 2005

From 1976 to 2005, when TLD stations were deployed to monitor the environmental radiation levels, 

the dose rate at each EA was estimated by summing the dose rates from the major direct dose 

contributors (i.e., K-65 silos, storage and production facilities, and the waste pits) using the measured 

dose rate data in Tables 4-19 and 4-20 and the estimated dose rate profile in Figure 4-12.
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The dose rates can be represented by Equation 4-12:

( ) ( )

         

= −× + −× + × +                          

2 2 2

0.01 BSs BSj k

i bkg bkg pit bkg

EAsi EAji EAki

d d d D BSs D Average BSj D D D

d d d (4-12)

where

Di = estimated dose rate at EAi 

BSs = measured dose rate at the fence line TLD station closest to the K-65 silos (BS6) 

(Starting in 1996, the K-65 silo fence line values in Table 4-19 were used in the 

equation and as EA-6 ambient dose rate.)

Dbkg = background dose rate

dBSs = distance from TLD stations to the K-65 silos

dEAsi = distance from K-65 silos to EAi 

BSj = measured dose rate at fence line TLD stations, where j represents the combination of 

BS1, 2, 3, 4, 5, 7, and 9; starting in 1996, BSj values were assigned the average 

background values because the onsite TLD station BS32 (Health and Safety Building) 

readings in Table 4-20 were below background

dBSj = distance from TLD stations to production plants

dEAji = distance from production plants to EAi

Dpit = average dose rate at the Waste Pit Area (the waste pit component of the equation was 

applied to all operating years from 1952 to 2002)

dk = distance from center of Waste Pit Area to edge of Waste Pit Area

dEAki = distance from Waste Pit Area to EAi

ORAUT (2015) documents Equation 4-12 results. These results were compared to the direct dose 

rate contour maps (Figures 4-13 and 4-14) from the EG&G aerial surveys in 1976 and 1985 (see 

Section 4.5.5). The comparisons have resulted in closely matched dose rates in each EA, as shown 

in Table 4-22.

4.5.4 ONSITE AMBIENT DOSE RATE ESTIMATES, 1952 TO 1975

Because no usable direct radiation measurements were found in the available reference materials for 

the period before 1976, it was necessary to estimate the ambient dose rates on the site for those 

years based on extrapolation from known data or correlation to historical events. The following 

sections present the methodologies that were used to extrapolate onsite dose rates for those years 

that predate the implementation of mature radiation monitoring programs at FMPC.

4.5.4.1 DOSE RATE DUE TO K-65 SILOS, BEFORE 1976 

Figure 4-15 is the histogram of the average direct dose rates at the K-65 silo fence line that was 

developed by a combination of interpreting the historic description of the K-65 silo radiation levels and 

the application of measured dose rate values.

As noted in Section 4.5.1, the ambient dose rates started to increase to measurable levels in 1955 

with the relocation of wastes from Niagara Falls, New York, to the silos. Before 1975, the dose rates 

in the vicinity of the North and South Tanks (i.e., the K-65 silos) that were measured by a Juno survey 

meter ranged from 10 to 30 mR/hr, respectively (Klein 1963). The source strength increased to the 

point that radiation levels require additional shielding to mitigate the direct radiation dose to personnel.
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Figure 4-13. 1976 exposure rate (μR/hr) contour map of FMPC survey area (Shipman 1985, p. 18).
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Figure 4-14. 1985 exposure rate (μR/hr) contour map of FMPC survey area (Shipman 1985, p. 15).
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Table 4-22. Comparison of estimated dose rates in EAs and aerial survey results (mrem/hr).

Exposure 

areas

1976 

Eq. 4-12

1976 

Aerial surveya

1977 

Eq. 4-12

1985 

Eq. 4-12

1985 

Aerial surveya

1986 

Eq. 4-12

EA-1 0.151 0.045–0.085 (J) 0.176 0.147 0.170–0.350 (L) 0.258

EA-2 0.107 0.045–0.085 (J) 0.124 0.104 0.170–0.350 (L) 0.181

EA-3 0.107 0.045–0.085 (J) 0.124 0.103 0.170–0.350 (L) 0.180

EA-4 0.0150 0.045–0.085 (J) 0.175 0.145 0.170–0.350 (L) 0.256

EA-5 0.0367 0.045–0.085 (J) 0.0342 0.047 0.045–0.65 (J) 0.046

EA-6 0.455 >0.350 (M) 0.279 0.659 >0.350 (M) 0.546

EA-7 0.0153 0.013–0.015 (F) 0.0145 0.0175 0.019–0.027 (H) 0.0176

EA-8 0.0194 0.013–0.015 (F) 0.0198 0.0211 0.027–0.045 (I) 0.0253

EA-9 0.0160 0.012–0.013 (E) 0.0159 0.0177 0.019–0.027 (H) 0.0194

EA-10 0.0151 0.012–0.013 (E) 0.0148 0.0169 0.019–0.027 (H) 0.0177

EA-11 0.0141 0.012–0.013 (E) 0.0128 0.0167 0.015–0.019 (G) 0.0152

a. See Figures 4-13 and 4-14.

Figure 4-15. Projected K-65 silos fence line dose rates histogram. 

In 1991, a bentonite layer was added (Environmental Report Fluor Fernald 2001, p. 13). An 

immediate dose rate reduction of approximately 1 order of magnitude occurred. Dose rates slowly 

increased again, and in 2002 monitoring showed an increase of about a factor of 4 in external 

radiation. Figure 4-15 shows that the dose rates progressively increased from 1952 to a “measurable” 

level in 1955. This level, as measured by a Juno survey meter in 1963, was interpreted to be an 

exposure rate of 30 mR/hr at 3 ft (1 m) from each (North and South) silo, a total of 60 mR/hr at 3 ft 

from the tanks. The K-65 fence line dose rate was calculated to be 0.0443 mrem/hr based on this 

measurement. The histogram shows that this radiation level was assumed to remain steady from 

1955 to 1963. Starting in 1964, the K-65 fence line dose rates once again incrementally increased 

each year to the 1976 level of 0.44 mrem/hr, which is the first year with TLD measurements. From 

1976 on, the histogram reflects that the radiation level increased until the early 1980s. Since then the 

dose rates built up again. Figure 4-15 forms the basis for the estimation of the radiation levels in each 

EA due to the K-65 silos for 1952 to 1976. For 1977 to the present, environmental TLD station 

measurements have provided the data for estimation of the EA dose rate levels. The resulting dose 

rates and doses for dose reconstruction are listed in ORAUT (2015).
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4.5.4.2 DOSE RATE DUE TO PRODUCTION AND STORAGE FACILITIES, BEFORE 1976

The onsite direct radiation levels due to radioactive materials that were processed, handled, and 

stored at site facilities before 1976 can be extrapolated by the site production rate and the capacity of 

the processing systems. This extrapolation is further supported by the fact that large amounts of 

thorium were not present in storage until 1976, when measurements from the environmental TLD 

stations (which were used to estimate the EA dose rate levels) were made available in the annual site 

environmental reports.

As production rates went up, the amount of radioactive materials in process equipment, in-process 

storage, material staging areas, and product storage also increased and thus caused the ambient 

radiation to increase. However, the radioactive materials increase was limited to the capacity of the 

site systems. When the systems were at their maximum capacities, the onsite ambient radiation was 

at its maximum level. Assuming that the site required two shifts that operated at 100% systems 

capacity to meet the peak production rate of 10,000 MTU/yr (Mead, Savage, and Fugate ca. 1985, 

p. 19), the sources of radiation were limited to full systems capacity of 5,000 MTU/yr. The peak 

production rate occurred from about 1954, when the production plants became operational, to 1964. 

In 1964, the production rates began to decline to a low in 1975 of approximately 1,230 MTU, which 

was about 25% of capacity. The production rate remained at 25% until 1981 when direction to restore 

the site facilities to accommodate projected product requirements close to the original production 

capacity was received. Thereafter, until production of uranium ended in 1988, the production rate 

increased to 3,700 MTU (3 times the 1979 level), which is about 74% of full capacity. The onsite 

ambient direct radiation dose rates for the years before 1976 were estimated by the ratio of the 

systems capacity to the corresponding measured dose rate in the 1980s. Table 4-23 lists the 

resulting onsite ambient dose rates. Figure 4-16 overlays the projected dose rate on the FMPC metal 

production rate profile.

The dose rate values in Table 4-23 represent the ambient dose rate in EA-4 (center of the FMPC 

production area) and were input into the storage and production component of Equation 4-12 to 

estimate the dose rate distribution for each EA. The calculations for dose reconstruction are provided 

in ORAUT (2015).

4.5.5 CALCULATED DOSE RATES AND RECORDED MEASUREMENTS COMPARISONS

In 1985, EG&G Energy Measurements conducted an aerial survey of FMPC and surrounding areas 

for radiation due to gamma emissions (Shipman 1985, p. 7). The survey resulted in dose rate contour 

maps of the site for 1976 and 1985 (Figures 4-13 and 4-14, respectively). The aerial survey data 

were compared with estimated dose rates for the EAs using site TLD measurements. The aerial 

survey data were particularly important because the 1976 data represent a low annual production rate 

of approximately 1,230 MTU, while those for 1985 represent a modest annual production rate of about 

3,700 MTU. The peak FMPC production rate was 10,000 MTU in 1960 (Mead, Savage, and Fugate 

ca. 1985, p. 19). The comparisons resulted in generally comparable dose rates in the EAs, as 

Table 4-23 shows.

4.6 ATTRIBUTIONS AND ANNOTATIONS

All information requiring identification was addressed via references integrated into the reference 

section of this document.
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Table 4-23. Projected onsite dose rates based on systems 

capacity. 

Year

Systems capacity 

(%)

Fraction of 1981 

valuesa

Dose rates 

(mrem/hr)

1951 0 0.00 0.00

1952 24.6 0.33 0.12

1953 30 0.41 0.15

1954 30 0.41 0.15

1955 50 0.68 0.24

1956 100 1.35 0.49

1957 100 1.35 0.49

1958 100 1.35 0.49

1959 100 1.35 0.49

1960 100 1.35 0.49

1961 100 1.35 0.49

1962 100 1.35 0.49

1963 100 1.35 0.49

1964 100 1.35 0.49

1965 100 1.35 0.49

1966 100 1.35 0.49

1967 100 1.35 0.49

1968 100 1.35 0.49

1969 100 1.35 0.49

1970 100 1.35 0.49

1971 86 1.16 0.42

1972 72 0.97 0.35

1973 57 0.77 0.28

1974 43 0.58 0.21

1975 31 0.42 0.15

1976 31 0.42 0.15

a. Systems capacity and the onsite dose rate in 1981 are 74% and 

0.362 mrem/hr, respectively.
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B. Shleien, and J. E. Till, 1995, The Fernald Dosimetry Reconstruction Project, Tasks 2 and 3, 

Radionuclide Source Terms and Uncertainties, RAC Report No. CDC-5, Radiological 

Assessments Corporation, Neeses, South Carolina, June. [SRDB Ref ID: 3767]

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Monitoring Annual Report for 1975, NLCO-1133, Cincinnati, Ohio, April 1. [SRDB Ref ID: 

3553, p. 1]

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Monitoring Annual Report for 1976, NLCO-1142, National Lead Company of Ohio, Cincinnati, 

Ohio, April 1. [SRDB Ref ID: 3553, p. 33]
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Boback, M. W., K. N. Ross, and D. A. Fuchs, 1978, Feed Materials Production Center Environmental 

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May 1. [SRDB Ref ID: 3553, p. 105]

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Annual Report for 1979, NLCO-1164, National Lead Company of Ohio, Cincinnati, Ohio, 

May 1. [SRDB Ref ID: 3553, p. 140]

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Annual Report for 1980, NLCO-1168, National Lead Company of Ohio, Cincinnati, Ohio, 

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Environmental Monitoring Annual Report for 1981, NLCO-1180, National Lead Company of 

Ohio, Cincinnati, Ohio, May 1. [SRDB Ref ID: 3553, p. 213]

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Annual Report for 1982, NLCO-1187, National Lead Company of Ohio, Cincinnati, Ohio, 

May 1. [SRDB Ref ID: 3553, p. 249]

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Annual Report for 1983, NLCO-2018, National Lead Company of Ohio, Cincinnati, Ohio, 

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Facemire, C.F., D. L. Jones, and R. W. Keys, 1985, Feed Materials Production Center Environmental 

Monitoring Annual Report for 1984, NLCO-2028, National Lead Company of Ohio, Cincinnati, 

Ohio, July 15. [SRDB Ref ID: 3553, p. 332]

Aas, C. A., D. L. Jones, and R. W. Keys, 1986, Feed Materials Production Center Environmental 

Monitoring Annual Report for 1985, FMPC-2047, Westinghouse Materials Company of Ohio, 

Cincinnati, Ohio, May 30. [SRDB Ref ID: 3553, p. 417]

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Environmental Monitoring Annual Report for 1986, FMPC-2076, Westinghouse Materials 

Company of Ohio, Cincinnati, Ohio, April 30. [SRDB Ref ID: 3554, p. 3]

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Ohio, April 30. [SRDB Ref ID: 3554, p. 148]

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Monitoring Annual Report for 1988, FMPC-2173, Westinghouse Materials Company of Ohio, 

Cincinnati, Ohio, June. [SRDB Ref ID: 3554, p. 337]

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Annual Environmental Report for Calendar Year 1989, FMPC-2200, Westinghouse Materials 

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Byrne, J. M., T. A. Dugan, and J. S. Oberjohn, 1991, Feed Materials Production Center Annual 

Environmental Report for Calendar Year 1990, FEMP-2245, Westinghouse Materials 

Company of Ohio, Cincinnati, Ohio, December. [SRDB Ref ID: 3556]

WEMCO (Westinghouse Environmental Management Company of Ohio), 1992, Annual Site 

Environmental Report for Calendar Year 1991, FEMP-2275, Fernald Environmental 

Management Project, Fernald, Ohio, December. [SRDB Ref ID: 3559]

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Environmental Report, FEMP-2290, Fernald Environmental Management Project, Fernald, 

Ohio, June. [SRDB Ref ID: 3563]

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Report, FEMP-2342, Fernald Environmental Management Project, Fernald, Ohio, June. 

[SRDB Ref ID: 3569]

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Report, with Addendum and Supplemental Addendum, FEMP-2518, Fernald Environmental 

Management Project, Fernald, Ohio, October 31. [SRDB Ref ID: 3578]

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Environmental Management Project, Fernald, Ohio. [SRDB Ref ID: 3586]

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Fernald Environmental Management Project, Fernald, Ohio, June. [SRDB Ref ID: 3592]

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Management Project, Fernald, Ohio, May. [SRDB Ref ID: 3516]

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Fernald, Ohio, June. [SRDB Ref ID: 126955]

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Fernald, Ohio, June. [SRDB Ref ID: 126956]

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Fernald, Ohio, May. [SRDB Ref ID: 129790]

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Stoller (S. M. Stoller Corporation), 2007, 2006 Fernald Site Environmental Report, DOE-LM/1461-

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129812]
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GLOSSARY

alpha radiation

Positively charged particle emitted from the nuclei of some radioactive elements. An alpha 

particle consists of two neutrons and two protons (a helium nucleus) and has an electrostatic 

charge of +2.

beta radiation

Charged particle emitted from some radioactive elements with a mass equal to 1/1,837 that of 

a proton. A negatively charged beta particle is identical to an electron. A positively charged 

beta particle is a positron.

curie (Ci)

Traditional unit of radioactivity equal to 37 billion (3.7 × 1010) becquerels, which is 

approximately equal to the activity of 1 gram of pure 226Ra.

dosimeter

Device that measures the quantity of received radiation, usually a holder with radiationabsorbing filters and radiation-sensitive inserts packaged to provide a record of absorbed dose 

received by an individual. See thermoluminescent dosimeter. 

dosimetry

Measurement and calculation of internal and external radiation doses.

dosimetry system

System for assessment of received radiation dose. This includes the fabrication, assignment, 

and processing of external dosimeters, and/or the collection and analysis of bioassay samples, 

and the interpretation and documentation of the results.

exposure

(1) In general, the act of being exposed to ionizing radiation. (2) Measure of the ionization 

produced by X- and gamma-ray photons in air in units of roentgens.

gamma radiation

Electromagnetic radiation (photons) of short wavelength and high energy (10 kiloelectron-volts 

to 9 megaelectron-volts) that originates in atomic nuclei and accompanies many nuclear 

reactions (e.g., fission, radioactive decay, and neutron capture). Gamma photons are identical 

to X-ray photons of high energy; the difference is that X-rays do not originate in the nucleus.

ionizing radiation

Radiation of high enough energy to remove an electron from a struck atom and leave behind a 

positively charged ion. High enough doses of ionizing radiation can cause cellular damage. 

Ionizing particles include alpha particles, beta particles, gamma rays, X-rays, neutrons, 

high-speed electrons, high-speed protons, photoelectrons, Compton electrons, 

positron/negatron pairs from photon radiation, and scattered nuclei from fast neutrons. See 

alpha radiation, beta radiation, gamma radiation, neutron radiation, photon radiation, and X-ray 

radiation. 

neutron radiation

Radiation that consists of free neutrons unattached to other subatomic particles emitted from a 

decaying radionuclide. Neutron radiation can cause further fission in fissionable material such 

as the chain reactions in nuclear reactors, and nonradioactive nuclides can become 

radioactive by absorbing free neutrons.
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photon

Quantum of electromagnetic energy generally regarded as a discrete particle having zero rest 

mass, no electric charge, and an indefinitely long lifetime. The entire range of electromagnetic 

radiation that extends in frequency from 1023 cycles per second (hertz) to 0 hertz.

photon radiation

Electromagnetic radiation that consists of quanta of energy (photons) from radiofrequency 

waves to gamma rays.

rem

Traditional unit of radiation dose equivalent that indicates the biological damage caused by 

radiation equivalent to that caused by 1 rad of high-penetration X-rays multiplied by a quality 

factor. The sievert is the International System unit; 1 rem equals 0.01 sievert. The word 

derives from roentgen equivalent in man; rem is also the plural.

roentgen (R)

Unit of photon (gamma or X-ray) exposure for which the resultant ionization liberates a positive 

or negative charge equal to 2.58 × 10-4 coulombs per kilogram (or 1 electrostatic unit of 

electricity per cubic centimeter) of dry air at 0 degrees Celsius and standard atmospheric 

pressure. An exposure of 1 R is approximately equivalent to an absorbed dose of 1 rad in soft 

tissue for higher energy photons (generally greater than 100 kiloelectron-volts).

thermoluminescence

Property that causes a material to emit light as a result of heat.

thermoluminescent dosimeter (TLD)

Device for measuring radiation dose that consists of a holder containing solid chips of material 

that, when heated, release the stored energy as light. The measurement of this light provides 

a measurement of absorbed dose.

working level (WL)

Unit of concentration in air of the short-lived decay products of 222Rn (218Po, 214Pb, 214Bi, and 

214Po) and 220Rn (216Po, 212Pb, 212Bi, 212Po) defined as any combination of the short-lived 

radioactive progeny of radon or thoron in 1 liter of air, without regard to the degree of 

equilibrium, that results in the ultimate emission of 130,000 megaelectron-volts of alpha 

energy; 1 WL equals 2.083 × 10-5 joules per cubic meter.

working level month (WLM)

Unit of exposure to radon progeny defined as exposure for 1 working month (170 working 

hours) to a potential alpha energy concentration of 1 WL; 1 WLM equals 1 WL times 

170 hours, which is 0.00354 joule-hours per cubic meter.

X-ray radiation

Electromagnetic radiation (photons) produced by bombardment of atoms by accelerated 

particles. X-rays are produced by various mechanisms including bremsstrahlung and electron 

shell transitions within atoms (characteristic X-rays). Once formed, there is no difference 

between X-rays and gamma rays, but gamma photons originate inside the nucleus of an atom.
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ATTACHMENT A 

INCORPORATION OF Q-11 SILOS AS A RADON RELEASE SOURCE, 1952 TO 1958

Figure A-1 shows the yearly maximum and mean exposures in WLMs to workers in the area of the 

Q-11 silos during the years when these silos were radon sources (1952 to 1958). The yearly radon 

releases from the Q-11 silos were estimated based on the maximum exposures in Figure A-1, and the 

source term was back-extrapolated using Plant 2/3 as the point of release. The derivation of the 

equation for the source term estimation is shown below.

Figure A-1. Mean and highest yearly worker radon exposure estimates due to Q-11 silos (Pinney et 

al. 2004, p. 61, Figure 13).

The radon concentrations at distances from a continuous source were represented by modifying 

Equation A-1:

Q11 ( ) EA

d

χ C R f RF

Q

  =     (A-1)

where

C = radionuclide concentration in EA (Ci/m3 EA ) 

RQ11 = radionuclide release rate from source Q-11 silos (Ci/yr)

(χ/Q)d = atmospheric dilution factor at distance d from Q-11 source (s/m3

) 

f = fraction of time that the wind blew in the EA direction from source

RF = respirable fraction of the source aerosol (<10 μm)
Document No. ORAUT-TKBS-0017-4 Revision No. 03 Effective Date: 12/03/2015 Page 62 of 65

ATTACHMENT A

INCORPORATION OF Q-11 SILOS AS A RADON RELEASE SOURCE, 1952 TO 1958 (continued)

Rearrangement of Equation A-1 yields:

( )

1

Q11 EA

d

χ R C f RF

Q

−

   

= ×         (A-2)

The radon concentration CEA was expressed in terms of exposure D in WLM using Equation 4-11, and 

the rearranged Equation A-2 becomes:

(pCi L ) = 0.0824 EA

D C (A-3)

CEA was substituted with Equation A-2 to yield:

( )

1

Q11 0.0824 d

D χ R f RF

Q

−

   

= ×         (A-4)

The radon release rate unit is converted to curies per year:

( )

1 3 3 7

3

Q11 12

10 L m (3.154 10 s yr ) (pCi L) (s m ) 10 pCi Ci 0.0824 dD χ R f RF

Q

−

× ×     = × ×         (A-5)

The Q-11 silo release rate RQ11 then reduced to:

( ) ( )

1

Q11 0.383

d

χ R D f RF

Q

−

   

= ××         (A-6)

where

RQ11 = annual radon release rate (Ci/yr)

D = dose (WLM)

(χ/Q)d = atmospheric dilution factor at distance d (m) from source, where d is a close-in 

distance of 25 m and (χ/Q)25m = 2.09 ×10-3 s/m3 with due consideration of building 

wake effects

f = fraction of time the wind blows toward the receptor from Plant 2/3 (because the 

receptor is in the immediate vicinity of Q-11 silos, 100% of the time the wind 

blows toward the receptor and f =1.0)

RF = respirable fraction of the source aerosol (1.0)

0.383 = conversion factor

Pinney et al. (2004) estimated the RQ11 exposures by building. The maximum dose was assumed to 

occur in the building nearest the Q-11 silos. The 25-m distance is a conservative assumption. The 

estimated annual radon release rate RQ11 from Equation A-6 was added to the K-65 source term as 

shown in Figure 4-4.
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ATTACHMENT B 

95TH-PERCENTILE WIND ROSE DATA FOR 16 COMPASS POINTS

Table B-1. Wind rose data and derived 95th-percentile values for environmental radionuclide concentration estimates 

(1991 to 2001 Environmental Reports: WEMCO 1992; FERMCO 1993 to 1996; FDF 1997 to 1999; Fluor Fernald 2000 to 

2002.). 

Years

N

f

N

u (mph)

N

u (m/s)

NNE

f

NNE

u (mph)

NNE

u (m/s)

NE

f

NE

u (mph)

NE

u (m/s)

ENE

f

ENE

u (mph)

ENE

u (m/s)

1991 0.040 6.10 2.73 0.034 5.80 2.60 0.049 5.40 2.42 0.073 6.00 2.69

1992 0.037 4.90 2.20 0.040 4.90 2.20 0.046 4.00 1.79 0.071 3.10 1.39

1993 0.030 5.50 2.46 0.040 5.00 2.24 0.048 4.50 2.02 0.070 4.40 1.97

1994 0.032 4.80 2.15 0.037 5.40 2.42 0.055 3.60 1.61 0.085 5.00 2.24

1995 0.045 5.70 2.55 0.034 5.80 2.60 0.054 5.50 2.46 0.080 5.00 2.24

1996 0.045 5.90 2.64 0.043 6.40 2.87 0.050 5.50 2.46 0.073 4.70 2.11

1997 0.046 5.20 2.33 0.048 6.00 2.69 0.050 6.00 2.69 0.080 5.50 2.46

2000 0.039 5.50 2.46 0.045 6.00 2.69 0.060 5.40 2.42 0.064 4.50 2.02

2001 0.030 5.50 2.46 0.032 5.80 2.60 0.053 4.80 2.15 0.063 4.50 2.02

95th percentile 0.045 5.906 2.646 0.045 6.157 2.758 0.057 5.713 2.560 0.081 5.568 2.495

Years

E

f

E

u (mph)

E

u (m/s)

ESE

f

ESE

u (mph)

ESE

u (m/s)

SE

f

SE

u (mph)

SE

u (m/s)

SSE

f

SSE

u (mph)

SSE

u (m/s)

1991 0.042 4.00 1.79 0.025 3.00 1.34 0.019 3.30 1.48 0.024 4.30 1.93

1992 0.030 3.00 1.34 0.020 2.50 1.12 0.025 3.00 1.34 0.031 3.90 1.75

1993 0.035 3.50 1.57 0.018 2.90 1.30 0.015 3.00 1.34 0.020 3.50 1.57

1994 0.045 3.80 1.70 0.015 1.50 0.67 0.018 2.50 1.12 0.018 2.50 1.12

1995 0.048 3.00 1.34 0.028 3.00 1.34 0.030 4.00 1.79 0.039 4.90 2.20

1996 0.045 3.30 1.48 0.028 2.80 1.25 0.025 3.30 1.48 0.045 4.50 2.02

1997 0.046 3.90 1.75 0.025 3.30 1.48 0.020 3.00 1.34 0.028 3.80 1.70

2000 0.040 3.60 1.61 0.020 3.00 1.34 0.020 3.20 1.43 0.030 4.00 1.79

2001 0.030 3.00 1.34 0.020 2.00 0.90 0.020 2.50 1.12 0.033 4.40 1.97

95th percentile 0.047 3.891 1.743 0.027 3.135 1.405 0.026 3.594 1.610 0.039 4.637 2.078
ATTACHMENT B

95TH-PERCENTILE WIND ROSE DATA FOR 16 COMPASS POINTS (continued)

Document No. ORAUT-TKBS-0017-4 Revision No. 03 Effective Date: 12/0

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Years

S

f

S

u (mph)

S

u (m/s)

SSW

f

SSW

u (mph)

SSW

u (m/s)

SW

f

SW

u (mph)

SW

u (m/s)

WSW

f

WSW

u (mph)

WSW

u (m/s)

1991 0.051 5.50 2.46 0.110 6.00 2.69 0.119 5.00 2.24 0.115 5.00 2.24

1992 0.061 4.80 2.15 0.103 4.70 2.11 0.115 4.70 2.11 0.090 3.70 1.66

1993 0.048 4.30 1.93 0.095 4.50 2.02 0.118 3.60 1.61 0.110 3.20 1.43

1994 0.018 4.00 1.79 0.050 4.30 1.93 0.090 4.80 2.15 0.018 4.70 2.11

1995 0.076 6.00 2.69 0.098 6.00 2.69 0.110 4.80 2.15 0.110 5.00 2.24

1996 0.075 5.50 2.46 0.108 5.40 2.42 0.100 4.50 2.02 0.090 4.60 2.06

1997 0.055 4.70 2.11 0.082 5.80 2.60 0.112 5.00 2.24 0.108 4.50 2.02

2000 0.066 5.60 2.51 0.125 6.50 2.91 0.129 4.60 2.06 0.109 4.00 1.79

2001 0.075 5.00 2.24 0.125 6.60 2.96 0.130 4.50 2.02 0.107 4.00 1.79

95th percentile 0.074 5.751 2.577 0.120 6.395 2.865 0.126 4.941 2.213 0.112 4.898 2.194

Years

W

f

W

u (mph)

W

u (m/s)

WNW

f

WNW

u (mph)

WNW

u (m/s)

SW

f

SW

u (mph)

SW

u (m/s)

WSW

f

WSW

u (mph)

WSW

u (m/s)

1991 0.104 5.00 2.24 0.081 5.20 2.33 0.064 4.20 1.88 0.050 5.20 2.33

1992 0.090 3.80 1.70 0.090 4.10 1.84 0.088 4.20 1.88 0.050 4.50 2.02

1993 0.112 3.40 1.52 0.105 3.50 1.57 0.075 3.50 1.57 0.040 4.80 2.15

1994 0.099 4.00 1.79 0.097 4.40 1.97 0.090 4.00 1.79 0.080 3.80 1.70

1995 0.112 4.80 2.15 0.100 5.00 2.24 0.070 4.00 1.79 0.058 5.00 2.24

1996 0.090 5.30 2.37 0.084 5.50 2.46 0.067 4.50 2.02 0.044 5.00 2.24

1997 0.118 5.20 2.33 0.088 5.00 2.24 0.075 4.30 1.93 0.047 4.70 2.11

2000 0.104 4.60 2.06 0.079 4.50 2.02 0.068 4.80 2.15 0.040 5.20 2.33

2001 0.095 4.00 1.79 0.080 4.50 2.02 0.075 5.00 2.24 0.038 5.50 2.46

95th percentile 0.114 5.161 2.312 0.100 5.276 2.364 0.085 4.774 2.139 0.064 5.312 2.380
 

Document No. ORAUT-TKBS-0017-4 Revision 

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ATTACHMENT C 

ANNUAL AVERAGE RADON BACKGROUND CONCENTRATIONS AND CONCENTRATION AT SELECTED RADON LOCATIONS, 

1989 TO 2005

Figure C-1. Annual average radon background concentrations and concentration at selected radon locations, 1989 to 2005 (Environmental

Report Fluor Fernald 2006).