ORAU TEAM

Dose Reconstruction

Project for NIOSH

Oak Ridge Associated Universities I Dade Moeller I MJW Technical Services

Page 1 of 65

DOE Review Release 03/24/2014

Feed Materials Production Center –

Occupational Environmental Dose

ORAUT-TKBS-0017-4 Rev 02

Effective Date: 03/13/2014

Supersedes: Revision 01

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

Document Owner 

Approval: Signature on File Approval Date: 03/07/2014

Karen S. Kent, Document Owner

Concurrence: Signature on File Concurrence Date: 03/11/2014

James P. Griffin, Deputy Project Director

Concurrence: Scott R. Siebert Signature on File for Concurrence Date: 03/12/2014

Edward F. Maher, Objective 3 Manager

Concurrence: Vickie S. Short Signature on File for Concurrence Date: 03/07/2014

Kate Kimpan, Project Director

Approval: Signature on File Approval Date: 03/13/2014

David S. Sundin, Deputy Director, DCAS

New Total Rewrite Revision Page Change

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

REVISION AND DISCARD / DESTROY ALL COPIES OF THE PRIOR REVISION.
Document No. ORAUT-TKBS-0017-4 Revision No. 02 Effective Date: 03/13/2014 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 non 

uranium 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.
Document No. ORAUT-TKBS-0017-4 Revision No. 02 Effective Date: 03/13/2014 Page 3 of 65

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.................................................................................................. 10

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

4.4 Internal Doses from Atmospheric Radionuclide Concentrations ............................................. 15

4.4.1 Radioactive Materials Air Concentrations and Intakes, 1951 to 1988.......................... 18

4.4.1.1 Radionuclide Discharges and Sources of Emissions.................................... 18

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...................................... 25

4.4.1.3.2 Far Field Atmospheric Dilution Factor, χ/Q................................ 27

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

4.4.1.3.4 Effects of Wind Direction, fi........................................................ 30

4.4.1.3.5 Concentration Fields Near Buildings, Modified χ/Q for 

Building Wake ........................................................................... 31

4.4.1.3.6 Episodic Releases..................................................................... 32

4.4.1.4 Radionuclide Intake Derivation..................................................................... 33

4.4.1.5 Results......................................................................................................... 35

4.4.2 Nonradon Radionuclide Air Concentrations and Intakes, 1989 to 2012 ...................... 35

4.4.2.1 Annual Air Emissions for Radionuclides Other than Radon.......................... 36

4.4.2.1.1 Monitored Stacks....................................................................... 36

4.4.2.1.2 Unmonitored Stacks .................................................................. 36

4.4.2.1.3 Scrubbers.................................................................................. 36

4.4.2.1.4 Building Vents ........................................................................... 36

4.4.2.1.5 Fugitive Emissions .................................................................... 36

4.4.3 Radon Air Concentrations and Intake Estimates, 1989 to 2012 .................................. 37

4.5 Onsite External Doses from Ambient Radiation...................................................................... 38

4.5.1 K-65 Silos and Production Plants External Dose Contribution .................................... 41

4.5.2 External Dose Contribution from Waste Pits............................................................... 44

4.5.3 Onsite Ambient Dose Rate Estimates, 1976 to 2005 .................................................. 44

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 .................................................. 48

4.5.4.2 Dose Rate Due to Production and Storage Facilities, Before 1976............... 50

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

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

References ........................................................................................................................................ 53

Glossary ............................................................................................................................................ 59
Document No. ORAUT-TKBS-0017-4 Revision No. 02 Effective Date: 03/13/2014 Page 4 of 65

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

4-4 Historical information and data available for environmental dose reconstruction .................... 18

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............................. 22

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

4-9 Releases of 222Rn used in this analysis .................................................................................. 25

4-10 Receptor downwind estimated distances from contributing emission sources ........................ 29

4-11 95th-percentile values from wind rose data ............................................................................ 31

4-12 Building and K-65 silo dimensions for building wake χ/Q calculations .................................... 32

4-13 Uranium enrichment isotopes................................................................................................. 33

4-14 RU contaminant radionuclides ............................................................................................... 34

4-15 Ratio of 226Ra and 230Th to uranium........................................................................................ 34

4-16 Alphabetical list of fields......................................................................................................... 35

4-17 Medians of the maximum values of all uranium air concentrations samples by year .............. 37

4-18 Medians of the maximum values of all radon air concentrations samples by year.................. 38

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

4-20 TLD measurement summary.................................................................................................. 43

4-21 Waste pit dose rate data ........................................................................................................ 45

4-22 Comparison of estimated dose rates in EAs and aerial survey results.................................... 49

4-23 Projected onsite dose rates based on systems capacity......................................................... 51

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

concentration estimates ......................................................................................................... 63
Document No. ORAUT-TKBS-0017-4 Revision No. 02 Effective Date: 03/13/2014 Page 5 of 65

LIST OF FIGURES

FIGURE TITLE PAGE

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

4-2 Estimate of total 222Rn releases from K-65 and Q-11 silos...................................................... 24

4-3 Measured emission stack particle size from 15 different FMPC stacks................................... 26

4-4 Percentage of stack emision less than approximately 10 μm AMAD. .................................... 26

4-5 Horizontal diffusion σy versus downwind distance from point source for Pasquill’s 

stability categories..................................................................................................................28

4-6 Vertical diffusion σz versus downwind distance from continuously emitting point source 

for Pasquill’s stability categories............................................................................................. 29

4-7 95th-percentile wind rose data, 10-m height, 1991 to 1997 and 2000 to 2001........................ 30

4-8 Offsite and fenceline radon monitoring locations .................................................................... 39

4-9 Onsite radon monitoring locations .......................................................................................... 40

4-10 Fenceline TLD locations, 1976 to 1995 .................................................................................. 43

4-11 TLD monitoring locations, 1996 to 2005................................................................................. 44

4-12 Beta/gamma dose rate contours in waste pit area.................................................................. 45

4-13 Waste pit dose rate profiles.................................................................................................... 46

4-14 1976 exposure rate contour map of FMPC survey area ......................................................... 47

4-15 1985 exposure rate contour map of FMPC survey area ......................................................... 48

4-16 Projected K-65 silos fenceline dose rates histogram .............................................................. 49

4-17 Projected sitewide direct radiation dose rates based on metal production rate....................... 52

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
Document No. ORAUT-TKBS-0017-4 Revision No. 02 Effective Date: 03/13/2014 Page 6 of 65

ACRONYMS AND ABBREVIATIONS

AMS air monitoring station

Bq becquerel

CFR Code of Federal Regulations

Ci curie

d day

DOE U.S. Department of Energy

DOL U.S. Department of Labor

DCF Dose Conversion Factor

EA Exposure Area

EEOICPA Energy Employees Occupational Illness Compensation Program Act of 2000

FR Federal Register

FEMP Fernald Environmental Management Project

FMPC Feed Materials Production Center

ft foot

g gram

GSD geometric standard deviation

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

NCRP National Council on Radiation Protection and Measurements

NIOSH National Institute for Occupational Safety and Health

ORAU Oak Ridge Associated Universities

pCi picocurie

POC probability of causation

RU recycled uranium

s second

SRDB Ref ID Site Research Database Reference Identification (number)
Document No. ORAUT-TKBS-0017-4 Revision No. 02 Effective Date: 03/13/2014 Page 7 of 65

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

μR microroentgen

§ section or sections
Document No. ORAUT-TKBS-0017-4 Revision No. 02 Effective Date: 03/13/2014 Page 8 of 65

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 sites or categories of sites. They will be revised in the event 

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

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

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

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

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

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

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

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

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

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

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

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

facility encompassed by EEOICPA.

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

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

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

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

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

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

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

determined in accordance with the POC [probability of causation1

] guidelines established under 

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

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

referred to in 42 U.S.C. § 7384n(b) to nuclear weapons work (NIOSH 2010).

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

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

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

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

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

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

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

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

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

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

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

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

occupationally derived (NIOSH 2010):

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

conventional structures

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

reasons

 1 The U.S. Department of Labor (DOL) is ultimately responsible under the EEOICPA for determining the POC.
Document No. ORAUT-TKBS-0017-4 Revision No. 02 Effective Date: 03/13/2014 Page 9 of 65

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 while on 

the site but outside the FMPC buildings or facilities from radioactive material discharges to the 

atmosphere from the facilities and from ambient external radiation that originated in the 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 workers who were unmonitored 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 emission nuclide quantities. Environmental measurements were used to establish intake rates 

for periods 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.

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 Special Exposure Cohort (SEC):

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 (77 FR 150; August 3, 2012). 

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, which employees at 

this facility may have been subjected (Sebelius 2012). 

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 (78 FR 229; November 

27, 2013). 
Document No. ORAUT-TKBS-0017-4 Revision No. 02 Effective Date: 03/13/2014 Page 10 of 65

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, which employees 

working at this facility may have been subjected (Sebelius 2013a).

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 (78 FR 229; November 27, 2013).

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

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

FMPC working at this facility may have been subjected. 

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

subcontractors. NIOSH has access to an electronic data set that contains the results of the 

uranium urinalysis bioassay program for all the 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 data set. However, the data set 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 employees’ bioassay data were retained by the site or captured by data 

capture efforts. Additionally, 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 before January 1, 1984, 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 < 250 

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

workers for the period January 1, 1951, through December 31, 1983, NIOSH has determined, 

and HHS has concurred, that it is feasible to reconstruct external radiation doses for all FMPC 

workers for the period from January 1, 1951, through December 31, 1983 (Sebelius, 2013b).

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 to 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).
Document No. ORAUT-TKBS-0017-4 Revision No. 02 Effective Date: 03/13/2014 Page 11 of 65

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, radium-bearing) 

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 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 in this TBD 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. This TBD presents airborne radionuclide 

intake rates in Table 4-2. A spreadsheet (ORAUT 2014) documents the various calculations that 

were used to develop annual exposures and intake rates for each source, EA, and year combination. 

The maximum site wide annual intakes are assigned in the Interactive RadioEpidemiological Program 

(IREP) as a lognormal distribution with a geometric standard deviation (GSD) of 3.0.
Document No. ORAUT-TKBS-0017-4 Revision No. 02 Effective Date: 03/13/2014 Page 12 of 65

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, 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

EA-1 EA-2

EA-4 EA-3

EA-8

EA-5

EA-6

EA-11 EA-7

EA-9

EA-10
Document No. ORAUT-TKBS-0017-4 Revision No. 02 Effective Date: 03/13/2014 Page 13 of 65

Table 4-2. Maximum site wide average (1951 to 1988) or median (1989 to 2012) annual intake rates (WLM/yr for 222Rn and Bq/yr for all 

other radionuclides) via inhalation.

a,c,d

Year

Total Ub

(Bq/yr)

Th-228 

(Bq/yr)

Th-232 

(Bq/yr)

Pu-239 

(Bq/yr)

Np-237 

(Bq/yr)

Tc-99 

(Bq/yr)

Ru-103/106 

(Bq/yr)

Zr/Nb-95

(Bq/yr)

(Sr-90) 

(Bq/yr)

Rn-222 

(WLM)

1951 5.87E-01 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 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 2.20E+01

1953 1.57E+03 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 3.95E+01

1954 1.14E+03 4.83E+00 4.83E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 3.93E+01

1955 2.84E+03 5.52E+00 5.52E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 5.68E+01

1956 1.18E+03 5.18E+00 5.18E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 5.68E+01

1957 8.35E+02 2.55E-03 2.55E-03 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 5.47E+01

1958 9.19E+02 5.83E-03 5.83E-03 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 4.72E+01

1959 1.06E+03 5.83E-03 5.83E-03 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 1.55E+00

1960 5.56E+03 7.51E-02 7.51E-02 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 1.55E+00

1961 1.32E+03 1.46E-01 1.46E-01 1.06E+01 4.21E+00 2.59E+02 1.85E+02 5.56E+01 7.41E+00 1.55E+00

1962 9.39E+02 2.86E-01 2.86E-01 7.54E+00 3.00E+00 1.85E+02 1.32E+02 3.96E+01 5.28E+00 1.55E+00

1963 1.01E+03 3.20E-01 3.21E-01 8.07E+00 3.21E+00 1.98E+02 1.41E+02 4.24E+01 5.66E+00 1.55E+00

1964 7.46E+02 2.46E+00 2.46E+00 5.99E+00 2.38E+00 1.47E+02 1.05E+02 3.15E+01 4.20E+00 1.55E+00

1965 3.34E+02 2.46E+00 2.46E+00 2.10E+00 8.36E-01 5.15E+01 3.68E+01 1.10E+01 1.47E+00 1.55E+00

1966 3.12E+04 3.12E+00 3.12E+00 1.96E+02 7.81E+01 4.81E+03 3.44E+03 1.03E+03 1.38E+02 1.55E+00

1967 2.80E+02 2.46E+00 2.46E+00 1.76E+00 7.00E-01 4.31E+01 3.09E+01 9.25E+00 1.23E+00 1.55E+00

1968 4.92E+02 2.47E+00 2.47E+00 3.09E+00 1.23E+00 7.57E+01 5.41E+01 1.62E+01 2.17E+00 1.55E+00

1969 1.95E+02 1.43E+01 1.43E+01 1.23E+00 4.88E-01 3.00E+01 2.15E+01 6.44E+00 8.59E-01 1.55E+00

1970 1.78E+02 1.05E+01 1.05E+01 1.12E+00 4.44E-01 2.74E+01 1.96E+01 5.87E+00 7.83E-01 1.55E+00

1971 1.07E+02 5.56E+00 5.56E+00 6.75E-01 2.68E-01 1.65E+01 1.18E+01 3.54E+00 4.73E-01 1.55E+00

1972 2.00E+02 1.03E+00 1.03E+00 1.26E+00 4.99E-01 3.07E+01 2.20E+01 6.59E+00 8.80E-01 1.55E+00

1973 2.80E+02 5.69E-01 5.69E-01 7.04E+00 2.20E+00 1.51E+02 3.08E+01 9.23E+00 1.23E+00 1.55E+00

1974 3.12E+02 7.44E-01 7.44E-01 7.85E+00 2.45E+00 1.69E+02 3.44E+01 1.03E+01 1.37E+00 1.55E+00

1975 3.65E+02 5.41E-01 5.41E-01 9.19E+00 2.87E+00 1.97E+02 4.02E+01 1.21E+01 1.61E+00 1.55E+00

1976 3.04E+02 6.63E-01 6.63E-01 7.66E+00 2.39E+00 1.65E+02 3.35E+01 1.01E+01 1.34E+00 1.55E+00

1977 1.31E+02 2.94E+00 2.94E+00 3.30E+00 1.03E+00 7.10E+01 1.45E+01 4.34E+00 5.78E-01 1.55E+00

1978 6.78E+01 2.94E+00 2.94E+00 1.71E+00 5.33E-01 3.67E+01 7.47E+00 2.24E+00 2.99E-01 1.55E+00

1979 1.26E+02 2.94E+00 2.94E+00 3.17E+00 9.89E-01 6.80E+01 1.39E+01 4.15E+00 5.54E-01 1.55E+00

1980 4.82E+01 7.51E-01 7.51E-01 1.21E+00 3.79E-01 2.60E+01 5.31E+00 1.59E+00 2.12E-01 3.41E-01

1981 5.74E+01 7.51E-01 7.51E-01 1.44E+00 4.51E-01 3.10E+01 6.32E+00 1.90E+00 2.53E-01 3.41E-01

1982 4.94E+01 7.51E-01 7.51E-01 1.24E+00 3.88E-01 2.67E+01 5.44E+00 1.63E+00 2.18E-01 2.03E-01

1983 8.31E+01 7.51E-01 7.51E-01 2.09E+00 6.54E-01 4.49E+01 9.15E+00 2.74E+00 3.66E-01 2.03E-01

1984 5.85E+01 7.51E-01 7.51E-01 1.47E+00 4.60E-01 3.16E+01 6.44E+00 1.93E+00 2.58E-01 2.03E-01

1985 4.65E+01 7.51E-01 7.51E-01 1.17E+00 3.66E-01 2.51E+01 5.12E+00 1.54E+00 2.05E-01 2.03E-01

1986 4.74E+01 7.51E-01 7.51E-01 1.19E+00 3.73E-01 2.56E+01 5.22E+00 1.57E+00 2.09E-01 2.03E-01
Document No. ORAUT-TKBS-0017-4 Revision No. 02 Effective Date: 03/13/2014 Page 14 of 65

Year

Total Ub

(Bq/yr)

Th-228 

(Bq/yr)

Th-232 

(Bq/yr)

Pu-239 

(Bq/yr)

Np-237 

(Bq/yr)

Tc-99 

(Bq/yr)

Ru-103/106 

(Bq/yr)

Zr/Nb-95

(Bq/yr)

(Sr-90) 

(Bq/yr)

Rn-222 

(WLM)

1987 4.53E+01 7.85E-01 7.85E-01 1.14E+00 3.56E-01 2.45E+01 4.98E+00 1.49E+00 1.99E-01 2.03E-01

1988 4.40E+01 8.14E-01 8.14E-01 1.11E+00 3.46E-01 2.38E+01 4.84E+00 1.45E+00 1.94E-01 1.07E-01

1989 3.65E+01 2.76E-05 3.94E-05 9.18E-01 2.87E-01 1.97E+01 4.02E+00 1.21E+00 1.61E-01 1.00E-01

1990 1.97E+01 1.49E-05 2.13E-05 4.96E-01 1.55E-01 1.07E+01 2.17E+00 6.51E-01 8.68E-02 9.22E-02

1991 3.06E+01 2.32E-05 3.31E-05 7.71E-01 2.41E-01 1.66E+01 3.38E+00 1.01E+00 1.35E-01 1.05E-01

1992 1.65E+01 1.25E-05 1.79E-05 4.16E-01 1.30E-01 8.94E+00 1.82E+00 5.46E-01 7.29E-02 1.05E-01

1993 1.70E+01 1.29E-05 1.84E-05 4.28E-01 1.34E-01 9.20E+00 1.88E+00 5.62E-01 7.50E-02 1.21E-01

1994 9.49E+00 7.18E-06 1.03E-05 2.39E-01 7.46E-02 5.13E+00 1.04E+00 3.13E-01 4.18E-02 1.21E-01

1995 1.46E-01 1.10E-07 1.58E-07 3.67E-03 1.15E-03 7.89E-02 1.61E-02 4.82E-03 6.43E-04 1.13E-01

1996 2.04E+01 1.55E-05 2.21E-05 5.14E-01 1.61E-01 1.10E+01 2.25E+00 6.75E-01 9.00E-02 1.54E-01

1997 2.34E+01 1.77E-05 2.52E-05 5.88E-01 1.84E-01 1.26E+01 2.57E+00 7.71E-01 1.03E-01 1.70E-01

1998 2.12E+01 1.60E-05 2.29E-05 5.32E-01 1.66E-01 1.14E+01 2.33E+00 6.99E-01 9.32E-02 5.52E-02

1999 2.85E+01 2.16E-05 3.08E-05 7.17E-01 2.24E-01 1.54E+01 3.14E+00 9.41E-01 1.25E-01 1.21E-01

2000 2.75E+01 2.08E-05 2.97E-05 6.92E-01 2.16E-01 1.49E+01 3.03E+00 9.08E-01 1.21E-01 1.05E-01

2001 2.54E+01 1.93E-05 2.75E-05 6.40E-01 2.00E-01 1.38E+01 2.80E+00 8.40E-01 1.12E-01 1.05E-01

2002 1.35E+01 1.02E-05 1.46E-05 3.39E-01 1.06E-01 7.28E+00 1.48E+00 4.45E-01 5.94E-02 1.21E-01

2003 5.69E+01 4.31E-05 6.15E-05 1.43E+00 4.48E-01 3.08E+01 6.27E+00 1.88E+00 2.51E-01 1.05E-01

2004 3.19E+02 2.41E-04 3.45E-04 8.02E+00 2.51E+00 1.72E+02 3.51E+01 1.05E+01 1.40E+00 1.05E-01

2005 9.07E+00 6.87E-06 9.81E-06 2.28E-01 7.14E-02 4.90E+00 9.99E-01 3.00E-01 4.00E-02 9.64E-02

2006 3.16E+00 2.39E-06 3.42E-06 7.96E-02 2.49E-02 1.71E+00 3.48E-01 1.04E-01 1.39E-02 1.09E-01

2007 1.22E+00 9.20E-07 1.31E-06 3.06E-02 9.57E-03 6.57E-01 1.34E-01 4.02E-02 5.36E-03 9.64E-02

2008 8.03E-01 6.07E-07 8.68E-07 2.02E-02 6.31E-03 4.34E-01 8.84E-02 2.65E-02 3.54E-03 1.09E-01

2009 6.57E-01 4.97E-07 7.10E-07 1.65E-02 5.17E-03 3.55E-01 7.23E-02 2.17E-02 2.89E-03 1.09E-01

2010 6.63E-01 5.01E-07 7.16E-07 1.67E-02 5.21E-03 3.58E-01 7.30E-02 2.19E-02 2.92E-03 1.09E-01

2011 6.63E-01 5.01E-07 7.16E-07 1.67E-02 5.21E-03 3.58E-01 7.30E-02 2.19E-02 2.92E-03 1.09E-01

2012 6.63E-01 5.01E-07 7.16E-07 1.67E-02 5.21E-03 3.58E-01 7.30E-02 2.19E-02 2.92E-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. Three short-duration episodic emissions (Table 4-6) were modeled using Equation 4-7 and averaged into the annual intake rates.

d. Intake rates for Th-230 and Ra-226 were calculated to be in the range of a few μBq/yr but are not shown because they are judged to be negligibly small.
Document No. ORAUT-TKBS-0017-4 Revision No. 02 Effective Date: 03/13/2014 Page 15 of 65

The external gamma dose rate site wide exposure rates by the EAs are presented in ORAUT (2014) 

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, then the 

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

distribution in IREP. For best-estimate and compensable claims, the annual site average gamma 

doses should be used with a lognormal distribution in IREP with a GSD of 1.3.

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.
Document No. ORAUT-TKBS-0017-4 Revision No. 02 Effective Date: 03/13/2014 Page 16 of 65

Table4-3. 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 86.51,615.0 34.4 46.5 36.2 33.2 31.3 307.5 465.01,615

1980 419.0 293.6 290.8 413.0 99.92,145.6 39.0 49.2 39.3 36.6 37.0 351.2 615.02,146

1981 740.5 515.5 512.8 734.4 106.52,150.8 44.1 65.6 48.4 43.0 38.1 454.5 631.92,151

1982 754.1 526.4 524.3 749.2 101.71,626.1 45.0 67.8 50.5 45.1 38.2 411.7 498.51,626

1983 766.2 534.5 532.2 760.9 106.41,802.3 45.8 68.7 51.1 45.6 39.1 432.1 543.51,802

1984 361.0 253.9 252.1 356.7 91.51,281.9 33.2 42.9 34.7 32.4 30.8 251.9 366.81,282

1985 294.3 208.6 206.8 289.9 94.91,318.8 35.0 42.2 35.5 33.9 33.4 235.8 374.21,319

1986 516.1 361.2 359.7 512.2 92.01,091.1 35.1 50.6 38.8 35.4 30.5 283.9 332.91,091

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

1988 252.2 181.0 179.0 247.3 103.41,517.4 40.8 45.9 40.2 39.0 39.5 244.2 430.71,517

1989 196.6 141.4 139.7 192.3 93.11,247.7 34.0 37.9 33.4 32.6 32.9 198.3 354.11,248
Document No. ORAUT-TKBS-0017-4 Revision No. 02 Effective Date: 03/13/2014 Page 17 of 65

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.

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 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

2010 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

2011 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

2012 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 since background doses were not available for all of the dose components.

b. Dose Reconstructors should use Ambient (H*10) DCFs for 1952-1976 (based on instrument readings) and Exposure (R) DCFs for 1977-2012 (TLD on a fenceline 

based), (Section 4.1.2 of NIOSH 2007a).

c. Ambient gamma radiation doses are based on 2000 hours/year.
Document No. ORAUT-TKBS-0017-4 Revision No. 02 Effective Date: 03/13/2014 Page 18 of 65

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

Period

Available historic information and data 

description Source documenta

1951–1988 Airborne uranium, thorium, and radon 

emissions including unmonitored and 

episodic releases.

History of FMPC Radionuclide Discharges (Boback 

et al. 1987)

Addendum to FMPC-2082, “History of FMPC 

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

1989–1996 Radon in ambient air from continuous, 

high-volume AMSs at the site fenceline

locations. Results for onsite locations not 

reported until 1997.

Environmental Reports for 1989 to 1996

1997–2005 Radon in ambient air data from the Radon 

Monitoring Program that included AMSs 

at onsite locations.

Environmental Reports for 1997 to 2005

2006–2012b 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

a. For convenience, the annual environmental reports for 1975 to 2012 are listed sequentially in the Environmental Reports 

section at the end of the Reference list. 

b. 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 time spans 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 FMPC 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 (from 1951 through 1988). Dolan and Hill (1988), an addendum to Boback et al.,

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 quantities from either 

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. 

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

progeny are from Voillequé et al. 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). 
Document No. ORAUT-TKBS-0017-4 Revision No. 02 Effective Date: 03/13/2014 Page 19 of 65

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

Year

Plant

1

Plant

2/3a Plant4

Plant

5

Plant

6b

Plant

7c Plant8d

Plant

9e PilotPlantf Buildingexhaustg

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
Document No. ORAUT-TKBS-0017-4 Revision No. 02 Effective Date: 03/13/2014 Page 20 of 65

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

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.

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 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.
Document No. ORAUT-TKBS-0017-4 Revision No. 02 Effective Date: 03/13/2014 Page 21 of 65

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) to show a summary of six episodic 

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

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 summarizes annual fugitive thorium emissions due to wind erosion from 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).

ORAUT (2011) provides a background discussion and summary NIOSH position on the activity of 

222Rn that was emitted from the K-65 silos, which is between 5,000 and 6,000 Ci/yr during years when 

the silos were filled. Appendix J of Voillequé et al. (1995) estimated emissions of 222Rn and radon 

progeny primarily from the K-65 silos and from other stored K-65 materials, including the drummed 

material stored at Plant 1, which would eventually be 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 obtained 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.
Document No. ORAUT-TKBS-0017-4 Revision No. 02 Effective Date: 03/13/2014 Page 22 of 65

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

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
Document No. ORAUT-TKBS-0017-4 Revision No. 02 Effective Date: 03/13/2014 Page 23 of 65

Year Plant 8 Plant 9 Pilot Plant Yearly total

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 

month (WLM) 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. is reproduced in Figure A-1 in 

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

Combining each of these emission sources, Figure 4-2 shows a graphic representation of the 222Rn 

emission quantity estimates and Table 4-9 summarizes the annual median radon release estimates 

from (Voillequé et al. 1995) and maximized radon emission rates derived from the data in Pinney et al. 

(2004). Waste Pits were also a source of radon at FMPC but the quantities released were negligible 

compared to those in Table 4-9. To ensure 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 background 

radon concentrations measured from 1989 to 2000 (Environmental Report Fluor Fernald 2001).

Attachment C shows background concentration measurements.

4.4.1.2 Exposure Site Locations

Airborne emissions data indicate that the quantity 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 in the previous section 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 (2014).
Document No. ORAUT-TKBS-0017-4 Revision No. 02 Effective Date: 03/13/2014 Page 24 of 65

100

1000

10000

1951 1954 1957 1960 1963 1966 1969 1972 1975 1978 1981 1984 1987

Time Period

Total Rn222 Release Rate (Ci/Yr)

1952-1953

1953-1958 1959-1979 1980-1987

1988

95th Percentile

Median

100000

Q-11

K-65

Figure 4-2. Estimate of total 222Rn releases from K-65 (Voillequé et al. 1995, p. 45) and Q-11 silos

(Pinney et al. 2004, p. 61).

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.

Year

Highest 

Exposures 

(WLM)

Emission 

(Ci) 

(Eq. B-5)

1953 45 8,250

1954 94 17,200

1955 93 17,000

1956 135 24,700

1957 135 24,700

1958 130 23,800

1959 112 20,500
Document No. ORAUT-TKBS-0017-4 Revision No. 02 Effective Date: 03/13/2014 Page 25 of 65

Table 4-9. Releases (Bq) of 222Rn used in this analysis (Voillequé et al. 

1995).

Year K-65 silos Q-11 silo

Plant 1 Pad, 

Drum Storage Annual total

1951 0.0E+00 0.00E+00 1.30E+12 1.30E+12

1952 7.4E+13 3.05E+14 2.15E+13 4.01E+14

1953 7.4E+13 6.37E+14 3.52E+12 7.15E+14

1954 1.9E+14 6.31E+14 0.00E+00 8.21E+14

1955 1.9E+14 9.15E+14 0.00E+00 1.11E+15

1956 1.9E+14 9.15E+14 0.00E+00 1.11E+15

1957 1.9E+14 8.81E+14 0.00E+00 1.07E+15

1958 1.9E+14 7.59E+14 0.00E+00 9.49E+14

1959 2.2E+14 0.00E+00 0.00E+00 2.20E+14

1960 2.2E+14 0.00E+00 0.00E+00 2.20E+14

1961 2.2E+14 0.00E+00 0.00E+00 2.20E+14

1962 2.2E+14 0.00E+00 0.00E+00 2.20E+14

1963 2.2E+14 0.00E+00 0.00E+00 2.20E+14

1964 2.2E+14 0.00E+00 0.00E+00 2.20E+14

1965 2.2E+14 0.00E+00 0.00E+00 2.20E+14

1966 2.2E+14 0.00E+00 0.00E+00 2.20E+14

1967 2.2E+14 0.00E+00 0.00E+00 2.20E+14

1968 2.2E+14 0.00E+00 0.00E+00 2.20E+14

1969 2.2E+14 0.00E+00 0.00E+00 2.20E+14

1970 2.2E+14 0.00E+00 0.00E+00 2.20E+14

1971 2.2E+14 0.00E+00 0.00E+00 2.20E+14

1972 2.2E+14 0.00E+00 0.00E+00 2.20E+14

1973 2.2E+14 0.00E+00 0.00E+00 2.20E+14

1974 2.2E+14 0.00E+00 0.00E+00 2.20E+14

1975 2.2E+14 0.00E+00 0.00E+00 2.20E+14

1976 2.2E+14 0.00E+00 0.00E+00 2.20E+14

1977 2.2E+14 0.00E+00 0.00E+00 2.20E+14

1978 2.2E+14 0.00E+00 0.00E+00 2.20E+14

1979 2.2E+14 0.00E+00 0.00E+00 2.20E+14

1980 4.4E+13 0.00E+00 0.00E+00 4.40E+13

1981 4.4E+13 0.00E+00 0.00E+00 4.40E+13

1982 4.4E+13 0.00E+00 0.00E+00 4.40E+13

1983 4.4E+13 0.00E+00 0.00E+00 4.40E+13

1984 4.4E+13 0.00E+00 0.00E+00 4.40E+13

1985 4.4E+13 0.00E+00 0.00E+00 4.40E+13

1986 4.4E+13 0.00E+00 0.00E+00 4.40E+13

1987 4.4E+13 0.00E+00 0.00E+00 4.40E+13

1988 1.9E+13 0.00E+00 0.00E+00 1.90E+13

Total 6.11E+15 5.04E+15 2.63E+13 1.12E+16

4.4.1.3.1 Radionuclide Emission Rates, Ri × RF

The average annualized emission rates were derived by dividing emission quantities 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 quantities 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 

data sets 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 
Document No. ORAUT-TKBS-0017-4 Revision No. 02 Effective Date: 03/13/2014 Page 26 of 65

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 re-sampled with a result of 9.0 μm. Figure 4-3 shows 

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

 Figure 4-3. Measured emission stack particle size from 15 different FMPC

 stacks (Spenceley 1985).

The measurements were made using a multi-stage particle size separator and, 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 in the neighborhood of 10 μm, the nominal cutoff 

value associated with respirable particles, was available in each case. Typically (in 13 out 16 

instances) the cutoff value was greater than 11 μm so the percentage of respirable particles tends to 

be slightly overestimated and this bias is favorable to the claimant for dose reconstruction purposes. 

Figure 4-4 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 4-4 is 74%. 

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

 AMAD (Spenceley 1985).

0

2

4

6

8

10

12

0 2 4 6 8 10 12 14 16

AMAD, micrometers

Stack observation number

Measured Emission Stack Particle Size from 15 

Different FMPC Stacks

0

20

40

60

80

100

0 2 4 6 8 10 12 14 16

Percentage

Stack observation number

Percentage of Emissions Less than 

Approximately 10 μm 
Document No. ORAUT-TKBS-0017-4 Revision No. 02 Effective Date: 03/13/2014 Page 27 of 65

Appendix E of 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.

 

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. 184):

2 2

2 2

(,) 1 1 exp

y z 2 y z

χxy y H

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)

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 4-5 and 4-6, 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

(,) 1 1 exp 2 2 y z y z

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 4-5 and 4-6, 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.

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 
Document No. ORAUT-TKBS-0017-4 Revision No. 02 Effective Date: 03/13/2014 Page 28 of 65

the emission sources to the first encountered occupied location in each EA. These distances are the 

shortest distance between the emission sources and the occupied locations. For EAs that contained 

Figure 4-5. Horizontal diffusion σy versus downwind distance from point source for 

Pasquill’s stability categories (Slade 1968, p. 115).
Document No. ORAUT-TKBS-0017-4 Revision No. 02 Effective Date: 03/13/2014 Page 29 of 65

Figure 4-6. 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

Contributing emission sources

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. 

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 Figures 4-5 and 

4-6, the “F” stability category represents a moderately stable condition that has the lowest σy and σz, 

which would result 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, which is also favorable to 

claimants, from Table 2 of Boback et al. (1987, p. 54), which 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 
Document No. ORAUT-TKBS-0017-4 Revision No. 02 Effective Date: 03/13/2014 Page 30 of 65

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 the 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 groundlevel 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 annual environmental reports for 1991 to 2001 (see 

the end of the Reference list) 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 4-7) to determine the wind direction frequency for each of the compass points. Table 4-11

presents the summary wind rose data that was used in this analysis. Attachment B contains more 

detail.

Figure 4-7. 95th-percentile wind rose data, 10-m height, 1991 to 1997 and 

2000 to 2001 (FDF 1998, p. 35; Fluor Fernald 2002, p. 28).

Percentage of time that the 

wind blew from this direction

N

S

W E

NNE

NE

ENE

ESE

SE

SSW SSE

SW

WSW

WNW

NW

NNW

10

15 LEGEND

Average wind speed from 

this direction (mph)

5
Document No. ORAUT-TKBS-0017-4 Revision No. 02 Effective Date: 03/13/2014 Page 31 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 methodology in 

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

0.7( /2) 0.0967( 3 ) σ W xH y b ′ =+− (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.
Document No. ORAUT-TKBS-0017-4 Revision No. 02 Effective Date: 03/13/2014 Page 32 of 65

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. 39) 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, such as those from Events 1, 2, and 3, were evaluated as quasi-instantaneous (puff) releases. 

The remaining episodic releases (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

yI zI 2 yI zI

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

)

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

X = downwind distance (m)

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)

h = effective release height (m)

For particulates, multiply Equation 4-7 by one-half. 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:
Document No. ORAUT-TKBS-0017-4 Revision No. 02 Effective Date: 03/13/2014 Page 33 of 65

0.93 σ X yl = × 0.14 (4-8)

The value of σzI is calculated as:

0.61 σ X ×= × zl 0.05 (4-9)

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

( ) ψ I q f Br RF Q

  = × ×× ×     (4-10)

where

I = radionuclide intake (g)

q = episodic release quantity (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 4-8)

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 that 

was 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).

Uranium 

enrichment

Specific 

activity 

(Bq/mg U) Isotope

Activity 

fraction

1% enriched 29 U-234 0.637

U-235 0.021

U-238 0.342

2% enriched 37 U-234 0.769

U-235 0.0268

U-238 0.204

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

documented the resolution of several internal dosimetry issues including the basis for recycled 

uranium (RU) default values in Table 4-14. RU was first processed in significant quantity at FMPC 

beginning in 1961, so none of the RU-associated contaminants was present before 1961. Between 

1961 and 1972, RU was present in the process stream at concentrations shown in Table 4-14. 

Beginning in 1973, some RU contaminants were present in higher concentrations.
Document No. ORAUT-TKBS-0017-4 Revision No. 02 Effective Date: 03/13/2014 Page 34 of 65

Table 4-14. RU contaminant radionuclides (Bq/g U) (ORAUT 2011, pg. 44).

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

Ru103/106 Zr/Nb-95 Sr-90

Before 1961 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00

1961–1972 2.33E+02 9.25E+01 5.70E+03 4.00E-02 2.80E-02 4.08E+03 1.22E+03 1.63E+02

After 1972 9.31E+02 2.91E+02 2.00E+04 4.00E-02 2.80E-02 4.08E+03 1.22E+03 1.63E+02

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 operating 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.

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 (2014) for derivation.

Radon

Radon-222 intake levels are expressed in units of working level months (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 in a month. 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

100 170

8.24 10

222 pCi Rn WLM in a year Concentration

L

pCi Concentration

L

−

×   = ×   ×  

  = ××    

(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.
Document No. ORAUT-TKBS-0017-4 Revision No. 02 Effective Date: 03/13/2014 Page 35 of 65

Radon Exposure Records

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

appear in “Personnel Exposure” records, separate from the “DOE Response file.” Dose 

Reconstructors 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, Dose 

Reconstructors should assign only the highest dose determined from this comparison, not both. The 

radon intakes from the “Personnel Exposure Records” are assumed to be a chronic exposure and 

assigned as a constant distribution in IREP for overestimates. For best-estimate or compensable 

claims, because the dose is almost instantaneous, Dose Reconstructors 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 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 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,” 

“Q11_TOTAL,” AND “TOTAL_EXP.” The alphabetical listing of fields and their content is shown in 

Table 4-16 which provides information to Dose Reconstructors on what the fields represent.

 Table 4-16. Alphabetical list of fields (Pinney et al. 2004, p. 67).

Field Content

SSN Social Security Number

SURVEY_TOT_EXP Yearly K65 + Q11 radon exposure (WLM) from survey data

SURVEY_YRLY_K65 Yearly K65 radon exposure (WLM) from survey data

SURVEY_YRLY_Q11 Yearly Q11 radon exposure (WLM) from survey data

TOTAL_EXP Yearly K65+Q11 radon exposure (WLM) from estimated shift

YR Start date year

YRLY_K65 Yearly K65 radon exposure (WLM) from estimated shift

YRLY_Q11 Yearly Q11 radon exposure (WLM) from estimated shift

4.4.1.5 Results

ORAUT (2014) 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 are values summarized and, for each radionuclide of interest, the intake rate 

in the EA with the highest value is assigned for site wide application. Table 4-2 presents these results 

in summary form.

4.4.2 Nonradon Radionuclide Air Concentrations and Intakes, 1989 to 2012

Starting in 1989, annual environmental reports to DOE reported estimates of the amount of 

radioactive airborne material that was released from the site. FMPC used those amounts to estimate 

how much radiation exposure to which workers, nearby residents, and the population in general might 

have received as the result of radioactive airborne emissions from the facilities at FMPC. The annual 

airborne uranium concentrations from the onsite and perimeter air monitoring stations were used with 

the model in ORAUT-PROC-0031, Site Profile and Technical Basis Document Development (ORAUT 

2012b), to determine the onsite air concentrations and intakes for environmental dose to workers for 

1989 to 2012. 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 
Document No. ORAUT-TKBS-0017-4 Revision No. 02 Effective Date: 03/13/2014 Page 36 of 65

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.

4.4.2.1.1 Monitored Stacks

Stack discharges are monitored by taking continuous isokinetic particulate samples, which are 

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 EPA 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 air in the plants is significantly lower than the calculated average of the continuous air 

monitor results.

4.4.2.1.5 Fugitive Emissions

These estimates were based upon 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.

Table 4-17 presents the median of maximum values of all uranium air samples that were collected 

after 1988.
Document No. ORAUT-TKBS-0017-4 Revision No. 02 Effective Date: 03/13/2014 Page 37 of 65

Table 4-17. Medians of the maximum values of all uranium

air concentrations (Bq/m3

) samples by year (ORAUT 2014).

Year Uranium air concentration

1989 2.03E-02

1990 1.09E-02

1991 1.70E-02

1992 9.19E-03

1993 9.46E-03

1994 5.27E-03

1995 8.11E-05

1996 1.14E-02

1997 1.30E-02

1998 1.18E-02

1999 1.58E-02

2000 1.53E-02

2001 1.41E-02

2002 7.49E-03

2003 3.16E-02

2004 1.77E-01

2005 5.04E-03

2006 1.76E-03

2007 6.76E-04

2008 4.46E-04

2009 3.65E-04

2010 3.65E-04a

2011 3.65E-04a

2012 3.65E-04a

a. Results based on 2009 air sampling data when particulate uranium 

site sampling ended.

Results

Uranium intake rates and intakes of associated trace contaminants are reported in Table 4-2. 

Calculation of the intake rates is presented in ORAUT (2014).

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 are changed

quarterly and sent to the supplier for analysis. At year’s end, 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.

The monitoring locations changed over time as remedial actions occurred. Figures 4-8 and 4-9 show 

monitoring locations in 1994 timeframe for:

• 11 locations offsite,
Document No. ORAUT-TKBS-0017-4 Revision No. 02 Effective Date: 03/13/2014 Page 38 of 65

• 21 locations along the fence line,

• 4 locations onsite 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 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.

Table 4-18. Medians of the maximum values of all radon air 

concentrations (Bq/m3

) samples by year (ORAUT 2014).

Year Radon air concentration

1989 7.50E-01

1990 6.50E-01

1991 8.00E-01

1992 8.00E-01

1993 1.00E+00

1994 1.00E+00

1995 9.00E-01

1996 1.40E+00

1997 1.60E+00

1998 2.00E-01

1999 1.00E+00

2000 8.00E-01

2001 8.00E-01

2002 1.00E+00

2003 8.00E-01

2004 8.00E-01

2005 7.00E-01

2006 8.50E-01

2007 7.00E-01

2008 8.50E-01

2009 8.50E-01a

2010 8.50E-01a

2011 8.50E-01a

2012 8.50E-01a

a. Results based on 2008 air sampling data, when radon site sampling 

ended.

4.5 ONSITE EXTERNAL DOSES FROM AMBIENT RADIATION

External environmental radiation dose at FMPC has been primarily 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.
Document No. ORAUT-TKBS-0017-4 Revision No. 02 Effective Date: 03/13/2014 Page 39 of 65

Figure 4-8. Offsite and fenceline radon monitoring locations (Environmental Report FERMCO 

1995, p. 156).

AMS-1
Document No. ORAUT-TKBS-0017-4 Revision No. 02 Effective Date: 03/13/2014 Page 40 of 65

Figure 4-9. Onsite radon monitoring locations (Environmental Report FERMCO 1995, p. 157).

AMS-8a

AMS-9b
Document No. ORAUT-TKBS-0017-4 Revision No. 02 Effective Date: 03/13/2014 Page 41 of 65

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 

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. 11). 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 one order of magnitude occurred as a result of the bentonite. However, the 

bentonite had provided only a temporary reduction. Dose rates slowly increased again, and in 2002 

monitoring showed an increase of about a factor of four in external radiation.

The only data found from before 1975 (Klein 1963, p. 2) 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, p. 2) 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 (2014) 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 (1976 Environmental Report Boback, 

Ross, and Fuchs 1977, p. 62). Table 4-19 is a summary of the maximum measured dose rates at the 

monitoring stations for 1976 through 1995. Figure 4-10 shows the locations of the fence line

monitoring stations.

Since 1996, FMPC expanded its direct radiation monitoring system to measure 32 locations 

continuously including several onsite monitoring locations. Table 4-20 provides summary level 

information about direct radiation measurements for 1996 through 2005. Figure 4-11 shows the 

locations of the TLD stations.

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.
Document No. ORAUT-TKBS-0017-4 Revision No. 02 Effective Date: 03/13/2014 Page 42 of 65

Table 4-19. Radiation dose rates at site boundary (mR/hr or mrem/hr).

a,b

Year

Maximum measured dose at TLD stations

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.

b. NA = not available.
Document No. ORAUT-TKBS-0017-4 Revision No. 02 Effective Date: 03/13/2014 Page 43 of 65

N

Center at 

Plant 4

Plant Perimeter

Production Area Perimeter

FEMP

Waste 

Pit Area

BS1

BS2

BS3

BS4

BS5

On Site

BS7 BS8

BS9

BS10

8

8 EA No.

9

1 2

10

5

6

11 7

4 3

TLD Locations

Radiation Source 

BS6

BS1A

Figure 4-10 Fenceline TLD locations, 1976 to 1995 (Environmental Reports

FERMCO 1993 to 1996).

Table 4-20. TLD measurement summary (mrem/hr).

a

Year

TLD monitoring results

Fence line Onsite Offsite

Background

Min. Max. average

Min.

(Health and

Safety Bldg)

Max.

(K-65 silos

fence line) Min. Max.

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.
Document No. ORAUT-TKBS-0017-4 Revision No. 02 Effective Date: 03/13/2014 Page 44 of 65

N

Center at 

Plant 4

Plant Perimeter

Production Area Perimeter

FEMP

Waste 

Pit Area

8

8 EA No.

9

1 2

10

5

6

11 7

4 3

Fenceline TLD Locations

Radiation Source 

Onsite TLD Locations

Silo

2

Silo

1

Figure 4-11. TLD monitoring locations, 1996 to 2005 (Environmental Report Fluor Fernald 2004).

4.5.2 External Dose Contribution from Waste Pits

Figure 4-12 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 4-13 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).

The beta and gamma radiation at the Waste Pit area is 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 typically is about 1% of the dose rate at the source.

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 waste pits) using the measured 

dose rate data in Tables 4-19 and 4-20 and the estimated dose rate profile in Figure 4-13. 
Document No. ORAUT-TKBS-0017-4 Revision No. 02 Effective Date: 03/13/2014 Page 45 of 65

Figure 4-12. Beta/gamma dose rate (mrad/hr) contours in waste pit area (DOE 1994a, SRDB 

Ref ID 3574, p. 1213).

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 NCRP 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, SRDB 

Ref ID 3574, p. 23).
Document No. ORAUT-TKBS-0017-4 Revision No. 02 Effective Date: 03/13/2014 Page 46 of 65

Figure 4-13. Waste pit dose rate profiles.

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 [j is 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 (2014) documents Equation 4-12 results. These results were compared to the direct dose 

rate contour maps (Figures 4-14 and 4-15) 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.

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
Document No. ORAUT-TKBS-0017-4 Revision No. 02 Effective Date: 03/13/2014 Page 47 of 65

Figure 4-14. 1976 exposure rate (μR/hr) contour map of FMPC survey area (Shipman 1985, p. 18).

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.
Document No. ORAUT-TKBS-0017-4 Revision No. 02 Effective Date: 03/13/2014 Page 48 of 65

Figure 4-15. 1985 exposure rate (μR/hr) contour map of FMPC survey area (Shipman 1985, p. 15).

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

Figure 4-16 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, p. 2). The source strength increased to 
Document No. ORAUT-TKBS-0017-4 Revision No. 02 Effective Date: 03/13/2014 Page 49 of 65

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

Exposure 

areas

Estimated dose rates compared to aerial surveys (mrem/hr)

1976 1977 1985 1986

Eq. 4-12 Aerial surveya Eq. 4-12 Eq. 4-12 Aerial surveya 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 4-14 and 4-15.

Figure 4-16. Projected K-65 silos fenceline dose rates histogram.

the point that radiation levels require additional shielding to mitigate the direct radiation dose to 

personnel. In 1991, a bentonite layer was added (Environmental Report Fluor Fernald 2001, p. 11). 

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 4-16 shows that the dose rates progressively increased from 1952 to a “measurable” 

level in 1955. The “measurable” level as measured by 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 silo 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 

0.00

0.20

0.40

0.60

0.80

1.00

1.20

19511953195519571959196119631965196719691971197319751977197919811983198519871989199119931995199719992001

Projected Dose Rates at the K-65 silos Fenceline 

(mrem/hr)
Document No. ORAUT-TKBS-0017-4 Revision No. 02 Effective Date: 03/13/2014 Page 50 of 65

1976 on, the histogram reflects that the radiation level increased until the early 1980s. Since then, the 

dose rates built up again until recently. Figure 4-16 forms the basis for the estimation of the radiation 

levels in each EA due to the K-65 silos for 1952 to 1976. Since then, the 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 (2014).

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 quantities 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, and 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% site systems capacity. The site production capacity 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 

site production rate increased to 3,700 MTU (3 times the 1979 level), which is about 74% of full 

production 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 4-17 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 (2014).
Document No. ORAUT-TKBS-0017-4 Revision No. 02 Effective Date: 03/13/2014 Page 51 of 65

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.

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 4-14 and 4-15, 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 is addressed via references integrated into the References 

section of this document.
Document No. ORAUT-TKBS-0017-4 Revision No. 02 Effective Date: 03/13/2014 Page 52 of 65

10,000 MTU (2 shifts 

@ 100% system 

capacity)

1,230 MTU (1 shift 

@ 24.6% 

capacity)

3,690 MTU (1 shift 

@ 74% capacity

1960 1964 1968 1975 1981 1989

5,000 MTU (1 shift 

@ 100% system 

capacity) 

1970 1973

2,000 

4,000 

6,000 

8,000 

10,000 

Metric Ton Uranium 

Years

1951 1953 1956

Plants 1, 2/3, 4, 5, 6, 8

Plants 7, 9

Pilot Plant

9,000 MTU (2 shifts @ 

90% system capacity)

7,000 MTU (2 shifts @ 

70% system capacity)

3,000 MTU (1 shift 

@ 60% system 

capacity) 

0.05

0.10

0.50

1.00

1.50

Estimated on-site environmental doserates 

based on production rates.

Legend

Estimated on-site environmental doserates 

derived from environmental TLD data.

Measured average background

0.01

Average Onsite Environmental Doserates 

(mrem/hour)

Figure 4-17. Projected site wide direct radiation dose rates based on metal production rate.
Document No. ORAUT-TKBS-0017-4 Revision No. 02 Effective Date: 03/13/2014 Page 53 of 65

REFERENCES

77 FR 150 (Federal Register, Volume 77, Number 150, p. 46438), 2012, “Final Effect of Designation 

of a Class of Employees for Addition to the Special Exposure Cohort,” U.S. Department of 

Health and Human Services, Washington, D.C., August 3. [SRDB Ref ID: 127271]

78 FR 229 (Federal Register, Volume 78, Number 229, p. 70949), 2013, “Final Effect of Designation 

of a Class of Employees for Addition to the Special Exposure Cohort,” U.S. Department of 

Health and Human Services, Washington, D.C., November 27. [SRDB Ref ID: 129226]

ASI (Advanced Sciences Inc.), ca. 1986, Remedial Investigation of the Feed Material Production 

Center, Part I: Evaluation of Current Situation. [SRDB Ref ID: 1858]

Boback, M. W., 1978, “Gamma Levels Inside K-65 Tank,” memorandum to R. C. Heatherton, National 

Lead Company of Ohio, Cincinnati, Ohio, September 11. [SRDB Ref ID: 3054]

Boback, M. W., T. A. Dugan, D. A. Fleming, R. B. Grant, and R. W. Keys, 1987, History of FEMP 

Radionuclide Discharges, FMPC-2082, Westinghouse Materials Company of Ohio, Cincinnati, 

Ohio, May. [SRDB Ref ID: 4270]

Cember, H., 1983, Introduction to Health Physics, Second Edition, Pergamon Press, New York, New 

York. [SRDB Ref ID: 127691]

DOE (U. S. Department of Energy), 1994a, Remedial Investigation Report for Operable Unit 1, 

Fernald Environmental Management Project, Fernald, Ohio, Remedial Investigation and 

Feasibility Study, Fernald Environmental Management Project, Fernald Field Office, Fernald, 

Ohio, August. [SRDB Ref ID: 3574, 3577, 3579, 3581, 3582, 3591]

DOE (U.S. Department of Energy), 1994b, Airborne Release Fractions/Rates and Respirable 

Fractions for Nonreactor Nuclear Facilities, Volume I - Analysis of Experimental Data, DOEHDBK-3010-94, U.S. Department of Energy, Washington, D.C., December. [SRDB Ref ID: 

56569]

Dolan, L. C., and C. A. Hill, 1988, Addendum to FEMP-2082, “History of FEMP Radionuclide 

Discharges”, Westinghouse Materials Company of Ohio, Cincinnati, Ohio, December. [SRDB 

Ref ID: 2165]

Hornung, R. W., S. M. Pinney, J. Lodwick, G. G. Killough, D. E. Brewer, and D. Nasuta, 2008, 

“Estimation of Radon Exposure to Workers at the Fernald Feed Materials Production Center, 

1952–1988,” Journal of Exposure Science and Environmental Epidemiology. [SRDB Ref ID: 

41619]

ICRP (International Commission on Radiological Protection), 1995, Dose Coefficients for Intakes of 

Radionuclides by Workers, Publication 68, Pergamon Press, Oxford, England. [SRDB Ref ID: 

22731]

Klein, F. J., 1963, “Contact Radiation Levels from K-65 Tanks,” memorandum to R. L. Pischoff, 

October 16. [SRDB Ref ID: 3017]

Mead, J., F. Savage, and R. Fugate, ca. 1985, History of the Operations of the Feed Materials 

Production Center by NLO, Inc., National Lead Company of Ohio, Feed Materials Production 

Center, Fernald, Ohio. [SRDB Ref ID: 26097]
Document No. ORAUT-TKBS-0017-4 Revision No. 02 Effective Date: 03/13/2014 Page 54 of 65

Napier, B. A., and J. V. Ramsdell, Jr., 2005, Atmospheric Data Package for the Composite Analysis, 

Pacific Northwest National Laboratory, PNNL-14599, Rev. 1, Battelle, Pacific Northwest 

National Laboratory, Richland, Washington, September. [SRDB Ref ID: 86460]

NCRP (National Council on Radiation Protection and Measurements), 1984, Evaluation of 

Occupational and Environmental Exposures to Radon and Radon Daughters in the United 

States, Report 78, Bethesda, Maryland, May.

NIOSH (National Institute for Occupational Health and Safety), 2007a, External Dose Reconstruction 

Implementation Guideline, OCAS-IG-001, Rev. 3, Office of Compensation Analysis and 

Support, Cincinnati, Ohio, November 21.

NIOSH (National Institute for Occupational Safety and Health), 2007b, Working Group Meeting, 

Advisory Board on Radiation and Worker Health, Fernald, Advisory Board on Radiation and 

Worker Health, Cincinnati, Ohio, October 24. [SRDB Ref ID: 129789]

NIOSH (National Institute for Occupational Safety and Health), 2010, Radiation Exposures Covered 

for Dose Reconstructions Under Part B of the Energy Employees Occupational Illness 

Compensation Program Act, DCAS-IG-003, Rev. 1, Division of Compensation Analysis and 

Support, Cincinnati, Ohio, October 5.

NIOSH (National Institute for Occupational Safety and Health), 2013a, Lung Dose Conversion Factor 

for Thoron WLM, DCAS-TIB-011, Rev. 4, Division of Compensation Analysis and Support, 

May 1.

NIOSH (National Institute for Occupational Safety and Health), ca. 2013b, Fernald worker exposures 

resulting from K-65 & Q-11 exposure from an electronic database titled “fernald_exposures 

(K65-1951-88),” Oak Ridge, Tennessee. [SRDB Ref ID: 78212]

NIOSH (National Institute for Occupational Safety and Health), ca. 2013c, Fernald worker exposures 

resulting from K-65 & Q-11 exposure (summary) from an electronic database titled 

“fernald_exposures (K65-Summary),” Oak Ridge, Tennessee. [SRDB Ref ID: 78213]

ORAUT (Oak Ridge Associated Universities Team), 2004, Technical Basis Document for the Fernald 

Environmental Management Project (FEMP) – Occupational Internal Dose, ORAUT-TKBS0017-5, Rev 00, Oak Ridge, Tennessee, May 28.

ORAUT (Oak Ridge Associated Universities Team), 2011, Feed Materials Production Center Internal 

Dose Topics, ORAUT-RPRT-0052, Rev. 00, Oak Ridge, Tennessee, July 12.

ORAUT (Oak Ridge Associated Universities Team), 2012, Site Profile and Technical Basis Document 

Development, ORAUT-PROC-0031, Rev. 04, Oak Ridge, Tennessee, November 5.

ORAUT (Oak Ridge Associated Universities Team), 2014, “Fernald Environmental Technical Basis 

Document (ORAUT-TKBS-0017-4), Rev 02-C spreadsheet,” Oak Ridge, Tennessee, 

December. [SRDB Ref ID: 130094]

Pinney, S. M., R. Hornung, J. Lodwick, D. Brewer, and J. Nasuta, 2004, Condensed Final Report for 

Radon and Cigarette Smoking Exposure Assessment of Fernald Workers, University of 

Cincinnati, College of Medicine and Institute of Health Policies and Health Services, 

Cincinnati, Ohio, October 22. [SRDB Ref ID: 26771]
Document No. ORAUT-TKBS-0017-4 Revision No. 02 Effective Date: 03/13/2014 Page 55 of 65

Randerson, D., editor, 1984, Atmospheric Science and Power Production, DOE/TIC-27601, U.S. 

Department of Energy, Office of Scientific and Technical Information, Oak Ridge, Tennessee, 

July. [SRDB Ref ID: 128885]

Sebelius, K. 2012, “HHS Designation of Additional Members of the Special Exposure Cohort under 

the Energy Employees Occupational Illness Compensation Program Act of 2000, Fernald, 

Ohio, June 27. [SRDB Ref ID: 129894]

Sebelius, K. 2013a, “HHS Designation of Additional Members of the Special Exposure Cohort under 

the Energy Employees Occupational Illness Compensation Program Act of 2000, Fernald, 

Ohio, September 30. [SRDB Ref ID: 127923]

Sebelius, K. 2013b, “HHS Designation of Additional Members of the Special Exposure Cohort under 

the Energy Employees Occupational Illness Compensation Program Act of 2000, Fernald, 

Ohio, September 30. [SRDB Ref ID: 129961]

Shipman, G. R., 1985, An Aerial Radiological Survey of the Feed Material Production Center and 

Surrounding Areas, Fernald, Ohio, Date of Survey, April 1985, EGG-10282-1084, EG&G 

Energy Measurements, October. [SRDB Ref ID: 10859]

Slade, D. H., editor, 1968, Meteorology and Atomic Energy, 1968, U.S. Atomic Energy Commission, 

Division of Reactor Development and Technology, Washington, D. C., July. [SRDB Ref ID: 

118302]

Spenceley, R.M., 1985, Partial Data for Major Emission Stacks – Third and Final Report, 

memorandum to J.A. Reafsnyder, National Lead Company of Ohio, Feed Materials Production 

Center, Fernald, Ohio. October 3. [SRDB Ref ID: 84388]

Voillequé, P. G., K. R. Meyer, D. W. Schmidt, S. K. Rope, G. G. Killough, M. Case, R. E. Moore, B. 

Shleiens, 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]

Annual Environmental Reports for 1975 to 2005 (in Date Order)

NLO (National Lead Company of Ohio), 1976, Feed Materials Production Center Environmental 

Monitoring Annual Report for 1975, NLCO-1133, Cincinnati, Ohio, April 1. [SRDB Ref ID: 

3553, p. 1]

Boback, M. W., K. N. Ross, and D. A. Fuchs, 1977, Feed Materials Production Center Environmental 

Monitoring Annual Report for 1976, NLCO-1142, National Lead Company of Ohio, Cincinnati, 

Ohio, April 1. [SRDB Ref ID: 3553, p. 33]

Boback, M. W., K. N. Ross, and D. A. Fuchs, 1978, Feed Materials Production Center Environmental 

Monitoring Annual Report for 1977, NLCO-1151, National Lead Company of Ohio, Cincinnati, 

Ohio, April 1. [SRDB Ref ID: 3553, p. 71]

Boback, M. W., and K. N. Ross, 1979, Feed Materials Production Center Environmental Monitoring 

Annual Report for 1978, NLCO-1159, National Lead Company of Ohio, Cincinnati, Ohio, May 

1. [SRDB Ref ID: 3553, p. 105]
Document No. ORAUT-TKBS-0017-4 Revision No. 02 Effective Date: 03/13/2014 Page 56 of 65

Boback, M. W., and K. N. Ross, 1980, Feed Materials Production Center Environmental Monitoring 

Annual Report for 1979, NLCO-1164, National Lead Company of Ohio, Cincinnati, Ohio, May 

1. [SRDB Ref ID: 3553, p. 140]

Boback, M. W., and K. N. Ross, 1981, Feed Materials Production Center Environmental Monitoring 

Annual Report for 1980, NLCO-1168, National Lead Company of Ohio, Cincinnati, Ohio, April 

1. [SRDB Ref ID: 3553, p. 176]

Fleming, D. A., M. W. Boback, and K. N. Ross, 1982, Feed Materials Production Center 

Environmental Monitoring Annual Report for 1981, NLCO-1180, National Lead Company of 

Ohio, Cincinnati, Ohio, May 1. [SRDB Ref ID: 3553, p. 213]

Fleming, D. A., and K. N. Ross, 1983, Feed Materials Production Center Environmental Monitoring 

Annual Report for 1982, NLCO-1187, National Lead Company of Ohio, Cincinnati, Ohio, May 

1. [SRDB Ref ID: 3553, p. 249]

Fleming, D. A., and K. N. Ross, 1984, Feed Materials Production Center Environmental Monitoring 

Annual Report for 1983, NLCO-2018, National Lead Company of Ohio, Cincinnati, Ohio, 

August. [SRDB Ref ID: 3553, p. 285]

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]

Aas, C. A., S. J. Clement, G. L. Gels, and C. A. Lojeck, 1987, Feed Materials Production Center 

Environmental Monitoring Annual Report for 1986, FMPC-2076, Westinghouse Materials 

Company of Ohio, Cincinnati, Ohio, April 30. [SRDB Ref ID: 3554, p. 3]

Gels, G. L., and C. A. Lojeck, 1988, Feed Materials Production Center Environmental Monitoring 

Annual Report for 1987, FMPC-2135, Westinghouse Materials Company of Ohio, Cincinnati, 

Ohio, April 30. [SRDB Ref ID: 3554, p. 148]

Gels, G. L., M. A. Gibb, and J. S. Oberjohn, 1989, Feed Materials Production Center Environmental 

Monitoring Annual Report for 1988, FMPC-2173, Westinghouse Materials Company of Ohio, 

Cincinnati, Ohio, June. [SRDB Ref ID: 3554, p. 337]

Dugan T. A., G. L. Gels, J. S. Oberjohn, and L. K. Rogers, 1990, Feed Materials Production Center 

Annual Environmental Report for Calendar Year 1989, FMPC-2200, Westinghouse Materials 

Company of Ohio, Cincinnati, Ohio, October. [SRDB Ref ID: 3555]

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. [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]
Document No. ORAUT-TKBS-0017-4 Revision No. 02 Effective Date: 03/13/2014 Page 57 of 65

FERMCO (Fernald Environment Restoration Management Company), 1993, 1992 Fernald Site 

Environmental Report, FEMP-2290, Fernald Environmental Management Project, Fernald, 

Ohio, June. [SRDB Ref ID: 3563]

FERMCO (Fernald Environment Restoration Management Company), 1994, 1993 Site Environmental 

Report, FEMP-2342, Fernald Environmental Management Project, Fernald, Ohio, June. 

[SRDB Ref ID: 3569]

FERMCO (Fernald Environment Restoration Management Company), 1995, 1994 Site Environmental 

Report, FEMP-2441, Fernald Environmental Management Project, Fernald, Ohio, July. [SRDB 

Ref ID: 3575]

FERMCO (Fernald Environment Restoration Management Company), 1996, 1995 Site Environmental 

Report, with Addendum and Supplemental Addendum, FEMP-2518, Fernald Environmental 

Management Project, Fernald, Ohio, October 31. [SRDB Ref ID: 3578]

FDF (Fluor Daniel Fernald), 1997, 1996 Site Environmental Report, FEMP-2538, Fernald 

Environmental Management Project, Fernald, Ohio. [SRDB Ref ID: 3586]

FDF (Fluor Daniel Fernald), 1998, 1997 Site Environmental Report, 51350-RP-0001, Fernald 

Environmental Management Project, Fernald, Ohio, June. [SRDB Ref ID: 3592]

FDF (Fluor Daniel Fernald), 1999, 1998 Site Environmental Report, 51350-RP-0005, Fernald 

Environmental Management Project, Fernald, Ohio, June. [SRDB Ref ID: 3604]

Fluor Fernald, 2000, 1999 Integrated Site Environmental Report, 51350-RP-0010, Fernald 

Environmental Management Project, Fernald, Ohio, June. [SRDB Ref ID: 3606]

Fluor Fernald, 2001, 2000 Integrated Site Environmental Report, 51350-RP-0015, Fernald 

Environmental Management Project, Fernald, Ohio, June. [SRDB Ref ID: 9010]

Fluor Fernald, 2002, 2001 Site Environmental Report, 51350-RP-0019, Fernald Environmental 

Management Project, Fernald, Ohio, June. [SRDB Ref ID: 3516]

Fluor Fernald, 2003, 2002 Site Environmental Report, 51350-RP-0022, Fernald Closure Project, 

Fernald, Ohio, June. [SRDB Ref ID: 126955]

Fluor Fernald, 2004, 2003 Site Environmental Report, 51350-RP-0024, Fernald Closure Project, 

Fernald, Ohio. [SRDB Ref ID: 126956]

Fluor Fernald, 2005, 2004 Integrated Site Environmental Report, 51350-RP-0026, Fernald Closure

Project, Fernald, Ohio. [SRDB Ref ID: 129790]

Fluor Fernald, 2006, 2005 Site Environmental Report, 51350-RP-0028, Fernald Closure Project, 

Fernald, Ohio, June. [SRDB Ref ID: 129791]

Stoller (S. M. Stoller Corporation), 2007, 2006 Fernald Site Environmental Report, DOE-LM/1461-

2007. [SRDB Ref ID: 129793]

DOE (U.S. Department of Energy), 2008, Fernald Preserve 2007 Site Environmental Report, DOELM/1607-2008, Office of Legacy Management, Washington, D.C., May. [SRDB Ref ID: 

129795]
Document No. ORAUT-TKBS-0017-4 Revision No. 02 Effective Date: 03/13/2014 Page 58 of 65

DOE (U.S. Department of Energy), 2009, Fernald Preserve 2008 Site Environmental Report, 

LMS/FER/S05082, Office of Legacy Management, Washington, D.C., May. [SRDB Ref ID: 

129801]

DOE (U.S. Department of Energy), 2010, Fernald Preserve 2009 Site Environmental Report, 

LMS/FER/S06109, Office of Legacy Management, Washington, D.C., May. [SRDB Ref ID: 

129804]

DOE (U.S. Department of Energy), 2011, Fernald Preserve 2010 Site Environmental Report, 

LMS/FER/S07409, Office of Legacy Management, Washington, D.C., May. [SRDB Ref ID: 

129807]

DOE (U.S. Department of Energy), 2012, Fernald Preserve 2011 Site Environmental Report, 

LMS/FER/S08629, Office of Legacy Management, Washington, D.C., May. [SRDB Ref ID: 

129811]

DOE (U.S. Department of Energy), 2013, Fernald Preserve 2012 Site Environmental Report, 

LMS/FER/S09665, Office of Legacy Management, Washington, D.C., May. [SRDB Ref ID: 

129812]
Document No. ORAUT-TKBS-0017-4 Revision No. 02 Effective Date: 03/13/2014 Page 59 of 65

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

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.
Document No. ORAUT-TKBS-0017-4 Revision No. 02 Effective Date: 03/13/2014 Page 60 of 65

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.
Document No. ORAUT-TKBS-0017-4 Revision No. 02 Effective Date: 03/13/2014 Page 61 of 65

ATTACHMENT A

INCORPORATION OF Q-11 SILOS AS A RADON RELEASE SOURCE, 1952 TO 1958

Figure A-1 reports the yearly maximum and mean exposures in terms of 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 Q-11 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:

Q-11 ( ) EA

d

χ C R f RF

Q

  =     (A-1)

where

CEA = radionuclide concentration in EA (Ci/m3)

RQ-11 = 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. 02 Effective Date: 03/13/2014 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

Q-11 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

Q-11 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

Q-11 12

10 L m (3.154 10 s yr ) (pCi L) (s m ) 10 pCi Ci 0.0824 d

D χ R f RF

Q

−

× ×     = × ×         (A-5)

The Q-11 silo release rate RQ-11 then reduced to:

( ) ( )

1

Q-11 0.383

d

χ R D f RF

Q

−

    = ××         (A-6)

where

RQ-11 = annual radon release rate (Ci/yr)

D = dose (WLM)

(χ/Q)d = atmospheric dilution factor at distance d (m) from source (d is a close-in 

distance of 25 m, and the (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 RQ-11 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 RQ-11 from Equation A-6 was added to the K-65 source term as 

shown in Figure 4-5.
ATTACHMENT B

95TH-PERCENTILE WIND ROSE DATA FOR 16 COMPASS POINTS

Document No. ORAUT-TKBS-0017-4 Revision No. 02 Effective Date: 03/13/2014 Page 63 of 65

Table B-1. Wind rose data and derived 95th-percentile values for environmental radionuclide concentration estimates 

(1991 to 2001 Environmental Reports).

Years

N NNE NE ENE

f u (mph) u (m/s) f u (mph) u (m/s) f u (mph) u (m/s) f u (mph) 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 ESE SE SSE

f u (mph) u (m/s) f u (mph) u (m/s) f u (mph) u (m/s) f u (mph) 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. 02 Effective Date: 03/13/2014 Page 64 of 65

Years

S SSW SW WSW

f u (mph) u (m/s) f u (mph) u (m/s) f u (mph) u (m/s) f u (mph) 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 WNW NW NNW

f u (mph) u (m/s) 0.081 5.20 2.33 f u (mph) u (m/s) f u (mph) u (m/s)

1991 0.104 5.00 2.24 0.090 4.10 1.84 0.064 4.20 1.88 0.050 5.20 2.33

1992 0.090 3.80 1.70 0.105 3.50 1.57 0.088 4.20 1.88 0.050 4.50 2.02

1993 0.112 3.40 1.52 0.097 4.40 1.97 0.075 3.50 1.57 0.040 4.80 2.15

1994 0.099 4.00 1.79 0.100 5.00 2.24 0.090 4.00 1.79 0.080 3.80 1.70

1995 0.112 4.80 2.15 0.084 5.50 2.46 0.070 4.00 1.79 0.058 5.00 2.24

1996 0.090 5.30 2.37 0.088 5.00 2.24 0.067 4.50 2.02 0.044 5.00 2.24

1997 0.118 5.20 2.33 0.079 4.50 2.02 0.075 4.30 1.93 0.047 4.70 2.11

2000 0.104 4.60 2.06 0.080 4.50 2.02 0.068 4.80 2.15 0.040 5.20 2.33

2001 0.095 4.00 1.79 0.100 5.276 2.364 0.075 5.00 2.24 0.038 5.50 2.46

95th percentile 0.114 5.161 2.312 0.081 5.20 2.33 0.085 4.774 2.139 0.064 5.312 2.380
ATTACHMENT C

ANNUAL AVERAGE RADON BACKGROUND CONCENTRATIONS AND 

CONCENTRATION AT SELECTED RADON LOCATIONS, 1989 TO 2005

Document No. ORAUT-TKBS-0017-4 Revision No. 02 Effective Date: 03/13/2014 Page 65 of 65

Figure C-1. Annual average radon background concentrations and concentration at selected radon locations, 1989 to 2005 (Environmental 

Report Fluor Fernald 2006).