ORAU Team

NIOSH Dose Reconstruction Project

Technical Basis Document for the Fernald Environmental 

Management Project (FEMP) – Occupational Internal Dose

Document Number:

ORAUT-TKBS-0017-5

Effective Date: 05/28/2004

Revision No.: 00

Controlled Copy No.: ______

Page 1 of 42

Subject Expert: Bryce L. Rich

Document Owner

Approval: Signature on File Date: 05/24/2004

Terry A. Kuykendall, TBD Team Leader

Approval: Signature on File Date: 05/24/2004

Judson L. Kenoyer, Task 3 Manager

Concurrence: Signature on File Date: 05/27/2004

Richard E. Toohey, Project Director

Approval: Signature on File Date: 05/28/2004 James W. Neton, OCAS Health Science Administrator

Supersedes:

None

TABLE OF CONTENTS

Section

Record of Issue/Revisions ...................................................................................................................

Page

3

Acronyms and Abbreviations ...............................................................................................................4

5.0 Occupational Internal Dose.........................................................................................................6

5.1 Introduction ...........................................................................................................................6

5.2 Radioisotopes of Concern .....................................................................................................8

5.2.1 Uranium Processes ......................................................................................................8

5.2.1.1 Enrichments ..........................................................................................................9

5.2.1.2 Chemical Forms and Compounds ....................................................................... 10

5.2.1.3 Airborne Dust Potential ....................................................................................... 11

5.2.1.4 Chemical Toxicity ................................................................................................ 12

5.2.2 Recycled Uranium Processes..................................................................................... 13

5.2.3 Thorium Processes..................................................................................................... 18

5.2.4 K-65 Silo Processes ................................................................................................... 23

5.3 Radiological Controls Program............................................................................................ 28

5.3.1 Comprehensive Radiological Controls Program.......................................................... 28

5.3.2 Air-Monitoring Program............................................................................................... 29

5.3.3 Bioassay Program ...................................................................................................... 30

5.3.4 Environmental Levels and Fecal Sampling Program................................................... 31

5.3.5 Analytical Program ..................................................................................................... 31

5.3.6 In Vitro Procedures for Other Isotopes........................................................................ 34

5.3.7 In Vivo Analysis .......................................................................................................... 34

5.3.8 In Vivo MDA ...............................................................................................................36

References ........................................................................................................................................37

Glossary ............................................................................................................................................41
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LIST OF TABLES

Table

5-1 Contractors during the history of the Fernald Site ....................................................................

Page

6

5-2 Primary isotopes of concern.....................................................................................................8

5-3 Uranium enrichments and associated isotopes ...................................................................... 10

5-4 Uranium chemical forms and inhalation typees at the Fernald Site ........................................ 11

5-5 Fernald plant processes, materials, and dust release potential .............................................. 12

5-6 Chemical toxicity and associated limits .................................................................................. 12

5-7 Comparison of percent total alpha activity with 10 ppb plutonium-239.................................... 13

5-8 Derived air concentrations from the 1989 RU contaminant analysis....................................... 14

5-9 Recycled uranium summary values by process subgroupings ............................................... 15

5-10 Plant history average RU contaminant radionuclides ............................................................. 16

5-11 PPB conversion factors.......................................................................................................... 17

5-12 Percent of total uranium activity added to air-monitoring samples ........................................ 18

5-13 Thorium production estimates (MT) by year, with uranium production.................................... 20

5-14 Thorium plant processes with chemical forms and solubility classes...................................... 20

5-15 Estimated thorium emissions (kg) in comparison with uranium .............................................. 21

5-16 Istopic comparison of K-25 Silos 1 and 2 ............................................................................... 24

5-17 Air monitoring analysis sheet codes ....................................................................................... 29

5-18 Cross-reference of FEMP air-monitoring methods with the applicable MDLs ......................... 30

5-19 Routine uranium in vitro bioassay capabilities at the FEMP ................................................... 30

5-20 Uranium urinalysis sample type codes ................................................................................... 32

5-21 FEMP sample type coding system and personnel assignment codes .................................... 33

5-22 MDAs for isotopes other than uranium ................................................................................... 34

5-23 Urinary uranium averages summary ...................................................................................... 34

5-24 Typical reporting of internal dose determined by direct lung count ......................................... 33

5-25 Typical MIVRML counting schedule at Fernald in the 1970s .................................................. 35

5-26 Uranium in vivo MDAs for the lung ......................................................................................... 36

LIST OF FIGURES

Figure

5-1 Ingrowth of 232Th decay products ...........................................................................................

Page

19
Effective Date: 05/28/2004 Revision No. 00 Document No. ORAUT-TKBS-0017-5 Page 3 of 42

RECORD OF ISSUE/REVISIONS

ISSUE 

AUTHORIZATION 

DATE

EFFECTIVE 

DATE

REV. 

NO. DESCRIPTION

Draft 01/29/2004 00-A New Technical Basis Document for the Fernald 

Environmental Management Project (FEMP) –

Occupational Internal Dose. Initiated by Bryce L. 

Rich.

Draft 04/14/2004 00-B Incorporates comments from OCAS and Task 5 

reviews. Initiated by Bryce L. Rich.

Draft 05/14/2004 00-C Incorporates additional comments from OCAS 

review. Initiated by Bryce L. Rich.

05/28/2004 05/28/2004 00 First approved issue. Initiated by Bryce L. Rich.
Effective Date: 05/28/2004 Revision No. 00 Document No. ORAUT-TKBS-0017-5 Page 4 of 42

ACRONYMS AND ABBREVIATIONS

AEC U.S. Atomic Energy Commission

BZ breathing zone

cm3 cubic centimeter

CEDE cumulative effective dose equivalent

DAC Derived Air Concentration

DL decision level

DOE U.S. Department of Energy

dpm disintegrations per minute

ERDA Energy Research and Development Agency

FCP Fernald Closure Project

FEMP Fernald Environmental Management Project

FERMCO Fernald Environmental Restoration Management Corporation

FMPC Feed Material Production Center

GA general air sample

hr hour

ICP-MS Inductively Coupled Plasma Mass Spectrometry

ICRP International Commission on Radiological Protection

IVEC In Vivo Examination Center

KPA kinetic phosphorescent analysis

LEU low enriched uranium

m meter

MAC maximum allowable concentration

MCW Mallinckrodt Chemical Works

MDL minimum detection level

mg U milligrams of uranium

mg milligram

MIVRML Mobile In Vivo Radiation Monitoring Laboratory – ORNL Y-12 

MPBB maximum permissible body burden

MPC maximum permissible concentration

MPLB maximum permissible lung burden

MT metric ton

MTU metric ton of uranium

NCG National Lead Concentration Guide (used interchangeably with MAC)

nCi nanocurie

NLO National Lead of Ohio, Inc.
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ORO Oak Ridge Operations Office

pCi picocurie

PGDP Paducah Gaseous Diffusion Plant

POOS plutonium out of specification

ppb parts per billion

RU recycled uranium

s second

SRS savannah river site

TBD Technical Basis Document

TRU Transuranic

U.S.C. United States Code

UCL upper confidence limit

UNH uranyl nitrate

UNSCEAR United Nations Scientific Committee on Energy and Atomic Radiation

WEMCO Westinghouse Environmental Management Company of Ohio, Inc. 

WL Working Level used for Radon levels = 1.3 x 105 Mev/L

yr year

μCi microcurie

μg microgram
Effective Date: 05/28/2004 Revision No. 00 Document No. ORAUT-TKBS-0017-5 Page 6 of 42

5.0

This section provides the technical basis for estimation of occupational internal radiation dose to 

workers from the U.S. Department of Energy (DOE) Fernald Environmental Management Project 

(FEMP; previously known as the Feed Material Production Center, or FMPC).

OCCUPATIONAL INTERNAL DOSE

This Technical Basis Document (TBD) provides information and assumptions for use in reconstructing 

employees’ occupational internal radiation doses. Specific characteristics of the monitoring 

procedures, identification of events or processes that were unmonitored; identification of the types 

and quantities of radioactive materials involved, evaluation of production processes and safety 

procedures, identification of the locations and activities of exposed persons, and identification of 

comparable exposure circumstances for which data is available on which to base assumptions.

5.1 INTRODUCTION

The discussions in this introductory section derive from information from a variety of references, 

primarily the DOE Ohio Field Office Recycled Uranium Project Report (DOE 2000), Highly Enriched 

Uranium Working Group Report (DOE 1996), and Dolan and Hill FMPC-2082 (Dolan and Hill 1988). 

These reports summarize the efforts of teams of process engineers and operations specialists (all 

with long-term plant experience) in reconstructing historical plant operations and effluent experience, 

and represent the best information available. The information provided in this section includes 

operational documentation and professional judgment applied to fill in areas where data was missing 

or inadequate.

Construction at FEMP began in 1951 on a 1,050-acre site near the small rural community of Fernald, 

Ohio. Table 5-1 lists the three contractors to DOE and its predecessor agencies [i.e., the U.S. Atomic 

Energy Commission (AEC) and the Energy and Research Defense Agency] that have operated 

FEMP.

Table 5-1. Contractors during the operational history of the Fernald Site.

Contractors Dates Time

National Lead Co. of Ohio (NLO), Inc. October 1951 to January 1986 34+ yr

Westinghouse Materials Company of Ohio (WEMCO) January 1986 to December 1992 7 yr

Fernald Environmental Restoration Management 

Corporation (FERMCO, Fluor Fernald, Inc)*

December 1992 to present +11 yr

* FERMCO received the first DOE Environmental Restoration Management contract for FEMP.

The original health and safety program was conducted with an industrial hygiene emphasis, based 

upon uranium heavy metal toxicology, with awareness and utilization of technology from other AEC 

site programs (e.g., Oak Ridge Y-12 programs). Basic changes in the radiological protection program 

occurred with the contractor change in 1986 with the addition of radiation safety staff and a greater 

emphasis on radiation protection principles.

The primary missions at FEMP were the processing of uranium ores to high-quality finished metal 

products and the production of thorium metal parts and thorium feedstock for processes at other 

AEC/DOE sites. The 10 plants that constituted the Fernald facility processed a variety of natural and 

low enrichment uranium ores and scrap materials for the creation of high-purity uranium metal parts. 

The thorium production included metal parts and feedstock for the weapons production programs. 

Operations involved thousands of metric tons of ores, dry powder products, and corrosive chemicals 

in processes that were inherently dusty, producing an environment with internal intake potential.
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Due to the inherent nature of the processes, the limitations of the ventilation and material confinement 

systems, and the volume (and mass) of the materials, significant environmental and in-plant releases 

of radioactive materials occurred during FEMP operations. The work environs included a 

continuous/chronic potential for internal exposure, as demonstrated by the comprehensive air 

monitoring program and the urine sampling program for uranium (documented by air sample and 

urine uranium data sheets). During the early years, plant workers were routinely required to wear 

respiratory protection because of significant radioactive dust levels approaching or exceeding FEMP 

airborne alpha activity guidelines referred to as maximum allowable concentrations (MAC) [or National 

Lead of Ohio Concentration Guide (NCG)] depending on the period (the Lead referred to the name of 

the contractor, not the contaminant). In addition, over the operational history of Fernald, metric tons 

(MT) of uranium and thorium products were released from the ventilation stacks to the environment.

The number of personnel at Fernald peaked at 2,891 in 1956 and slowly declined to 538 in 1979; an 

estimated 7,300 personnel worked at FEMP through 1990. Given the site conditions of routine 

measurable air activity in the workplace and significant routine releases to the surrounding areas, all 

of these personnel had some potential for internal exposure from either being in process or other 

areas associated with loose contamination, or as a result of environmental releases. Additional

radionuclides were introduced from other feedstock sources: radium and thorium from pitchblende 

ores from the Belgian Congo, transuranics and fission products from recycled uranium, and thorium. 

Pitchblende ores from the Belgian Congo, which have unusually high radium and thorium activities, 

were processed from 1953 to 1955 as part of the uranium ores processing program. The liquid 

extraction column raffinates from the uranium extraction process were stored in two dedicated 

concrete silos (by agreement, they belonged to the Belgian government). Additional pitchblende 

residues from another AEC site (Mallinckrodt Chemical Works) were added to the silos starting from 

the time of completion of the silos in 1953 and continuing until 1958. The materials in these silos 

became the property and responsibility of the United States when the lease agreement expired in 

1983. This material remains on the Fernald site in the K-65 Silos and represents a continuing internal 

exposure potential (for any operations in which direct contact with these residues is required) from the 

unusually high concentrations of 226Ra and its progeny, 210Pb and 210Po (these three radionuclides 

comprise approximately 90% of the total activity in the K-65 materials). The uranium daughter 230Th is 

also present in significant quantities (7 to 10%) in this location (see Table 5-16). Historical 

environmental releases from this location, with the resultant internal dose potential, occurred from the 

ingrown radon and thoron gases and their daughter products. 

Fernald processed thorium during from 1954 to 1979 and was the national thorium materials 

repository for the DOE starting in 1972. About two-thirds of the material in the repository was 

processed and stored at Fernald, while the remainder originated at other DOE facilities. The thorium 

was stored as metal and stable chemical compounds [e.g., thorium gel, Th(OH)4] containing primarily, 

232Th and its long-lived daughter 228Th. In addition, the beta emitter 228Ra was present. The original 

thorium inventory was contained in approximately 15,000 storage containers of various sizes. In 

2002, the inventory was verified to be approximately 500,000 pounds stored in approximately 200 

metal boxes (Tomes 2001). Internal thorium exposure potential is associated with all of the thorium 

processes and during handling and repackaging of the thorium in storage drums.

The initial uranium processing started in 1951 with virgin stock from uranium mines and mills. 

Recycled uranium (RU) was received and processed at Fernald as early as 1961. Recycled uranium 

is that which has been used in a reactor environment (such as fuel elements or transuranic production 

targets) and then processed to separate the unspent uranium from fission products, activation 

products, and transuranic elements. RU contains trace quantities of plutonium isotopes, 237Np, and 

99Tc as the primary contaminants. Some recycled materials from the uranium enrichment facilities (in 
Effective Date: 05/28/2004 Revision No. 00 Document No. ORAUT-TKBS-0017-5 Page 8 of 42

the form of tower ashes) had higher than average transuranic contaminant concentrations. The levels 

of contaminants in these RU materials were known before receipt and blended with other uranium 

feed stock for processing. Personnel exposed to uranium contamination could also be exposed to the 

RU contaminants, which could have contributed to unmonitored internal exposure.

5.2 RADIONUCLIDES OF CONCERN

Table 5-2 lists the primary radionuclides that could have led to internal doses during the production 

history of FEMP:

Table 5-2. Primary radionuclides of internal exposure concern.

Uranium & Daughters Recycled Uranium 

Contaminants

Thorium & Daughters

Isotopes Daughters Isotopes Daughters

U-234 Th-230 Pu-238 Th-232 Ra-226

U-235 Ra-226 Pu-239 Th-230* Rn-222

U-236** Rn-222 Pu-240 Ra-228

U-238 Pa-231 Pu-241 Th-228

Ac-227 Am-241*** Ac-228 a

Rn-222 Np-237 Pb-212 a

Tc-99 Ra-228 a These Thorium daughters are listed, since they are used in In vivo counting to quantify the parent Th-232 

They do not contribute to the internal dose above 1% 

*Present primarily as a uranium decay product. 

**Though listed as an expected RU component, U-236 represents <1% of the dose resulting from 

exposure to uranium. 

***Recognized as a potential TRU contaminant in RU and detected with sophisticated analytical 

techniques in the plant after 1989, but not in personnel and <1% of RU contaminants

5.2.1

The FEMP production facilities consisted of ten production plants, each of which produced a product 

for the next process in a series of steps, terminating in high-purity uranium metal stock or parts. 

Processes

The Pilot Plant, as an operational prototype of the entire production process, began operations in 

October 1951. The purpose was to develop performance data for designing large-scale equipment for 

the rest of the plant. At the same time, limited quantities of uranium metal were produced during the 

initial Pilot Plant operation. Also, UF6 was converted to UF4. On February 14, 1966 an accidental 

release of approximately 1200 kg of uranium occurred during conversion operations that resulted in 

elevated personnel exposures (Warner 1966). The enrichment is unknown and 2% should be 

assumed for any claimant identified as a subject of this incident. In addition, more than 70% of the 

thorium at FEMP was handled and processed from 1964 to 1979 in the Pilot Plant. 

Plant 1, the Sampling Plant, started operations in December 1953 and was later designated the 

official AEC sampling station for uranium and isotopic assays of uranium ores and concentrates 

(usually U3O8). The sampling process frequently involved the dusty operations of mixing, blending, 

etc., of large quantities of uranium and thorium materials.

Plants 2 and 3, the Ore Refinery Plants, began operations in December 1953 and achieved full-scale 

production early in 1954. The chemical processing in these facilities involved a three-step operation 

that began with nitric acid leaching of uranium from dry solid feed materials. This was followed by 
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solvent extraction processing using a modified PUREX1

Plant 4, the Green Salt Plant, which converted UO3 to UF4 (green salt), began in October 1953 and 

achieved full-scale operation in 1954. The two-step process involved hydrogen reduction of UO3

powder to UO2 solids and then to UF4 by the addition of anhydrous hydrofluoric acid. These 

processes mixed, agitated, and transferred metric tons of uranium (MTU) solids and produced some 

of the higher levels of airborne radioactivity at the Site. Some air sample data sheet information 

indicates that at least a limited amount of thorium may have been processed in Plant 4.

process to produce a high-purity solution of 

uranyl nitrate (UNH). The final step was a high-temperature thermal decomposition of the UNH 

solution to produce UO3 product in dry, solid form and nitric oxides as effluent. After startup, these 

two plants generally operated as a single plant due to the integral process of the operations. During 

operations, the Plant 2/3 complex was a single large plant (2A), which was surrounded by several 

small buildings and tank farms that performed various support functions, e.g., the Hot Raffinate 

Building (3E) was utilized to filter insolubles from UNH and to process raffinates.

Plant 5, the Metals Production Plant, provided a chemical conversion of UF4 powder to a uranium 

metal “derby” by a thermite furnace reduction process with magnesium metal. This facility began 

operations late in 1953. This process concentrated the very small concentrations of remaining

transuranic (TRU) elements and uranium daughters in the MgF2 slag. The transport and mixing of 

feed materials and MgF2 slag led to inhalation exposure potential that were also in the higher level 

categories at the Site.

Plant 6, the Metals Fabrication Plant, and Plant 9 produced metal parts in rolling mills and machining 

lathes. Plant 6 began operations in 1953. Uranium metal fires were common, resulting in elevated 

airborne uranium concentrations.

Plant 7, the Hexafluoride Reduction Plant, was a chemical process facility operated from 1954 to 1957 

that provided an additional source of UF4 by hydrogen reduction of UF6. UF6 was converted by 

temperature control to the gaseous phase, which increased the potential for uranium inhalation 

exposures.

Plant 8, the Scrap Recovery Plant, provided a scrap uranium recovery function. Plant 8 began 

operations in 1954 and was also ranked in the higher air activity level areas at the Site. Thorium 

scraps and residues were processed in 1966, 1969, 1970, and 1971.

Plant 9, the Special Products Plant, included casting of uranium metal and high-purity recycled metal 

scraps into ingots. Operations began late in 1953. Thorium was processed as metal and briquettes 

in 1954 and 1955.

5.2.1.1 Uranium Enrichments

Late in 1964 the Fernald site provided the first production of 1.95% 235U billets for the Hanford Site. 

During the following production years uranium was processed in a variety of enrichments ranging from 

depleted to as high as 20%. The quantities of enriched material above 2% was not documented, but 

was qualitatively reported to be small and/or insignificant in total mass. The reported highest 

enrichment level processed in quantity was 2%. As another point of reference, a 1961 to 1984 history 

(Dolan and Hill 1988) of the average uranium enrichments in dust collector stack discharges 

demonstrates that 235U enrichments ranged from 0.20% to 1.68% with an average of 0.70%. 

 1

 The PUREX process is a chemical extraction technology for the separation of radionuclides from uranium and was used at 

several DOE sites in a variety of applications.
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Of the total quantity of uranium received and processed at FEMP <25% was enriched above normal 

(60,181 MTU of the total 246,683 MTU). Approximately 95% (208 gms of the total 218 gms) of the 

Pu-239 which was received at the Site came in the enriched uranium receipts. (DOE 2000)

Table 5-3 lists the primary assumptions for FEMP uranium enrichments and the isotopes associated 

with these enrichments. The mass percentages, relative activities in pCi/μg, and the total pCi/μg 

values are based upon IMBA NIOSH default values.

Table 5-3. Uranium enrichments and associated isotopes.*

Category Isotopes Mass %

Relative activity 

(pCi ug-1)

Activity

%

Total

pCi ug-1

Uranium–depleted U-234

U-235

U-236

U-238

0.001

0.20

0.00031

99.80

0.062

0.004

0.0002

0.3354

15.46

1.07

0.05

83.42

0.402

Uranium–natural U-234

U-235

U-236

U-238

0.0054

0.7204

0

99.274

0.334

0.016

0

0.334

48.86

2.28

0

48.86

0.683

Uranium -1% enriched

(EU assumption used 

in early In vivo 

calculations)

U-234

U-235

U-236

U-238

0.01

1.0

0

98.99

0.62

0.02

0

0.333

63.7

2.1

0

34.2

0.973

Uranium–2% enriched

(Recommended dose 

evaluation default for 

this TBD)

U-234

U-235

U-236

U-238

0.02

2.0

0

97.98

1.243

0.043

0

0.329

76.9

2.68

0

20.4

1.616

*U-236 is less than 1% activity in DU, Ntl, 1% EU, and 2% EU

In the absence of specific enrichment information, and considering the above available data related to 

processing experience of uranium enrichments at FEMP, the default assumption for time periods after 

1964 is 2% enrichment for bioassay data in milligram quantities of uranium. Prior to 1964 natural 

uranium should be assumed.

5.2.1.2 Chemical Forms and Compounds

There are approximately seven steps in the process of conversion of uranium ore or other scrap 

recovery materials to metallic uranium. Those steps produce a number of compounds, each of which 

has specific chemical characteristics that are associated with different internal exposure parameters. 

Each of the compounds identified in Table 5-4 was handled in MT quantities. Most of the compounds 

were dry powder or granular in form and represented a dust hazard potential as the material was 

processed, transferred, and otherwise handled. 
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Table 5-4. Uranium chemical forms at the Fernald Site.

Uranium Compound

Chemical 

Formula

Uranium hexafluoride (gas) UF6

Uranyl fluoride UO2F2

Uranyl nitrate UO2(NO3)2

Uranium trioxide (orange oxide) UO3

Uranium tetrafluoride (green salt) UF4

Uranium dioxide (brown oxide) UO2

Uranium tetraoxide UO4

Uranium oxide (yellow cake) U3O8

High fired uranium oxides UO2

Uranium metal (exposure generally 

from an oxide)

U

5.2.1.3 Airborne Dust Potential

Production operations that involved handling dry uranium materials were generally equipped with 

engineered ventilation systems for controlling dusts. Standard operating procedures required the use 

of respiratory equipment when dusty conditions were anticipated. Good housekeeping involving the 

immediate cleanup of spilled uranium products was also a standing policy and practice. In spite of 

this emphasis on engineered and administrative contamination controls and policy to reduce the 

release of radioactive materials, spills and routine releases occurred. In addition to the routine 

releases at FEMP, there were frequent “upset” conditions (i.e., spills, effluent filter ruptures, etc.) that 

produced episodic airborne radioactivity in the work areas and plant effluents, and were of a 

magnitude that the ventilation systems were unable to contain all of the releases.

In 2000 a FEMP team working on the DOE Ohio Field Office Recycled Uranium Project Report (DOE 

2000) qualitatively rated various plant processes in relation to the potential for producing airborne 

dusts in high, medium, or low categories, based on the criteria below. Though qualitative, this 

evaluation was consistent with historical FEMP air activity measurements and recorded internal 

exposures. The evaluation involved the following ratings and the qualitative descriptions of “very 

large quantities”, “small quantity, intermittent” and “highly reliable” are used as quoted in the 

reference:

• High ratings were assigned for operations that processed very large quantities of materials 

using processes or equipment susceptible to occasional events. Examples of such events 

include digestion tank fuming and denitration pot eruptions in Plant 2/3, failure of 

hydrofluorination banks in Plant 4, temperature excursions in Plant 8 furnace operations, and 

failure of ventilation system dust bags during both operation and maintenance. The ventilation 

systems were inadequate to contain the dust emissions from these events, and all plant 

workers were susceptible to exposure from these releases.

• Low ratings were assigned for small quantity, intermittent operations, those that did not involve 

appreciable amounts of dry materials, and those considered highly reliable.

• Medium ratings were assigned for operations that were neither of clearly high nor clearly low 

radioactive material release potential.

Table 5-5 lists the summary results of this evaluation including information related to the typical 

materials. 
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Table 5-5. Fernald plant processes, materials, and dust release potential.

Plant Process Compound* Airborne Dust Potential (qualitative)

5 Metal reduction & casting UF4, U, U3O8, Residues High

8 Feed Preparation Furnaces U3O8, U, Residues High

4 Banks 7-9, Packing Stations UO3, UO2, UF4 High

2/3 Digestion, Extraction Denitration Prepared Feed, U3O8, 

UO2, UO3

Medium

9 Reduction, Casting UF4, U, U3O8, Residues Medium

Pilot Hex Reduction, Metal Operations UF4, U, U3O8, Residues Medium

1 Milling U, U3O8, Residues Low

* MgF2 (one of the residue products) in Plants 1, 2/3, 5, 9, and the Pilot Plant, contributed to the dust potential rating, but 

did not carry more than trace quantities of radionuclides. The MgF2 typically was associated with nonuranium radionuclides 

(i.e., uranium daughter products and RU contaminants).

5.2.1.4 Chemical Toxicity

Uranium is an unusual radioactive material in that its chemical toxicity can be the dominant hazard in 

cases of readily or moderately soluble compounds of depleted, natural, and low-enriched uranium. 

Uranium is a heavy metal and a sufficient heavy metal exposure can lead to kidney damage.

In 1959, a chemical toxicity threshold concentration of 3 μg of uranium per gram of kidney tissue 

(based upon animal studies and some human data) was established by the International Council on 

Radiation Protection and Measurements (ICRP), and has been used since that time as the reference 

limit. Concentrations in excess of 3 μg of uranium per gram of kidney tissue have been shown to 

result in renal tubule damage. The no-effect level of uranium in the kidney has been calculated to be 

1.1 μg of uranium per gram of kidney tissue, which averages to a kidney burden of 0.337 mg (Rich et 

al. 1988). Table 5-6 lists a summary of the FEMP chemical toxicity limits with their associated 

radiological limits.

Table 5-6. Chemical toxicity and associated limits.

*

Solubility –

Enrichment

mg intake = approx. 0.337 

mg kidney – toxicity limit

OSHA Limit μCi cm-3 limit

mg m based on rad. exp. -3 (toxicity) Equiv. rad. (μCi cm-3

)

Sol. – depleted Type F – 7 mg

Type M – 37 mg

Type S – 696 mg

0.05 1.8E-11 6.0E-10

Sol. – natural 0.05 3.5E-11 3.0E-10

Sol. – 2% enriched 0.05 8.1E-11 2.0E-11

Insol. – depleted 0.25 9.2E-11 6.0E-10

Insol. – natural 0.25 1.8E-10 3.0E-10

Insol. – 2% enriched 0.25 1.8E-10 3.0E-10

* Italicized entries indicate limits that are controlled by radiological concerns.

The early basis for conducting routine urine analysis was to assure that uranium exposure controls 

were adequate to prevent chemical toxicity. Additionally the urinalysis results were used to restrict 

further worker exposure when control levels were exceeded. Urine samples were taken as frequently 

as weekly (or more frequently as indicated by unusual events). Internal radiation doses, derived from 

urine sample results, were not calculated until the mid to late 1980s.

Though the primary exposure control was based on toxicity concerns, some radiological exposures 

were reported in addition to a few work restrictions based upon radiological dose limits. The 

radiological dose determinations were based upon In vivo lung counting. In vivo lung counts were 

routine from 1968 to 1989 using a Mobile In Vivo Radiation Monitoring Laboratory (MIVRML) from 

ORNL Y-12. From 1989 to 2001 the lung counts took place in the Fernald In Vivo Examination Center 
Effective Date: 05/28/2004 Revision No. 00 Document No. ORAUT-TKBS-0017-5 Page 13 of 42

(IVEC). The results were reported in milligrams of 235U and total milligrams of uranium (mg U) as well 

as in percent of the maximum permissible lung burden (MPLB) or maximum permissible body burden 

(MPBB). The in vivo counting results were recorded in individual dosimetry files and summarized in 

formal FEMP reports. The in vivo lung counts were preferentially used at FEMP to determine the 

MPLB percentages. 

5.2.2

The first uranium feedstock for the Fernald plant in 1952 and 1953 was yellow cake (U3O8) and 

pitchblende from the Belgian Congo. On February 13, 1961, the first RU was introduced to Fernald in 

a shipment of UF6 that was to be converted to UF4. During the next 3 years, RU was introduced to all 

of the processes with the plutonium concentrations remaining below the established specifications of 

10 ppb of uranium on a mass basis (Bihl 2003). It was calculated that 218 grams of Pu239, 19,048

grams of Np237, and 328,740 grams of Tc99 was received at FEMP in 246,683 MTU of RU during the 

history of the plant. (DOE 2000). In 1976 the decision was made to process tower ash and other 

residues from the Paducah Gaseous Diffusion Plant (PGDP), which introduced RU contaminants with 

Plutonium-Out-of-Specification (POOS), i.e. above the 10 ppb established specification. This material 

was both 1) sent through the solvent extraction process and 2) blended to an oxide. The PGDP 

shipment accounted for a major portion of the TRU inventory at the Site. (Bassett et al. 1989). The 

solvent extraction process concentrated TRU in the raffinate, while the blended oxide resulted in 

storage of uranium with a maximum concentration of Pu of 43 ppb U. The total historical average 

FEMP concentrations of RU contaminants in the 246,683 MTU (RU) is approx. 1 ppb Pu239, 104 ppb 

NP237, and 1344 ppb Tc99 of RU. 

Recycled Uranium (RU) Processes

The RU was uranium that had been recovered from irradiated production reactor fuel and plutonium 

production fuels. The RU was separated in the chemical processing plants at Hanford and Savannah 

River Site, but was known to contain traces of TRU elements and fission product impurities. Most of 

the RU delivered to Fernald originated from the Hanford Site, PGDP, and the Savannah River Site. 

The primary contaminants were 238Pu, 239Pu, 241Am, 237Np, and a fission product, 99Tc. The 239Pu, 

237Np, and 99Tc were the radionuclides of greatest concentrations and were tracked and documented 

for control purposes.

The levels of contaminants were recorded in parts per billion on a uranium mass basis. Limits were 

set for Hanford product at 10 ppb total plutonium (Bihl 2003). No limits were specified for 237Np or 

99Tc. The established 10 ppb plutonium limit resulted in radiological impurities which did not exceed 

0.1% of the alpha activity from uranium. Table 5-7, which lists the actual activity percentages, shows 

that the activity varied with enrichment.

Table 5-7. Comparison of percent total alpha activity with 10 and 100 ppb plutonium-239.

Uranium 

Enrichment

Uranium Enrichment 

Activity (pCi μg-1)

Pu-239 (pCi μg-1) at: % Pu-239 of Total U Alpha Activity at:

10 ppb 100 ppb 10 ppb 100 ppb

Depleted 0.402 6.175E-04 6.175E-03 0.154 1.54

Natural 0.683 6.175E-04 6.175E-03 0.09 0.90

2% enriched 1.616 6.175E-04 6.175E-03 0.04 0.38

POOS feed stock was sent to Fernald for reprocessing with the prior knowledge that the material was 

above the routine contamination limits. The intent was to blend these materials with the inventories at 

the plant and still maintain the plant-wide limits that had been determined to provide an adequate level 

of safety, i.e., staying below the average site wide plutonium specification of 10 ppb. Workers 

handling the POOS materials directly (Plant 1 and other locations) were protected with airline 

respiratory equipment, particularly for the 1976 shipment of tower ashes from the PGDP. 
Effective Date: 05/28/2004 Revision No. 00 Document No. ORAUT-TKBS-0017-5 Page 14 of 42

In addition, several of the uranium purification processes resulted in concentrating the TRU 

contaminants in waste streams. Formal investigations that traced the TRU contaminants through the 

various plant processes were conducted; one such study defined contaminant affinity for the metal 

production furnace mold materials in 1975 (some TRU studies are referenced in DOE 2000). These 

analyses demonstrated that the extraction and metal conversion processes concentrated plutonium, 

neptunium, and uranium daughter product contaminants in the raffinate waste of the liquid extraction 

columns and in the MgF2 from the thermite metal conversion furnaces. MgF2 slag from the thermite 

metal reduction process was also processed through the chemical extraction plant. These 

concentration points or areas were identified by analytical methods in early years. The discussions in 

this TBD section make use of studies of the data from those early reports (DOE 2000). 

The Fernald Health and Safety staff was aware that TRU and fission product contaminants were 

present in the recycled uranium, although they believed it to be <0.1% of the total alpha activity. 

Therefore the biological concerns were thought to represent an increase of less than 10 to 15% in the 

hazard level. Radiological control was administered continually on the basis of both air analysis 

(gross alpha counts) and urine bioassay for uranium. Before 1989 no TRU analyses for radiological 

safety were performed on a routine basis for either airborne or urine activity, and exposure controls 

remained based on chemical toxicity under the assumption that these controls would be sufficient for 

all the radiological issues (Bassett et al. 1989). Although the alpha activity from the TRU alpha 

emitters would have been collected and detected on the air samples, the reported results were all 

considered to be uranium and compared to the MAC.

A review of the Oak Ridge Y-12 radiation protection program and internal dosimetry technology 

(Cofield 1959a; Scott 1964; Steckel and West 1966; West 1979) indicates that the internal dose 

technology, techniques, procedures, and philosophy similar to Y-12’s were used at Fernald. The fact 

that the Oak Ridge Y-12 MIVRML provided routine service tp the Fernald in vivo internal dosimetry 

program is an indicator of the close working relationship the Fernald staff had with the Y-12 program. 

However, this technology during the years until 1986 did not provide adequate detection for TRU or 

Thorium. Typical results of the MIVRML were uranium lung burdens, although a few thorium analyses 

results were reported at specific times. Though TRU analyses were attempted, the limitations of the 

MIVRML, which included limited detection sensitivities for TRU isotopes and infrequent counts, 

presented capabilities that lacked the ability to detect the anticipated levels at FEMP or in fact could 

not detect levels that met regulatory limits for TRU isotopes. 

A more formal program of internal dosimetry was introduced in 1986. An immediate emphasis was 

placed on evaluation of the transuranic materials and fission products. From April 1986 to 1989 the 

site collected 675 plutonium bioassay samples from 441 workers. Only 10 individuals exhibited 

quantities of plutonium in the urine above 0.02 dpm/sample (this contractual MDL was determined to 

be unrealistically low and adjusted to the ANSI 13.30 recommended MDL of 0.13 dpm/sample . 

Further investigation (follow up samples) revealed that 3 of the 10 were below the specified detection 

limit of 0.02 dpm per sample. All of the 10 individuals were flown to Hanford, where in vivo counts 

were conducted for plutonium. All results were below the detection limit of the Hanford In vivo 

analyses. The conclusion was that the seven of ten were technically at the urine MDL, although so 

close that the results were questionable false positive. Regardless, the directions were to record the 

results in the individual dosimetry records and perform a dose estimate calculation. The dose 

calculation results were not found that would serve the purposes of this report, but could appear in 

individual claimant files (Bassett 1989). The FEMP radiation protection practices were adjusted to 

account for those trace isotopes within the DOE radiation protection standards in effect at the time 

(e.g., DOE 1987).
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Before February 1989, no smears or air sampling filters were analyzed specifically for plutonium, 

neptunium, or thorium isotopes (Basset 1989), although these radionuclides would have been 

detected by gross alpha counting. In 1989, several sets of air and surface smear samples from Plants 

4 and 8 were analyzed for total uranium, 239Pu/240Pu, 238Pu, 237Np, 228Th, 230Th, and 232Th. Table 5-8 

lists the results in derived air concentrations (DAC) from these analyses. DAC ratios are used not for 

dose reconstruction purposes, but only to indicate the measured relative levels of TRU contaminants 

in the plant in the 1989 time period, which in turn lends credence to the default assumptions in Table 

5-8 for accounting for the unmeasured TRU in the plants.

Table 5-8. Derived air concentrations from the 1989 RU contaminant analyses.

Isotopes Derived Air Concentrations

Pu-238, 239, 240 0.1-1% of Type S

Np-237 <0.03% to 0.1% of Type S

Th-228, 232 Ranged up to 3% of the Type S

Th-230 6 to 10% of Type S

Note: These initial results indicated that for those samples where the activity fraction of each radionuclide could be 

compared (generally those with a total alpha activity exceeding E-12 μCi/mL) uranium was the controlling activity in air 

samples. In addition the results were reported in percentage of the Derived Air Concentration limits in effect at the time, 

which were also in the range of E-12 μCi/cc.

During the next several years a major sampling and analytical program examined hundreds of 

process samples, air samples from the process areas, and biological samples for plutonium, 

neptunium, thorium, technetium, and uranium. Early in 1989 an in vivo counting facility was 

constructed on the Fernald site to replace the mobile in vivo counting facility from Y-12. 

The process data were used by several FEMP teams to reconstruct the historical RU process 

experience, starting in about 1985 or 1986; the most recent report was issued May 15, 2000 (DOE 

2000). The movement of recycled contaminants and calculated levels were exhaustively 

reconstructed from records of recycled uranium inventories, RU contaminant analyses in each plant 

from analytical data from the 1970s and 1980s, and process knowledge of professional staff 

members, who had been at the plant throughout most of its history (DOE 2000).

Process subgroups were established to guide the reconstruction as indicated in Table 5-9, which lists 

a summary of the RU contaminants based on the different process/chemistry that the material 

encountered in the indicated process subgroups. For internal dose reconstruction, this table may be 

used to develop activity ratios in the absence of specific bioassay data.

Table 5-9. Recycled uranium summary values by process subgroups.

*

Subgroup # Description Pu-239 ppb U Np-237 ppb U Tc-99 ppb U

1A Miscellaneous 16.035 1328.11 2399.22

1B Miscellaneous – Minor Offsite 0.889 109.07 0.55

2 UF6 Source UF4 (GDP Tails) 0.502 54.90 201.61

3 UF6 source metal & scrap 0.007 2.54 9.12

4 Normal U product, residues & scrap 0.091 67.09 26.55

5 Enriched UF6 source products/Res. 1.259 81.39 2109.61

6A UO3 Purex source (A508)(Unblended) 2.884 388.97 8552.23

6B LEUproductsA508UO3/UF4 (low cross) 2.321 332.94 8934.58

6C LEUproductsA508UO3/UF4(highcross) 23.969 1045.29 2789.56

6D A500 Coded Enriched Residues 4.556 143.75 1085.45

6E SR UNH 16.527 -- --

6F SR UO3 – not shipped to FEMP 2.805 -- --

7A A508 based derbies 9.305 311.97 1721.00
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Subgroup # Description Pu-239 ppb U Np-237 ppb U Tc-99 ppb U

7B A508 based ingots & Metal 1.165 263.48 447.81

8 Enriched MgF2 96.618 1881.53 1651.23

9 Incinerator ash& scrap residues-GDPs 47.616 3164.53 263.32

10A Tower ash & decon. Residues 412.177 10503.53 2618.36

10B UO3 from tower ash 20.772 498.17 2405.28

11 Waste residues 84.817 3999.32 4110.05

* Though the results in the table are all reported in ppb U, this measure is meaningless in subgroups in which there is very 

little uranium, such as subgroup 8, in which the MgF2 did accumulate some isotopes, but was low in uranium by design.

Because the data from the many process subgroups did not conform well to normal or lognormal 

distribution patterns, the presentation of the data was not straightforward. In cases of data skew, the 

use of a “bootstrap” analysis is employed, which is a relatively recent form of analysis that employs a 

sampling approach to approximate representative values for the data set (DOE 2000). Table 5-10 

lists the average contaminant concentrations on a site wide basis over the nearly 38-yr RU history 

(1961-1999) of FEMP. Note: Processing of a single project of RU started at FEMP in February 1961 

and was followed in July 1962 with start-up of WINLO process. Processing continued until 1989. 

Receipts are listed through 1999.

Table 5-10. Plant RU history with 38-year average RU contaminant radionuclides.

Uranium 

Enrichm

ents

MTU 

Receipts

Percent 

Uranium 

Receipts 

Bootstrap Mean Calculations

Total Pu239 (gm)Pu-239 

(ppb)

Total Np237 (gm)Np-237 

(ppb)

Total Tc-99 

(gm)

Tc-99 

(ppb)

Enriched 60,180.7 24.4 207.9 3.5 19,047.5 316 328,740 5,382

Normal 89,649.2 36.3 4.1 <0.1 3,025.9 34 1,197.4 13

Depleted 96,853.2 39.3 5.7 <0.1 3,668.7 38 2,060.5 21

Total/ave 246,683.1 217.7 0.9 25,742.1 104 331,998.1 1,346

Because there were no TRU workplace or bioassay contaminant analyses before about 1986, another 

approach for internal dose reconstruction must be taken on the following bases:

• Only uranium urinalysis was performed routinely from the 1950s to 1986. There was no , 

direct measurement of TRU contaminants during this period, and the dose determined from 

only uranium urine results obtained during this period may underestimate a worker’s internal 

dose.

• Direct in vivo lung monitoring began in 1968 with the MIVRML and continued in 1989 with the 

Fernald IVEC counting facility until 2001. No RU contaminant constituents were reported in 

the analyses before 1989. Although there would have been some ability to detect some TRU 

materials in the range of 0.1 to 10’s of nanocuries depending on the radionuclide, the in vivo

counts were not performed with a consistency or frequency to be of significant value in TRU 

dose reconstruction. 

• Before DOE Order 5480.11 (effective in 1989), bioassay data at Fernald was not routinely 

used to estimate intakes and internal organ doses; rather lung counting was used to estimate 

percent MPLB and infer annual dose by multiplying by 15 rem yr-1 per MPLB. Measurements 

of uranium in urine were compared to limits based upon preventing toxic effects from heavy 

metals. Workers were removed from the higher exposure jobs when they approached the 

uranium urinalysis limit or placed on restrictions based on the % MPLB results from in vivo 

counting analyses, or both. However, the in vivo counting frequency seldom exceeded once 

per year – even for high exposure potential work groups.
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• The air monitoring program was used to establish work controls, such as respiratory protection 

requirements for workers, and was not routinely used to establish internal intake or exposure 

estimates.

• Annual exposure reports listed uranium lung burdens in percent of MPLB only and did not 

address the systemic radiological burdens.

• Because the few studies performed indicated historical variability of TRU contaminant levels 

among locations and processing campaigns, it is not possible to develop an estimate of TRU 

exposure to individuals based upon their work place history, thus a reasonable default 

maximum is recommended for all of the processes.

• According to the Fernald Closure Project (FCP) Internal Dosimetry TBD (Tomes 2001) most of 

the POOS uranium on the site is less than 80 ppb plutonium. However, an examination of the 

detailed statistical data sheets for the DOE (2000) RU report (summarized in Tables 5-9 and 

5-10) indicates that maximum values ranged up to 100 ppb in some subgroups as listed in 

Table 5-9. One subgroup had a maximum value of 412 ppb. This maximum level was 

associated with the tower ash from PGDP, which was handled with knowledge and additional 

protective precautions. 

• Though the long-term average RU contaminant levels in the plant are below 10 ppb U for 

plutonium, there are places and materials in the plant that could have provided RU 

contaminants above these average values. Based upon the preceding facts and conditions, 

the most technically defensible and claimant favorable approach to assure that missed internal 

dose from unmonitored and/or undetected TRU activities (that were present throughout all of 

the Fernald plants from 1961) were accounted for is to determine the uranium intake and add 

a ratio of TRU to that intake. Therefore, it would be reasonable and claimant-favorable to add 

100 ppb U for 239Pu, 3500 ppb U for 237Np, and 9,000 ppb U for 99Tc to the calculated uranium 

gram value intakes calculated by the dose reconstruction staff from uranium after 1961. Table 

5-11 lists conversion factors for this approach:

Table 5-11. PPB conversion factors. 

RU 

Contaminant

ppb U × (value) 

= pCi gm-1 ppb U × (value) = Bq gm-1 ppb U × (value) = dpm gm-1

Pu-239 62.89 2.327 139.6

Np-237 0.714 0.0264 1.59

Tc-99 17.15 0.6346 38.07

• The chemical forms of the RU contaminants are not known, although it is apparent from the 

chemical processes to which the materials were subjected during uranium processing, a 

variety of forms would be expected. Hence the dose reconstructor should assume the most 

claimant favorable solubility type for the target organ.

• In the case of evaluating internal intake from long-term calculated environmental air activity

concentrations, the approach of adding a fraction of TRU contaminants to the uranium effluent 

values as a default is recommended. This is based upon the fact that there is no data for TRU 

effluent other than those calculated from analyses of feedstock contamination levels, a few 

sets of plant sample analyses, and calculations, based upon process knowledge. Although 

there were materials in various plant locations that could have been dispersed to the environs 

at higher levels, by far the bulk of releases to the environs were from stacks and 
Effective Date: 05/28/2004 Revision No. 00 Document No. ORAUT-TKBS-0017-5 Page 18 of 42

representative of the average plant contaminants. For consistency, the claimant-favorable RU 

contaminant levels for plutonium, neptunium, and technetium in Table 5-12 could apply. For 

example, Table 5-12 is a format guide that presents the calculated percentage increase in 

count rate of total uranium count on air samples:

Table 5-12. Percent of total uranium activity added to air monitoring samples.

Uranium 

Enrichment

Uranium 

(pCi μg-1)

Plutonium 100 ppb Neptunium 3500 ppb Technetium 9,000 ppb*

(pCi μg-1) % U act. (pCi μg-1) % U act. (pCi μg-1) % U act.

Depleted 0.402 6.175E-3 1.54 2.45E-3 0.61 0.152 37.8

Natural 0.683 0.90 0.36 22.3

2% enriched 1.616 0.38 0.15 9.4

*Technetium-99 is a beta emitter and at least 5 orders of magnitude less hazardous by activity. If beta counts were taken on 

the air sample, it is possible in some unusual circumstances that the count could be predominantly due to the Tc-99.

With an alpha count, which was the preferred analysis method for counting air samples, and at the 

100 ppb Pu contaminant level, the alpha count on the filter would normally be less than 1% due to 

plutonium or neptunium contamination (i.e., the increase would be indistinguishable in the normal 

uncertainty of the counting statistics). As a default, the calculated air concentrations should be 

increased by the percentages in Table 5-12. For the 2% enrichment default, the total plutonium 

activity would be increased by just 0.4%, the neptunium activity by just 0.15%, and the technetium 

activity by 9.4%.

5.2.3

Much of the thorium production data has been lost, and the plant and bioassay monitoring data 

recovered to date has been sparse. A comprehensive effort to reconstruct the effluent of uranium and 

thorium from the Fernald plants in 1988 discovered that a large number of records and files were

destroyed in the early 1970s during declassification efforts (Dolan and Hill 1988). Reviews of AEC 

records in Oak Ridge and Atlanta failed to uncover additional details. 

Thorium Processes

Thorium processes had been shut down and most of the thorium processing equipment had been 

removed prior to the effluent data reconstruction, which made the reconstruction more difficult. The 

estimates in this section are based on process flow descriptions and available production yields. 

Where production data was not available, estimates were based on product volume and yield 

information. These estimates were researched from the various files of FMPC, Oak Ridge, AEC, and 

FMPC customers. This information was used to develop process flow sheets and locations of 

possible emission sources that were identified from the components of each process.

Part of the reconstruction process involved interviews of long-time current and retired employees 

about thorium production. The interviews included the following questions:

• With which processes are you familiar?

• How long did you work in that area?

• What was your job function?

• What type of emission monitoring/control equipment existed in that process?

• What was the production rate?

• What problems were experienced in this operation?

• Where would you say were the greatest sources of emissions?

• How many and what types of scrubbers existed in this system?

• Did all the dry processes vent through a dust collector?
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Information from these interviews was used to refine processing times, plant locations, processes, 

and working conditions.

Though the processes and products involved with uranium production were not precisely the same as 

those in the processing and production of thorium products, comparison of the volume and mass of 

the thorium materials with those of uranium is instructive and is used in this report as bases of some 

default recommendations. The data reconstruction indicates that thorium processing was limited to 

three plants over short periods in the 38-year production history of FEMP. 

In 1972 the site was appointed as the national repository for thorium. For the purposes of dose 

reconstruction, effective equilibrium is a logical, claimant-favorable assumption because thorium was 

present from the earliest times and was stored at FEMP after the industry need for thorium products 

was past. However, for purposes of perspective Figure 5-1 presents thorium equilibrium growth 

curves.

Figure 5-1. Ingrowth of 232Th decay products. 

Table 5-13 provides a tabular presentation of the thorium production estimates in MT by year, 

compared to uranium production in the same plants. The production values in Table 5-13 represent 

only the uranium and thorium production in the specific areas as identified in the studies. A total of 

980,048 MTU was processed site wide, compared to 2,855 MT of thorium (or approximately 0.3% of 

total uranium mass). In the three plants that processed thorium, the largest quantity (71%) was 

processed in the Pilot Plant. However, even in this plant, thorium represented only 25% of the total 

mass of uranium processed. Table 5-14 lists a summary of the thorium plant processes with chemical 

forms and solubility types.
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Table 5-13. Thorium production estimates (MT) by year, with uranium production.

Fiscal 

Year

Pilot Plant Plant 9 Plant 8

Thorium (gel, 

oxalate, extraction)

Uranium

UF4

Thorium

(metal, briquette)

Uranium

(metal, briquette, Zirnlo)

Thorium

(oxalate, hydroxide)

1953 0 0 0 0 0

1954 0 0 212 0 0

1955 0 0 244 0 0

1956 0 0 0 0 0

1957 0 0 0 0 0

1958 0 0 0 1,735 0

1959 0 0 0 2,965 0

1960 0 0 0 3,289 0

1961 0 0 0 5,603 0

1962 0 1,342 0 6,310 0

1963 0 1,197 0 8,818 0

1964 98 1,468 0 12,698 0

1965 98 658 0 12,850 0

1966 251 478 0 3,165 59

1967 158 13 0 3,095 0

1968 243 0 0 824 0

1969 287 0 0 874 149

1970 249 0 0 603 100

1971 74 0 0 443 61

1972 122 0 0 599 0

1973 49 0 0 452 0

1974 50 0 0 1,031 0

1975 2 0 0 1,189 0

1976 0 0 0 304 0

1977 117 0 0 401 0

1978 117 0 0 558 0

1979 117 0 0 671 0

1980 0 0 0 460 0

1981 0 0 0 827 0

1982 0 0 0 1,050 0

1983 0 0 0 1,523 0

1984 0 0 0 1,766 0

1985 0 622 0 1,287 0

1986 0 462 0 1,984 0

1987 0 542 0 2,787 0

1988 0 1,642 0 4,443 0

Total 2,030 8,424 456 84,604 369

Table 5-14. Thorium plant processes with chemical forms and 

solubility types.

Plant and Process

Production

Years

Chemical 

Form

Plant 9–metal briquetting 1954-55 Th, ThO2

Pilot Plant–extraction 1964-80 ThNO3, ThO2

Gel 1977-79 Th(OH)4

TNT crystals 1966 ThNO3

Metal 1969-71 Th, ThO2

Oxalate 1971-76 ThNO3

Plant 8–hydroxide 1966 Th(OH)4

Oxalate 1969-71 ThNO3
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The ICRP has assigned oxides and hydroxides of thorium to inhalation type S. All other compounds 

of thorium are type M. The claimant-favorable assumption would be either type M or S, based on the 

organ of interest, because all of the compounds in Table 5-14 were handled and could have resulted 

in intakes. The default isotopes would be 228Th and 232Th in equilibrium, since the degree of 

equilibrium is impossible to estimate due to the variable times of separation of the isotopes from the 

feed stock (see Figure 5-1). Table 5-15 lists the estimated thorium emissions from FEMP plant 

facilities in comparison with uranium emissions.

 Table 5-15. Estimated thorium emissions (kg) in comparison with uranium.

Year

Total Uranium 

Emissions

Total Thorium 

Emissions

1951 123.0 0

1952 499.0 0

1953 2084.6 0

1954 15,345.0 1,028

1955 33,751.4 1,176

1956 15,376.0 0

1957 10,832.0 0

1958 8,844.5 0

1959 8,979.3 0

1960 8,941.5 0

1961 7,202.8 0

1962 6,682.1 0

1963 6,078.0 0

1964 5,295.9 344

1965 7,266.9 344

1966 2,340.9 462

1967 3,572.2 344

1968 5,787.2 344

1969 4,605.7 1,434

1970 1,812.0 1,198

Year

Total Uranium 

Emissions

Total Thorium 

Emissions

1971 1,062.1 492

1972 1,574.4 141

1973 1,718.6 50

1974 2,677.8 100

1975 3,111.5 3

1976 3,515.0 0

1977 952.3 408

1978 230.6 408

1979 162.6 408

1980 275.2 0

1981 627.0 0

1982 342.5 0

1983 325.0 0

1984 962.9 0

1985 218.7 0

1986 43.8 0

1987 246.9 0

1988 97.5 0

Total 173,574.8 8,684

Realizing that the thorium data are not measurements, but are the best values that the TBD technical 

staff (Dolan 1988) could reconstruct on the basis of available records, recollections of professional 

engineers, and best estimates on the basis of process knowledge, this information represents the best 

available. From these data estimates it is clear that thorium represents less than about 5% of the total 

emissions from the plant processes and that processing occurred during fewer years. In addition, 

these emissions give some qualitative indication of the estimated availability of the materials to the 

airborne pathway. Because the contamination controls for thorium processing (ventilation, etc.) were 

the same or equivalent to those for the uranium processes, certain assumptions in relation to 

contaminants in the work place apply to both processes.

The same air sampling procedures were followed for thorium processing as for uranium processing. 

Some records have been recovered that indicate that basic air activity levels were recorded in 

fractional MAC (70 dpm m-3 prior to 1970 and100 dpm m-3 thereafter) for thorium processing. The 

thorium air sampling results are similar to the uranium air sample results. However, from the limited 

data examined, the measured MAC levels during the thorium campaigns do not appear to approach 

the higher MAC levels measured during the processing of uranium. The practices of wearing 

respiratory protection preventatively for operations known to produce dusty conditions were 

administered for thorium operations as they were for the uranium operations. 
Effective Date: 05/28/2004 Revision No. 00 Document No. ORAUT-TKBS-0017-5 Page 22 of 42

A fundamental difficulty of dose reconstruction for thorium processing is that either 1) in vitro 

bioassays for thorium were not performed or 2) data is not available until after 1986. An additional 

consideration is that air sampling data was not used to calculate intake and dose until after 1986. Air 

monitoring was used only to control exposures to levels below the MAC. A number of internal 

memoranda identify those areas with concentrations at or above the MAC of 100 dpm m-3 (4.5 ×10-11

μCi cm-3

) as areas requiring respiratory protection. However, recorded examples of exposure to 

multiple MAC levels without respirators indicate these violations of policy were not uncommon. In 

addition, the urine sampling was performed for uranium only. The only discovered record of thorium 

exposure has been in vivo lung count data sheets in a few claimant records and a single claimant 

record which indicates thorium urine results, counted for beta and at essentially no detectable results, 

from before 1986. Thorium processing was completed in 1979, with exposure from that time being 

limited to repackaging and shipping operations. 

After 1986 thorium air sampling was used to estimate internal exposure using continuous lapel air 

samples as breathing zone (BZ) evaluations. From that time until the present air monitoring is used to 

conservatively estimate internal intake even when the worker wore respiratory protection.

The data from the report (Dolan 1988) indicates that just the Pilot Plant, Plant 8, and Plant 9 

processed thorium. A single air sampling data sheet was found that indicated a thorium equipment 

repair operation in Plant 4 during which there were air activity concentrations above MAC. Therefore, 

the three plants mentioned should be considered the primary processing sites, although there is some 

evidence that a few isolated thorium operations occurred in other locations. Based on evaluation of 

the available information, dose reconstructors should assume thorium exposure for any employee 

whose records establish work, and therefore exposure potential, primarily in the Pilot Plant from 1964 

to 1979, in Plant 9 in 1954 or 1955, or in Plant 8 in 1966, 1969, 1970, or 1971. 

This reconstruction of the operational history of the large volume thorium processing is the best 

available at the time of this report. Th230 is also found in the plant as a uranium decay product. Most 

of the thorium from this source is removed from the ores during solvent extraction and is found in 

quantity in the raffinates. See Table 5-16 for content of the Silos 1 & 2, where Th230 make up 7 to 10% 

of the total activity. Some thorium as a contaminant follows the uranium streams through the plant in 

trace quantities but will constitute <1% of the thorium default assumptions below.

In vivo counting was performed on the workers in the more likely exposed groups at least once each 

year. There is some evidence of urine analyses for thorium in claimant files as early as 1955, but to 

date no information has been found regarding how to interpret it. Although urinalysis can offer some 

information regarding thorium intake, it is not the preferred bioassay technique, since the material is 

is predominantly insoluble. Fecal sampling and in vivo analyses are the preferred default. This is a 

difficult default to derive with any degree of technical basis because: 

1- There was primarily gross alpha and some gross beta air monitoring during thorium operations 

for the purpose of controlling worker exposures to below MAC levels. A few in vitro analyses 

for thorium were discovered primarily in claimant file records; only a few in vivo analyses were 

found. The thorium results are questionable because of the lack of information for readily 

interpreting them (e.g., there is no information regarding the in vitro separation method or 

counting procedure/equipment, nor is there information regarding the assumptions made to 

derive the in vivo results). 

2- It is known that respiratory protection (both preventative and following MAC-level air sample 

results) was provided and would have resulted in at least a factor of 10 protection when used 

properly. However it is also known that workers were exposed to >MAC levels without 
Effective Date: 05/28/2004 Revision No. 00 Document No. ORAUT-TKBS-0017-5 Page 23 of 42

respiratory protection. Considering this information, standard respiratory protection factors 

cannot be assumed.

3- Limited operation times and smaller volumes and mass (which also would presuppose a more 

effective ventilation confinement) reduced the exposure potential, all of which would result in 

an assumption for limited periods of higher-level contamination.

4- The MAC of 100 dpm m-3 (4.5 × 10-11 μCi cm-3

) is 20 to 100 times larger than the current 

derived air concentrations for 232Th. 

Based upon the above information and assumptions, the recommended claimant-favorable default 

exposure approach to assign thorium intakes is to assume:

• An intake for an exposure period of 100 hours per year at an assumed exposure of 10 MAC is 

judged adequate to account for the higher levels of exposure indicated by air sampling, since 

few samples above 10 MAC were reported and these primarily represented short term 

maximized sampling (based upon descriptions on the sample sheets). Also typical and more 

extensive uranium air sample data demonstrate that 10 MAC is a reasonable assumption of 

the higher level of exposure.

• No respiratory protection factor, although not wearing respirators when air concentrations were 

above MAC represented procedural violations. This violation was known to have occurred 

and was not unusual.

• An intake for an exposure period of 500 hours per year at an average air activity of 0.1 MAC 

during normal operations

Using these assumptions, the claimant-favorable assumption would be: 

500 hr × 0.1 MAC + 100 hr x 10 MAC = 1050 MAC-hr exposure (5-1)

This exposure results in an intake of about 60 nCi per year. Therefore, in the absence of monitoring 

data a claimant-favorable default intake is 30 nCi/y (82 pCi/d) each of Th-232 and Th-228 (the alpha 

emitting isotopes detected on the air samples) plus a 60 nCi/y (164 pCi/d) intake of Ra-228. The Ac228 beta emitter adds about 0.1% to the effective dose and therefore can be ignored in the dose 

calculations. Although an equal intake of Ra-224 might have occurred, because of its alpha 

emissions, it is adequately accounted for in the thorium intake assumption, which is based on 

detection of all alpha emitters on the air samples. This default intake rate applies to the most exposed 

craft personnel (e.g., chemical operators, process maintenance personnel, safety personnel, and first 

line supervisors/foremen) at the locations and during the periods noted below. For workers whose 

location cannot be determined the claimant favorable assumption is that they were exposed to 

thorium and daughter products when they were employed during the listed periods in Table 5-16. 

Exposures to casual workers who worked in the immediate vicinity of the plants should be evaluated 

in the Environmental Occupational Dose section. 

Thorium chest counts may indicate that lower exposures occurred. When available, the chest count 

data should be used to constrain an employee’s intake. In other words if there are chest count results 

for an employee, the smaller of the default thorium intake or the chest count determined intake 

should be assigned for full dose reconstructions.
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Table 5-16 Default thorium exposure.

Period Area Worker categories Intakes 

(pCi/d)

1954-1955 Pilot Plant Chemical operators, 

process maintenance 

personnel, etc.

Th-232 82 

Th-228 82 

1964-1979 Plant 8 Ra-228 164

1966, 1969-1971 Plant 9

5.2.4 K-65 Silo Processes

Information on the K-65 silo processes is derived from air sample and external radiation dose data

sheets from the 1953 period, Krach 1998, Tomes 2001 and RAC 1995. The raffinates from the Plant 

2/3 processing or uranium ores were relatively high level wastes (in the μCi/gm range) and required 

special storage. The waste especially from processing African pitchblende was particularly high in 

226Ra (and daughter products) content. Large concrete tanks, called K-65 silos, were constructed 

from August 1951 to July 1952 for the “interim” storage of the pitchblende waste, since it technically 

belonged to the African Metals Corporation. The chemical extraction plant 2/3 was complete and 

started operation in December 1953 and disposed of the raffinates directly to the storage silos 1 and 

2. However, Mallinckrodt Chemical Works (MCW) facility in St. Louis had processed African 

pitchblende ores prior to completion of the Fernald extraction process facilities and the K-65 storage 

silos, and did not have sufficient storage space. Therefore, MCW shipped their pitchblende raffinate 

wastes in lined 55-gallon drums to Fernald for interim storage. Shipments from MCW began in 1951 

and by the end of July 1952, when the silos were complete, approximately 13,000 55-gallon drums of 

K-65 (approximately half the capacity of one silo) were stored on the ore storage pad around Plant 1, 

the Sampling Plant. The drummed waste was transferred to silos 1 and 2 from July 1952 through 

September 1958. The total of the radium-containing residues in Silos 1 & 2, resulting from the 

processing of uranium ores is approximately 10,000 MT.

Silo 3 was used for the raffinate storage from “cold metal oxide” extraction separations and contains 

approximately 138,000 cubic feet of raffinate. The feedstock for these processes was uranium 

concentrates from a variety of uranium mills in the United States and abroad. Though this material 

contains radioactive material, most of the 226Ra had been removed at the mills and the remaining 

levels were not a significant source of exposure in comparison with Silos 1 and 2. The materials in 

Silo 3 were calcined prior to storage and are a fully-oxidized, fine powder in contrast to the K-65 

material in Silos 1 & 2, which are approximately 30% moisture. Silo 4 was never used as a waste 

storage silo and contains only a small amount of low level contaminated water.

The silo waste became the property of the United States in 1983 and has become a storage problem. 

The concrete silos cracked, leaked and were the object of several upgrade efforts to reduce both the 

liquid leaks and the release of radon and its daughter products. A dirt berm was added around the 

silos (primarily as silo wall support but provides shielding also), as were roof sealants and a semipermeable cap on the waste itself, all of which reduced the routine releases of radon plus daughters 

from the silos.

The operation of handling the large number of drums of K-65 waste materials and dumping them into 

the silos in the period of July 1953 to September 1958 was an operation in which external and internal
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exposure to the radium and other associated contaminants occurred. In addition there were several 

operations to improve the confinement of the silos for the radon gases and associated daughter 

products, although the waste materials have not been directly disturbed since they were put into 

storage. Air sample data sheets from the 1953 time period have been found and provide insight into 

the operation and exposure levels. The only bioassay information related to internal exposures to 

radium or the associated contaminants are a few radion breath samples in the 1953 time period. In

vivo counts were periodically performed by the Y-12 mobile counting laboratory; however these in vivo 

counts did not begin until 1968, a long period after completion of the handling and dumping of the 

13,000 drums of waste.

Table 5–16 is a summary of the radiochemical analyses of Silos 1 and 2 core samples taken in 1993, 

and is provided for internal dose reconstruction purposes in visualizing the potential particulate source 

term (Tomes 2001). It should be noted that in Silo 1 approximately 92% of the total activity is 

composed of 226Ra and the longer lived daughters 210Pb and 210Po. In Silo 2 the same isotopes 

comprise approximately 88% of the total activity.

 Table 5–16. Isotopic composition of K-65 Silos 1 and 2.

Isotope Silo 1 Silo 2

Activity (nCi/gm) Activity Fraction Activity (nCi/gm) Activity Fraction

Uranium – Total 1.68 1.61 E-3 2.37 3.04 E-3

Ac-227 (β) 7.67 7.36 E-3 6.64 8.50 E-3

Pa-231 (α) 4.04 5.17 -3

Pb-210 (β) 202 1.94 E-1 190 2.43 E-1

Po-210 (α) 281 2.70 E-1 231 2.96 E-1

Ra-226 (α) 477 4.58 E-1 263 3.36 E-1

Th-228 (α) 2.28 2.19 E-3 7.36 9.42 E-3

Th-230 (α) 68.9 6.62 E-2 76.2 9.75 E-2

Th-232 (α) 1.11 1.07 E-3 0.99 1.26 E-3

The information that is available upon which to base estimates of intake of radium and its daughter 

products to an unknown number of workers consists of a few air sample data sheets in late 1952 to 

early 1953 with alpha analyses (“O” drive Data Capture\601-881\010000784\000784.pdf) and 

personnel dosimetry records of external dose to workers specifically identified as working on the 

MCW raffinate dumping operation (“O” drive Data Capture\601-881\010000867\000867.pdf). The 

basic facts from this data are as follows:

• Movement of the 13,000 55-gal drums of MCW pitchblende raffinates to Silos 1 and 2 during the 

six-year period of July 1952 to Sept 1958 (74 months) could have been a continuing effort 

averaging 176 drums per month (approximately 8 per day). However, one of the air sample data 

sheets in 1953 indicated an 80 drum per day rate (assumed to be a 3 shift-24 hour day), or 3.3 

drums per hour, this seems reasonable, based upon other notes relating to the need for shoveling 

the contents, “knocking” on the barrels, etc. At a rate of 80 drums per day the movement could 

have been accomplished in 163 days. 

• The description of this operation was inferred from brief descriptions contained on the air sample 

data sheets from 1952 and 1953. Both elevated air sample activity and the existence of elevated 

radiation levels were clear from the sampling location descriptions, e.g. “behind shielding”, etc. 

and analytical results. The operational descriptions came primarily from descriptions on air 

sample sheets, i.e. removing/replacing lids and liners, beating the rim to dislodge contents,
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digging into the drum contents with a shovel, standing on the conveyer, sweeping and washing, 

etc.

• From the information in the series of air sample data sheet analyses in 1952 and 1953, it is clear 

that an air monitoring program was in place. The results indicate both GA (5 to 30 minute 

duration) and BZ (1 to 5 minutes) sampling with results <0.2 MAC, many >1 MAC, and a few in the 

50-60 MAC range - and one at about 180 MAC. On a couple of data sheets there was a note that 

no respirators were worn. It is presumed that the note was made to record an unusual event, 

since standing air monitoring policy was for the purpose of evaluating the need for respiratory 

protection.

• There were air samples taken for the determination of radon levels during the barrel dumping 

operation (as inferred from a single “Air Radon Sample” data sheet on 10-29-53), the results of 

which were 1 and 2.3 x 10-10 Ci/L Rn.-or 1 to 2.3 WL.

• No records were found of any bioassay results for radium or daughter products during this time 

period – with the exception of a series of radon breath samples. No In vivo counts were made on 

personnel during this period.

• The information on the data sheets indicates that in spite of the fact that the contents of the drums 

were wet, the operations resulted in significant airborne contamination.

• A group of external dose data sheets were found during the information search that recorded 

radiation doses for 22 workers specifically identified as working on the K-65 dumping operation. 

Conclusions derived from these data sheets can be summarized as follows:

 Dosimeters were processed weekly during 10 one-week time periods from 9/8/52 

through 11/31/52.

 During this period the 22 assigned individuals worked for periods ranging from 1 to 6 

weeks (average 4.7 weeks) each and received an average of 275 mrem/week.

 Of those 13 workers who received the highest doses (ranging from158 to 500 mrem 

average/week) 3 worked 5 weeks and 10 worked 6 weeks with an overall average of 

312 mrem per week. 

 The highest individual dose in any week was 1200 mrem. Though radiation dose rate 

survey records are not available, it is apparent that working levels would be in the 40 to 

60 mrem/hr., i.e. 1200 mrem/20 to 30 hours per week in the highest radiation fields = 

40 to 60 mrem/hr. Actual radiation fields could have been higher.

 Work schedules on 1/16/52 recorded 3 shifts of 4 workers per shift, indicating that 

during the most intense transfer period work proceeded 24 hours per day.

 Assuming that those workers receiving the highest weekly doses were those workers 

who would also be the most likely to be in the highest particulate air activity, the 

external dose limits would restrict exposure times to approximately 3 months per year, 

i.e. (4000 mrem/312 mrem wk-1

) (4.3 wks/month)-1 = 3 months. This assumes that 

doses above 4 rem would not be administratively planned to avoid exceeding the 5 

rem/year limit. In any event external dose limits would restrict personnel exposure to 

the recorded maximum air activity to approximately 3 months per year.
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 From the information derived in the external dose data sheets and the air monitoring 

sample sheets, it appears that the transfer could have been limited to a period of 10 

weeks per year with no individual working more than a period of 6 weeks in the year

For internal dose reconstruction of workers in the Silos 1 & 2 areas in the absence of specific in vitro 

and/or in vivo data, a maximization approach, based upon air sample results, is proposed: 

in 

order to control external dose within the regulatory limits. This could have resulted in 

80 drums per day x 10 weeks x 5 days per week = approximately 4000 barrels per 

year, which would have been double the rate necessary to transfer the 13000 barrels in 

the 74 month time actually utilized by the Site.

• assume a worker was assigned to all of the dumping operations during 6 weeks of each of the 

years in the 6 year period; 

• maximum dumping rate, and hence maximum exposure rate, 80 drums/day; and 

• 100 MAC (4.5 E-9 uCi/ml alpha analysis) exposure levels with no respiratory protection at 

9.6 E+06 cm3 per day breathing rate. The calculation is: 

4.5 E-9 μCi/cc x 9.6E+06 cc/day x 30 days = 1.3 μCi intake per year for the period of July 1953 

to September 1958. The contaminants are assumed to be composed of Table 5-16 Isotopes 

in their indicated activity fractions. For purposes of dose reconstruction, it should be noted 

that the MAC air concentrations were determined by gross alpha counts, requiring the addition 

of the appropriate percentage of the beta emitting isotopes.

From a single radon sample data sheet on which the analyses of two samples were recorded on 

10/29/53, the higher of the two samples indicated a result of 230 pCi/L radon gas, which verifies the 

logical assumption that radon gas was released as the drum lids were removed. In addition to the 

default particulate intake (determined as previously stated), a conservative/bounding analysis of 

possible radon plus daughter product exposures can be derived:

• Assume 230 pCi/L (2.3 WL) with 100% daughter product equilibrium for 1304 hrs. (163 day x 8 

hr/day)/ 74 months of the dumping operations = 17.6 average hours/month exposure. Then 2.3 

WL 17.6/167(the fraction of a full working month) x 12 months = 2.9 WLM exposure per year.

This assumed bounding exposure to radon plus daughter products would be in addition to any 

assigned exposure that may be derived from the Section 4, Occupational Environmental Dose 

calculations.

It is evident that these estimates are based upon assumptions that are cumulatively conservative, 

claimant favorable, and establish an upper bound of intake for workers involved in the transfer 

operation of the 13,000 barrels of the stored MCW raffinates to the K-65 silos. Calculations of internal 

intakes resulting from exposures to the raffinate dusts generated during dumping operations should 

be used only for claimants for whom a work history on this specific project can be established. An 

examination of external penetrating radiation dose for workers who were known to have worked with 

and handled these drums of raffinate wastes show significant (several100 mrem per week) 

penetrating dose accumulation. Therefore a criteria to determine and/or verify that a worker had 

indeed been exposed to internal intake from raffinate dusts would be a record of penetrating external 

dose, i.e. no detectable dose would clearly indicate little direct contact or work with the barrels of 

waste.
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As previously stated, the contents of the silos have not been disturbed during storage to any large 

degree. However, it has been calculated that during the 1953 to 1978 period 5,000 to 6,000 Ci/year 

of 222Rn were released from the silos (RAC 1995). Considering the expected large differences in 

release rates due to barometric pressure changes, the release rates would average up to 15 to 20 

Ci/day after addition to the Silos were complete. Again, no monitoring or operational descriptions of 

the maintenance and/or silo upgrades are available, which would involve extensive work near the 

silos at the time of this writing. If it is known that a specific claimant was a part of a crew which spent 

significant time near the silos during the upgrade operations, intake of Rn222 plus daughters will best 

be calculated as defined in section 4, Occupational Environmental Dose from the data in Table 4-9a. 

Radon and thoron plus daughter product sampling was conducted inside each plant building

5.3 RADIOLOGICAL CONTROLS PROGRAM

at the 

Fernald Site primarily in the late 1980s (Weaver 1987). Documentation at that time indicated levels in 

terms of percentage of 0.4 WL (ICRP 32 limit) generally in the 0.5 to 7% range, with a few locations 

that ranged in the 10 to 30% levels. This would indicate minimal occupational internal exposure 

above normal expected radon/thoron backgrounds from the Silos source. However, a default of 5% of 

0.4 WL could be added to plant exposures. This equals an exposure of 0.24 WLM per year.

The initial health and safety organization (the Industrial Hygiene and Radiation Division) at FEMP was 

organized and directed by an occupational medical physician and staffed primarily with industrial 

hygienists. The Fernald staff worked with the Oak Ridge National Laboratories Y-12 staff, because of 

similar radiation safety issues associated with their uranium processing, although Y-12 typically had a 

more highly enriched uranium source term. FEMP also used Y-12’s Mobile In Vivo Radiation 

Monitoring Facility (MIVRML).for over 20 years. 

5.3.1

Radioactive contamination was routinely measured in occupied areas of the plant, and there were 

significant radioactive material releases to the environment. It was necessary for every facility at the 

FEMP to address issues involving airborne contamination, even though engineered confinement 

barriers were used in conjunction with process and work area ventilation. Elevated airborne 

radioactivity resulted from processing thousands of metric tons of dispersible radioactive materials 

with a variety of chemical and physical characteristics. Throughout the process history of the site 

there were high potentials for intakes of uranium, thorium, and their radioactive progeny. 

Radiological Controls Program

A radiological controls program was in place from the beginning of FEMP operations. The internal 

dose control program consisted of:

• An air sampling program in all processing areas to evaluate internal exposure potential via 

inhalation

• Urine samples submitted after at least a two-day work break to allow elimination of uranium 

cleared rapidly via the GI tract (this material causes relatively little dose) 

• In vivo analysis once a month for high exposure-potential workers on a frequent urinalysis 

program and once a year for workers with a low potential for internal intake.

Other elements of the protection program included routine monitoring of the workplace and personnel 

for radiation and contamination, personnel protection in the form of protective clothing and respiratory 

protection in all of the operational areas as needed, and restricting workers from workplaces with 
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elevated airborne radioactivity concentrations when the level of uranium in the urine or in vivo

counting results exceeded specified plant action limits.

5.3.2

The large quantity of data in the archives shows that FEMP maintained an aggressive air monitoring 

program from the beginning of operations. Both high and [primarily] low-volume general area (GA) 

and breathing zone (BZ) air samples were collected (most for 3 to 30 min) and counted for alpha 

contamination in the 1950s. A few sample records and claim file records indicate that some beta 

counts were performed. In the 1960s the samples were counted for both alpha and beta activity. The 

results were compared to the National Lead of Ohio MAC guidelines or the National Lead 

Concentration Guide (NCG) of 100 dpm m-3 (70 dpm m-3 was used as the MAC/NCG until the 1970s). 

The 100 dpm m-3 equates to about 4.5 × 10-11 μCi cm-3

.

Air Monitoring Program

Routine air samples were taken in every plant and operational area. This program was the primary 

means of controlling intakes. Workers were directed to use respiratory protection in the form of dust 

masks or supplied air respirators depending on the anticipated or measured airborne radioactivity 

concentration. From a historical viewpoint, extensive, long-term air activity summary sheets that 

covered 15 or more years were periodically prepared that indicated routine detectable air activity in all 

working areas of each plant. These summaries detailed annual average exposures to workers 

without respiratory protection and average air activities associated with job assignments that required 

respirators. The air activity ranged from a fraction of the MAC levels to hundreds of times those 

levels.

Since 1989 BZ samples were (and continue to be) converted to DAC-hours in the case of thorium 

exposures. On occasion, these air sample analyses are converted to intake, dose, or both. The 

calculated doses observed in the records were low, i.e., 10s of mrem CEDE. The air sample results 

are used with the thorium isotope (230 or 232) and a conversion to % DAC (2E-02 Bq m-3 for Th232 

and 9E-02 Bq m-3 for Th230), and then converted directly to dose in CEDE. 

For reference, Table 5-17 lists the abbreviations and codes on the sample and dose calculation 

datasheets. Table 5-18 cross-references FEMP air-monitoring methods with the applicable MDLs.

Based on the preceding information on FEMP air monitoring, the following default explanations and 

assumptions may be applied for the purpose of dose reconstruction.

Table 5-17. Air monitoring analysis sheet codes.

Code Description

MAC NLO maximum allowable concentration – 100 dpm m-3 and 70 

dpm m-3 prior to the 1970 time period

NCG NLO concentration guide – used interchangeably with MAC

R Sample collection rate in m3 min-1

T Sample time in minutes

Q Sample volume in m3 Q = R × T

GA Sample collected in a general area

BZ Sample collected as close to the breathing zone as practicable

Analyzed for – Alpha Gross alpha count

Analyzed for – Beta Gross beta count

Analyzed for – Ra Alpha count on radium sample separation

Analyzed for – Th (#33) Alpha count on thorium sample separation

Analyzed for – all others Chemical analysis for non-radiological samples
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Table 5-18. Cross-reference of FEMP air-monitoring methods with the applicable MDLs.

Type of 

Analysis Method Period Frequency MDL*

Air sampling Breathing zone 

sampling

1952–1993 Variable and frequent, depending 

upon work

0.2–0.02 MAC depending on 

sample volume

Air sampling Breathing zone 

lapel samplers

1993–present Job specific: when Th-230 or Th232 exposure potential exists. 

BZ Sampling for Uranium is also 

performed for work controls

1–2 DAC-hrs with a Minimum 

Detectable Intake** of 0.001 -

0.002 corrected DAC-hrs. for 

thorium

* From 1952 to 1993 BZ samples were taken 5 to 30 min at 20 L min-1 ( 0.1 to 0.6 m3,of air) while today the lapel sampler is worn 

continuously at about 4 L min-1 for 8 hours/day (approximately 2 m3 of air collected) and then counted for longer times with more sensitive 

laboratory counters. ** Detectable Intake is lower due to the application of a respiratory protection factor.

From 1953 to 1986, the air monitoring program was conducted as a primary control element. 

However, the measured air concentrations levels from the routine sampling program typically were not 

used to establish worker intakes, and workers were required to submit routine urine samples for 

uranium analysis only for the purpose of verification of site air sample-based controls. The MDL for 

routine air sampling was in the 0.2 to 0.02 MAC range, and when multiplied by an 8-hr workday, MDL 

levels in MAC-hr of approximately 1.5 oer day is derived (based on informal calculations conducted 

from data taken from recorded sample sheet volume and radiation counting data). The most typical 

sample volume was a 3-min sample (0.06 m3

) in dusty areas, which equates to the 1.6 MAC-hr MDL, 

and up to a 30-min sample (0.6 m3

) with an MDL of 0.2 MAC-hr. BZ samples were generally of the 

same flow rate but shorter duration. Long term averages at 0.1 MAC levels were reported, but should

be interpreted as numerical averages only.

Current (1993 to present) air monitoring is performed with the intent of calculating thorium intake and 

dose if necessary for low-dose cases in addition to maintaining site control of personnel exposures. 

The air sample data is used to calculate a specific concentration of a specific nuclide in terms of 

microcuries per cubic centimeter and conversion to a specific DAC percentage. This value is directly 

converted to intake and organ dose, if the results are positive.

5.3.3

A urinalysis program was administered at FEMP starting in 1953 or possibly before. The NLO 

industrial physician and industrial hygienists performed and documented a number of studies to 

establish the uptake or MPBB for workers exposed to uranium. The initial study was based solely on 

heavy-metal toxicological limits for kidney damage. Table 5-19 lists the derived limits, which 

remained essentially unchanged throughout the history of the site.

Bioassay Program

Table 5-19. Routine uranium in vitro bioassay capabilities at the FEMP. 

Type of 

analysis Method Period Frequency MDL

Urine - uranium Fluorophotometry 1952–1993 Weekly to annual – job 

specific

14 μg L-1*

Chemchek KPA 1993 to 9/2002 Bimonthly 0.17 μg L-1 (total U)

ICP-MS Sept. 2002 to 

Present

As requested 0.15 μg L-1 (total U)

Fecal** Fluorophotometry Various No routine schedule Unknown – assume 

environmental levels of 

2 μg per sample 

Contract lab Occasional No routine schedule 0.1 pCi per sample

* Y-12 listed a sensitivity of 1.6 μg L-1 in 1973 using the fluorometric process for 0.7% U-235. Fernald frequently listed less than 0.003 mg 

L-1 in the bioassay data reports. Several blank samples on intercomparison studies also list results as 0.003 mg L-1

. A value of 0.008 mg L-1

has also been quoted in the records as the MDL. However, a formal response on January 21, 1993, (Blalock 1993) to a deficiency in the 
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ability to detect 100 mrem CEDE with the existing 0.014 mg L-1 MDA is accepted as the most reliable representation for historical MDAs for 

this analytical procedure.

** Fecal sampling was performed as a part of special studies or specific incident investigations on occasion. Records of this activity are 

generally not present in the dosimetry records for individuals. 

Nearly all employees provided urine samples for uranium analysis at the time of their annual 

physicals. Workers with work assignments such that exposure could be expected on a routine basis 

were sampled weekly, monthly, or [at least] bimonthly.

5.3.4

The value of fecal samples was recognized even in the early years and has been well understood 

since 1986. As an example, fecal sampling for uranium was performed on several workers in 1968 as 

a part of a DOE Health and Safety Laboratory – National Lead of Ohio (HASL-NLO) Study (as 

recorded on analytical data sheets of October 4, 1968). However, fecal sampling has never been a 

part of the routine bioassay program at FEMP, except in special conditions. 

Environmental Levels and Fecal Sampling Program

As documented in the most recent FCP Technical Basis for Internal Dosimetry at the Fernald Closure 

Project, SD 2008 (Tomes 2001), UNSCEAR and ICRP publications estimate natural intake of uranium 

in the general U.S. population in the range of 0.7 to 2 μg a day with an expected excretion of 0.05 to 

0.5 μg a day in urine and 1.4 to 1.8 ug per day in the feces: 

“Current bioassay methods are sufficiently sensitive to detect dietary intakes of 

uranium in urine. Therefore, an environmental Decision Level (DL) was determined 

and used in place of the MDA for establishing bioassay sampling frequency. The DL 

was determined from the measured results of samples analyzed from newly hired 

employees (a control group) who had no known occupational exposure to uranium. 

The DL was set at a 99.99 percent confidence interval. This means a sample whose 

concentration measures above this level has a 10-4 probability of being the result of 

dietary intake. This would be a false positive result for occupational exposure. 

Therefore, a value above the DL is considered indicative of occupational exposure and 

requires additional sampling and follow-up. The DL for uranium was determined to be 

0.8 μg U/l [of urine]” (Tomes 2001).

The same procedure was followed at other DOE sites to determine the occupational indicator. The 

DL value for urine at the Idaho National Engineering and Environmental Laboratory, for example, has 

been established at 0.16 μg L-1 (Rielly 2001).

It has been recognized at Fernald, particularly since 1986, that fecal sampling can provide useful 

information, particularly in cases of exposure to less soluble compounds (types M & S). However, it is 

also recognized that the natural environmental levels vary considerably and for that reason are less 

reliable for routine use. FEMP did not conduct routine fecal sampling; however, since the mid 1990s, 

when fecal samples were collected, they were analyzed under contract with a specified maximum 

MDA of 0.1 pCi per sample. If lower sample MDAs are found in records associated with contracted 

fecal uranium analyses, these lower MDAs should be used. Uranium results that are within a factor of 

10 of the DL should be adjusted for full dose reconstructions.

5.3.5

The urinanalyses procedure was conducted using the fluorometric fusion process, which fused 

uranium from raw urine with sodium fluoride and measured the fluorescence created by ultraviolet 

Analytical Program
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light. This method provided a measure of the total amount of elemental uranium in the sample. 

Comparison studies between the other laboratories provided assurance that the analytical process 

was consistent with industry standards. As indicated in Table 5-19, the MDA (MDL) was not formally 

established in the early periods (as evidenced by a records search). In addition, the large fluctuation 

in uranium in the diet of nonoccupationally exposed personnel provided implied limits (although 

apparently not well understood in the early periods). The values listed in the bioassay data sheets 

generally range from 0.003 mg L-1 to 0.008 mg L-1 in urine. The effective MDA is now understood to 

have been near 0.014 mg L-1

; therefore, values below this can be accepted as positive but without 

statistical value as indicators of occupational exposure.

In early 1993 other DOE laboratories were changing to kinetic phosphorescent analysis (KPA), with 

detection capabilities as low as 0.02 μg L-1

, which is far below the natural environmental background 

in most people. The MDA of 0.17 μg L-1 was established consistent with other laboratories. The DL 

of 0.8 μg L-1 at FEMP since 1993 represents the best estimate of the non-occupational excretion of 

uranium, although values below the DL are reported in the dose history records. Specifically, when 

Fernald changed to the KPA, 0.8 μg/L became the reporting level until 1997, after which all calculated 

results were reported as derived. Some database printouts for samples after 1997 will include < signs 

for results that are less than 0.17 ug/L, e.g., <0.06 μg U/L. These results should be interpreted as 

less than 0.17 ug/L. This represents a database formatting issue rather than a lowering of detection 

limits.

The method currently being used at Fernald for urinalysis is inductively coupled plasma mass 

spectrometry (ICP-MS) which has an a priori MDA of 0.15 μg L-1

.

For reference, Table 5-20 lists the FEMP uranium urinalysis sample type codes. Table 5-21 lists the 

FEMP sample type coding system and personnel assignment codes.

Table 5-20. Uranium urinalysis sample type codes.

Types of Samples – General

Code Description

00 No Code

10 Pre-Employment Sample – collected during pre-employment physical exam or before beginning 

work on first day. Establishes individual U background

20 Annual Sample – collected as a part of each employee’s annual physical exam

30 Routine Sample – samples from plant workers who are on a routine schedule for the purpose of 

insuring that airborne levels of uranium in the work place are being controlled within safe limits

40 Incident – Follow-up Sample – samples from employees involved in an event or circumstances 

which presents a potential for elevated exposure

49 An incident sample left at the end of the shift on the day of the incident.

50 Special Sample – samples collected as a part of a study to provide data related to uranium 

exposure and/or excretion characteristics of specific areas and/or conditions

5C Special Correlation Sample -

60 Termination/Retirement Sample – sample obtained from employee during post-employment physical 

exam.

70 Rehire – sample obtained during a former employee’s physical exam prior to being re-employed

R Recall Sample (example: 3R – Routine Sample Recall) – samples taken to verify positive sample 

results and/or to follow elimination pattern

VF Visitor First Sample

VR Visitor Routine Sample

VE Visitor Exit Sample
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Types of Samples – General

Code Description

BDL Below Detection Level

NA Not Applicable/Available

Table 5-21. FEMP sample type coding system and personnel assignment 

codes.

Sample Type Coding System – Specific

1 During 1st hour of shift

2 During 2nd hour of shift

3 During 3rd hour of shift

4 During 4th hour of shift

5 During 5th hour of shift

6 During 6th hour of shift

7 During 7th hour of shift

8 During 8th hour of shift

9 After completing work

Sample Type Coding System – Further Definition

5A Off the job, overnight composite sample

5B Off the job, overnight individual specimen

5C Sample to test for possible correlation with abnormal clinical lab findings

5D 24-hour individual sample from confined patients

5E 24-hour composite sample from confined patients

5F 24-hour individual sample from unconfined patients

5G 24-hour composite sample from unconfined patients

5H On-the-job individual sample collected in the work area

Personnel Assignment Codes

150 Plant 6 Inspection

200 General Project (Plant 8)

210 Plant 1

220 Plant 2 and 3

240 Plant 4

250 Plant 5 and 9

261 Plant 6 Rolling Mill

262 Plant 6 and 9

2623 Plant 6 and 9 Chemical

2625 Plant 6 and 9 Machining

270 Pilot Plant

280 Plant 8

430 Boiler Plant

436 Mechanical Department

452 Cafeteria

455 Security Department

462 Stores

463 Transportation Department

465 Garage

466 Service – Porters & Laundry
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5.3.6

As previously stated, the fundamental and primary bioassay for the first 35 years (1951 – 1986) of 

Fernald operational experience was urine analysis for uranium metal, reported in milligrams per liter. 

Radionuclides other than uranium have been analyzed on occasion through the years, predominantly 

by contract laboratories. Even for those special cases, they have been so few in number that the 

review of records for the TBD efforts did not reveal a dose record with non-uranium urinalysis results. 

There are records of special studies, but no documented intent to analyze for radionuclides other than 

uranium. The primary contract laboratory for FEMP in vitro analyses was United States Testing 

Company in Richland, Washington. Table 5-22 lists MDAs for various analyses.

In Vitro Procedures for Other Radionuclides

Table 5-22. MDAs for non-uranium radionuclides.

Type of Analysis Method Period Sample Frequency MDA*

Urine – plutonium Chemical extraction/ 

gross count

Prior to 1988 No routine schedule –

unknown

~0.1 pCi/sample

Extraction/alpha spec. 

count

1990s to 

present

Special study samples 

only

0.1 pCi/sample

Fecal – plutonium Extraction/alpha spec. 

count

1990s to 

present

Special study samples 

only

0.1 pCi/sample

Urine – thorium Extraction/alpha spec. 

count

1990s to 

present

Not performed 0.1 pCi/sample

Fecal – thorium Extraction/alpha spec. 

count

1990s to 

present

Only in cases of 

significant exposure

0.1 pCi/sample

* 0.1 pCi/sample was and is the contractual MDA for all analyses of this type, since the mid 1990s .

As early as 1958, the Fernald site reported internal dose experience to the AEC in an annual report. 

Table 5-23 summarizes the data from 1958 to 1966. The data provide instructive indication of 

recorded annual urinary experience in summary form.

Table 5-23. Urinary uranium averages summary.

Urine Average 

(μg L-1)

* Total Number of Exposed Workers in Each Category

1958 1959 1960 1961 1962 1963 1964 1965 1966

25-30 22 51 125 75 25 26 18 5 3

31-35 18 37 44 41 5 9 9 1 1

36-40 17 21 25 18 5 1 6 - -

41-45 9 5 15 6 1 1 - - -

46-50 7 4 6 3 1 - - 1 -

51-55 5 3 7 1 - - - - -

56-60 2 1 3 1 - - - - -

61-70 3 4 4 - - - - - -

71-90 3 5 - - - - - - -

91-110 3 3 - - - - - - -

111-155 45 2 - - - - - - -

Total 94 136 229 145 37 37 33 7 4

* Various annual reports reported the units in mg L-1, which is an obvious typographical error. The permissible urine 

concentration, averaged throughout the year, was 0.05 mg L-1.

5.3.7

Lung counting became available to FEMP in 1968 in the form of the Mobile In Vivo Radiation 

Monitoring Laboratory. The mobile van visited the Fernald plant on a routine schedule and counted 

the workers on a schedule based on their internal exposure potential and their urine sampling results. 

When lung counting became available, the annual reports to the AEC listed the number of workers

In Vivo Analysis
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who exceeded 50% of the MPLB and the calculated annual doses to the lung in rem. The uranium 

MPLB used was 0.0175 μCi, assuming 1% enrichment. Table 5-24 presents the typical reporting of 

internal dose as determined by direct lung counting.

Industrial Hygiene & Radiation Department Internal Deposition Action Levels procedures from about 

1970 indicate actions related to the determination of percent maximum permissible lung burden to 

either uranium or thorium. Uranium-235 was detected primarily by the emission of its 186 keV 

photon. Uranium-238 was calculated from measurement of the Th-234 progeny assumed to be in 

equilibrium with the U-238. Thorium-232 and Th-228 activities were determined based on equilibrium 

assumptions and detection of their progeny, most likely Ac-228 for Th-232, but Pb-212 may have 

been used for assessment of both Th isotopes. Thorium-230 is not readily detectable by in vivo

measurements. There appeared to be no attempt to detect TRU contaminants with the MIVRML. In 

fact, the only determination made with the mobile van was a quantification of the uranium lung burden 

in micrograms of uranium, with the assumption of 1% enrichment and of occasional thorium lung 

burdens as indicated by some claim records.

Table 5-24. Typical reporting of 

internal dose determined by direct lung 

count.

Year Number Above 50% MPLB

1974 21

1975 21

1976 9

1977 8

1978 10

1979 13

1980 6

The results from the MIVRML were calibrated in μCi of 235U and reported in mg of uranium in the lung, 

which was translated to MPLB based on the assumed enrichment (generally 1%). The percent body 

burden was then multiplied by 15 rem to obtain the assigned annual dose in rem.

The workers, who had known exposures to high air concentrations, had high urine results or were 

involved in an incident, were counted on first priority each time the MIVRML visited the site. Other 

workers were counted based upon their job exposure potentials, as shown in Table 5-25:

Table 5-25. Typical MIVRML counting schedule at Fernald in the 1970s.

Labor Category Description In vivo Counting Schedule

All chemical operators Once per year

Members of Project Labor Pool Once per year

Mechanical Department crafts During each MIVRML visit, 25% of the employees in these classifications 

were scheduled to be counted, and each worker would be counted at 

least once during a 2-yr period.

Mechanical Department laborer

Laundry group

Industrial truck operator

Locomotive operator

Switchman

Graphite shop machinist

Machine tool operator

Degreaser

Crane operator

Stamper
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Labor Category Description In vivo Counting Schedule

Plant 6 laborer

Furnace operator heater

Mill man

Decontaminator

Transportation laborer

Cafeteria Salaried personnel and workers in these classifications were not 

routinely counted because of low chronic exposure and low potential for 

unobserved acute exposures.

Water treatment group

Power plant group

Heavy equipment operator

Motor vehicle operator

Stores Warehouse attendant

Checker

Industrial mechanic

Security Police officer

Porter

Toolmaker

Machine set-up

Tool room machinist 

Gauge set up

Inspector

5.3.8

The In Vivo Examination Center (IVEC) operated at the FCP from 1989 to 2001, a subject with a 2.5

cm chest wall thickness had the MDAs listed in Table 5-26 at the 95% confidence interval for a 3,600 

sec count. The previous mobile counting system, which serviced Fernald from 1968 to 1989, 

provided reports to the site. However, no system performance characteristics have been discovered 

to date. Table 5-26 lists the MDAs for FEMP in vivo analyses.

In Vivo MDA

Table 5-26. Uranium in vivo MDAs for the lung.

Radionuclide Facility MDA (nCi)

Depleted U

MDA (mg)

Natural U

MDA (mg)

2% enriched

MDA (mg)

U-235 MIVRML

1968 to 1989

*

100μg(20%)

U-238 *6.5 mg(1%)

Th-232 6 mg thorium

U-238(Th-234) FCP IVEC

1989 to 2001

2.5 7.4 7.5 7.6

U-235 0.18 36.0 11.3 4.2

Pu-239 190

Am-241 0.25

Th-232 (Ac-228)** 1.2

Th-232(Pb-212) 0.3

Ra-226 ≥ 3.0

* The recorded MDA reported in Scott et. al. (1969, p. 169). Values reported in a claimant record were below 10% MPLB, 

which implies an MDA of less than 1 nCi total 1% (assumed) U-235.

** The degree of equilibrium is seldom known in a plant setting because these daughter products constitute a complex decay 

chain and the first controlling daughter has such a long half life (5.75 yr). As a consequence, in vivo measurements are used 

only as a screening technique.
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REFERENCES

Bassett, A. C., et al., 1989, Health and Safety Provisions for Recycle Uranium Processing at the 

FMPC, Fernald Environmental Management Project, Fernald, OH.

Cited References

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Health Physics, volume 2.

DOE (U.S. Department of Energy), 1987, Radiological Controls for Processing Out-of-Specification 

Uranium, Standard Operating Procedure, SOP 22-C-101.

DOE (U.S. Department of Energy), 1996, Highly Enriched Uranium Working Group Report, DOE/EH0525.

DOE (U.S. Department of Energy), 2000, DOE Ohio Field Office Recycled Uranium Project Report.

Dolan, L. C., and C. A. Hill, 1988, History of FMPC Radionuclide Discharges, Addendum to FMPC2082. 

ICRP (International Commission on Radiological Protection), July 1994, ICRP Publication 68, Dose 

Coefficients for Intakes of Radionuclides by Workers.

Krach, R. and Crawford, R.M., Integrated Hazards Analysis for Operable Unit 4 (OU4) Silos, 40000-

RP-0027, Rev.2, May 1998 

RAC (Radiological Assessments Corporation), 1995, Radionuclide Source Terms and Uncertainties, 

The Fernald Dosimetry Reconstruction Project, Tasks 2 & 3.

Rielly, P. S., 2001, INEEL M & O Contractor Technical Basis for Internal Dosimetry, General 

Technical Basis and Facility Specific Documents, Idaho Operations Office, Idaho Falls, Idaho, 

May 30.

Scott, L. M., et al., 1969, “Design and Development of a Mobile In vivo Radiation Monitoring 

Laboratory,” American Industrial Hygiene Association Journal, March-April 1969.

Scott, L. M., and C. M. West, 1964, Detection and Evaluation of Uranium Exposures, International 

Atomic Energy Agency.

Steckel, L. M. and C. M. West, 1966, Characterization of Y-12 Uranium Process Materials Correlated 

with In Vivo Experience, Y-1544-A.

Tomes, T., 2001, Technical Basis for Internal Dosimetry at the Fernald Closure Project, SD-2006, 

Fernald Closure Project, Fernald, OH.

Warner, W. T., March 16, 1966, letter to C. L. Carl, Report of UF6 Gas Release on February 14, 1966 

at the Feed Materials Production Center, Fernald, Ohio.

West, C. M., L. M. Scott, and N. B. Schultz, 1979, “Sixteen Years of Uranium Personnel Monitoring 

Experience – In Retrospect,” Health Physics, volume 36.
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Author unknown, September 1995, Technical Basis for Workplace Air Monitoring, SD-1024.

Other References

Bihl, D., 2003, Technical Basis Document for the Hanford Site – Occupational Internal Dose, NIOSH 

Dose Reconstruction Project.

Blalock, W. O., 1993, “Fluorometric fusion method MDL of 0.014 mgU/l”, Corrective Action Report, 

93:0008, January 26.

Boback, M. W., and R. A. Dugan, 1979, Internal Deposition Action Levels, Internal NLO Procedure, 

National Lead of Ohio, Inc., Fernald, OH.

Christiansen, H., 1988, Determination of Uranium in Urine: Fluorometric Fusion Method, Standard 

Operating Procedure.

Cofield, R. E., 1959, In Vivo Gamma Counting Method of Determining Uranium Lung Burden in 

Humans, Y-1250.

DOE (U.S. Department of Energy), 1999, DOE Standard Internal Dosimetry, DOE-STD-1121-98, 

Washington, D.C.

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Results, May 1987.

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FEMP sample data sheets from United States Testing Corp (UST) (mostly air sample for TRU, 

thorium and others).

Miscellaneous References

FEMP air sample data sheets (mostly on uranium contents, with thorium records included).

FEMP air, water, soil and radon sample data sheets from the K-65 Area

Selected examples of FEMP dosimetry data records in claim files.
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GLOSSARY

Class

The respiratory tract classification scheme in ICRP 30 for inhaled material according to its rate 

of clearance from the pulmonary region of the lung.. Materials are classified as D (days), W 

(weeks), or Y (years), according to how fast they clear the lungs; the half-times are class D in 

less than 10 days; class W in 10 to 100 days; class Y in more than 100 days. Recent 

recommendations in ICRP Report 66 modified the lung model and now instead of class D, W, 

and Y with lung absorption Types F (fast), M (moderate), and S (slow).

Becquerel

A unit of radioactivity equal to one disintegration per second

curie

A special unit of activity. One curie exactly equals 3.7 × 1010 nuclear transitions per second.

Depleted uranium

Uranium nuclide that has undergone a process to remove the 235U isotope, resulting in a 

relative increase in the percentage of the 238U isotope

Dose equivalent (H)

The product of the absorbed dose (D), the quality factor (Q), and any other modifying factors. 

The dose unit is rem.

Dose

A specific amount of energy from ionizing radiation absorbed per unit of mass.

Dosimetry

The science of assessing absorbed dose, dose equivalent, effective dose equivalent, etc., 

from external or internal sources of radiation.

Enriched uranium

Uranium which has been processed to contain a higher abundance of the isotope 235U.

exposure

As used for external dosimtry, exposure refers to a measure expressed in roentgens (R) of the 

ionization produced by photon radiation (i.e., gamma rays) in air. 

In vitro

In glass. Outside the living body and in an artificial environment. Typically used for bioassay of 

a contaminant in excreta, such as in fecal or urine samples.

In vivo

In the living; In the living body of a plant or animal. Bioassay counting analysis of 

radionuclides in the human body.

Isotope

Nuclides having the same number of protons in the nuclei (same atomic number), but having a 

differing number of neutrons (different mass number).
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maximum permissible lung burden (MPLB)

During the 1970s the occupational limit for radionuclides in the lung was expressed in terms of 

a quantity of the isotope that could be present at any given time to deliver 15 rem per year to 

the lung at the end of a 50-y period of chronic exposure.

Maximum permissible body burden (MPBB)

Same as MPLB except that any internal organ – or combination of a number of organs could 

be the subject of the determination.

Millirem

A unit of radiation dose equal to one-thousandth of a rem (see rem).

Microcurie

A measure of radioactivity equal to one-millionth of a curie.

Natural uranium

Uranium occurring in the natural state that has not been through a 235U enrichment process.

Rad

A unit of absorbed dose equal to 100 ergs/gm of any material.

Radiation

Energy transferred through air or some other media in the form of particles or waves (see 

ionizing radiation)

Radionuclide

A radioactive species of an atom characterized by the constitution of it nucleus specified by the 

number of protons, neutrons, atomic number, and mass number.

rem

A unit of dose equivalent, equal to the product of the rad absorbed dose and the quality factor.

Type

Refers to the rate of material absorption from the lung to the blood and includes types F (fast), 

M (moderate), and S (slow).

Transuranic (TRU) materials

Radioisotopes of nuclides having an atomic number greater than 92.