ORAU Team

NIOSH Dose Reconstruction Project

Technical Basis Document for the Fernald Environmental 

Management Project (FEMP) – Site Description

Document Number:

ORAUT-TKBS-0017-2 

Effective Date: 05/20/2004

Revision No.: 00

Controlled Copy No.: ______

Page 1 of 83

Subject Expert: Samuel L.T. Chu

Document Owner

Approval: Signature on File Date: 05/13/2004

Terry A. Kuykendall, TBD Team Leader

Approval: Signature on File Date: 05/19/2004 Judson L. Kenoyer, Task 3 Manager

Concurrence: Signature on File Date: 05/19/2004 Richard E. Toohey, Project Director

Approval: Signature on File Date: 05/20/2004

James W. Neton, OCAS Health Science Administrator

Supersedes:

None

TABLE OF CONTENTS

Section

RECORD OF ISSUE/REVISIONS .......................................................................................................

Page

5

ACRONYMS AND ABBREVIATIONS..................................................................................................6

2.0 SITE DESCRIPTION ..................................................................................................................7

2.1 Introduction................................................................................................................................. 7

2.1.1 Site Background and History ...........................................................................................8

2.1.2 Site Location and Facilities..............................................................................................8

2.1.3 Site Operations................................................................................................................8

2.1.4 Presentation for Dose Reconstruction Purposes ........................................................... 10

2.2 Production Facilities Activities and Processes........................................................................... 10

2.2.1 Pilot Plant – UF6 to UF4 Reduction Plant ....................................................................... 10

2.2.1.1 Processes and Activities ................................................................................. 11

2.2.1.2 Radiation Sources........................................................................................... 12

2.2.2 Plant 1 – Sampling Plant ............................................................................................... 14

2.2.2.1 Processes and Activities ................................................................................. 15

2.2.2.2 Radiation Sources........................................................................................... 15

2.2.3 Plant 2/3 – Refinery....................................................................................................... 18

2.2.3.1 Processes and Activities ................................................................................. 19

2.2.3.2 Radiation Sources........................................................................................... 20

2.2.4 Plant 4 – Green Salt Plant ............................................................................................. 22
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2.2.4.1 Processes and Activities ................................................................................. 23

2.2.4.2 Radiation Sources........................................................................................... 24

2.2.5 Plant 5 – Metals Production Plant.................................................................................. 27

2.2.5.1 Processes and Activities ................................................................................. 28

2.2.5.2 Radiation Sources........................................................................................... 29

2.2.6 Plant 6 – Metals Fabrication Plant ................................................................................. 33

2.2.6.1 Processes and Activities ................................................................................. 34

2.2.6.2 Radiation Sources........................................................................................... 34

2.2.7 Plant 7 – Hexafluoride Reduction Plant ......................................................................... 36

2.2.7.1 Processes and Activities ................................................................................. 36

2.2.7.2 Radiation Sources........................................................................................... 37

2.2.8 Plant 8 – Scrap Recovery Plant..................................................................................... 38

2.2.8.1 Processes and Activities ................................................................................. 39

2.2.8.2 Radiation Sources........................................................................................... 39

2.2.9 Plant 9 – Special Products Plant ................................................................................... 43

2.2.9.1 Processes and Activities ................................................................................. 44

2.2.9.2 Radiation Sources........................................................................................... 45

2.3 Waste Management Facilities Activities and Processes............................................................ 48

2.3.1 Production Wastes ........................................................................................................ 49

2.3.2 Production Area Waste Treatment and Storage Facilities.............................................. 50

2.3.2.1 Thorium Buildings and Overpacking Operations ............................................. 50

2.3.2.2 Ferrous Metal Scrap Pile................................................................................. 51

2.3.2.3 Pilot Plant Tank Farm...................................................................................... 52

2.3.2.4 Pilot Plant Warehouse..................................................................................... 52

2.3.2.5 KC-2 Warehouse ............................................................................................ 52

2.3.2.6 Refinery Storage Pad...................................................................................... 53

2.3.3 Waste Facilities Outside Production Area...................................................................... 53

2.3.3.1 Storage Silos .................................................................................................. 53

2.3.3.2 Low-Level Radioactive Waste Pit Area ........................................................... 56

2.4 FEMP REMEDIATION AND ENVIRONMENTAL RESTORATION ACTIVITIES SINCE 1989 ... 68

2.5 FEMP Radioactive Materials and Characteristics...................................................................... 70

2.5.1 Uranium and Uranium Compounds ............................................................................... 71

2.5.2 Recycle Uranium ........................................................................................................... 72

2.5.3 Thorium and Thorium Compounds ................................................................................ 73

2.5.4 Radon and Thoron ........................................................................................................ 74

2.5.5 Specific Activities and Conversion Factors .................................................................... 75

REFERENCES..................................................................................................................................76

GLOSSARY.......................................................................................................................................78
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APPENDIX 2A Default Model of FEMP Production Rate History for Dose Reconstruction 

Purposes. .......................................................................................................................82

APPENDIX 2B FEMP uranium emissions summary by plant (kg)............ ..................................83 

APPENDIX 2C Fugitive uranium and thorium emissions from wind erosion of Waste Pits 1, 2 and

3...................................................................................................................................83 

APPENDIX 2D Fugitive uranium and thorium emissions from wind erosion of Waste Pits 1, 2 and

3...................................................................................................................................85 

LIST OF TABLES

Table Page

2-1a Pilot Plant air dust concentration by operation or location...............................................12

2-1b Projected annual uranium intake estimates for Pilot Plant workers....................................13 

2-1c Projected annual thorium intake estimates for Pilot Plant workers....................................14

2-2a Plant 1 1955 airborne radioactivity concentration by worker groups.................................16

2-2b Projected annual uranium intake estimates by Plant 1 worker groups..............................17

2-3a Plant 1 1955 airborne alpha concentration by operation or location.................................17

2-3b Projected annual uranium intake estimates for Plant 1 workers...................................... 19

2-4 Plant 2/3 gamma fluxes .........................................................................................21

2-5a Plant 2/3 airborne radioactivity concentrations in late 1955 by worker groups.....................21

2-5b Projected annual uranium intake estimates by Plant 2/3 worker groups............................21

2-6a Plant 2/3 airborne radioactivity concentration by operation or location in late 1955..............22

2-6b Projected annual uranium intake estimates by Plant 2/3 workers....................................23

2-7a Plant 4 airborne radiactivity concentrations by operation or location.................................24

2-7b Projected annual uranium intake estimates by Plant 4 workers.......................................28

2-8 Plant 5 gamma fluxes.............................................................................................29

2-9a Plant 5 airborne radioactivity concentration by operation or location.................................30

2-9b Projected annual uranium intake estimates by Plant 5 workers.......................................33

2-10 Plant 6 gamma fluxes.............................................................................................35

2-11a Plant 6 airborne radioactivity concentration by operation or location.................................35

2-11b Projected annual estimated uranium intake by Plant 6 workers.......................................36

2-12 Plant 7 airborne radioactivity concentration and uranium intakes by worker groups...............37

2-13a Plant 7 airborne radioactivity concentration by operation or location.................................38

2-13b Projected annual estimated uranium intake by Plant 6 workers.......................................38

2-14a Plant 8 average airborne radioactivity concentrations by operation or location....................40

2-14b Projected annual estimated uranium intake by Plant 8 workers.......................................43

2-14c Projected annual estimated thorium intake by Plant 8 workers........................................43

2-15 Plant 9 gamma fluxes.............................................................................................45

2-16 Plant 9 LeBlonde Rapid Bore skin doses....................................................................45

2-17a Plant 9 airborne radioactivity concentration by operation or location.................................47

2-17b Projected annual estimated uranium intake by Plant 9 workers.......................................48

2-17c Projected annual estimated thorium intake by Plant 9 workers........................................48

2-18 Location and number of thorium drums......................................................................49

2-19 Dose estimates for thorium overpacking operations......................................................50
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2-20 Average exposure for Building 68 and 67 overpacking staff (800 drums).......................... 51

2-21 Summary of dispersible thorium in a typical thorium drum..............................................51

2-22 Expected airborne dust during the overpacking operation..............................................51

2-23 Isotopic composition of K-65 Silo 1...........................................................................54

2-24 Isotopic composition of K-65 Silo 2...........................................................................54

2-25 Isotopic composition of K-65 Silo 3...........................................................................55

2-26 Waste Pit 1 isotopic content....................................................................................56

2-27 Airborne uranium concentration at Waste Pit 1 by years................................................58

2-28 Waste Pit 2 isotopic content....................................................................................58 

2-29 Airborne uranium concentration at Waste Pit 2 by years................................................59

2-30 Airborne thorium concentration at Waste Pit 2 by years................................................60 

2-31 Waste Pit 3 isotopic content....................................................................................60

2-32 Airborne uranium concentration at Waste Pit 3 by years................................................61 

2-33 Airborne thorium concentration at Waste Pit 3 by years................................................61

2-34 Waste Pit 4 isotopic content....................................................................................61

2-35 Airborne uranium concentration at Waste Pit 4 by years................................................63

2-36 Airborne thorium concentration at Waste Pit 4 by years................................................64

2-37 Waste Pit 5 isotopic content....................................................................................64

2-38 Airborne uranium concentration at Waste Pit 5 by years................................................65

2-39 Airborne thorium concentration at Waste Pit 5 by years................................................66

2-40 Waste Pit 6 isotopic content....................................................................................65

2-41 Airborne uranium concentration at Waste Pit 6 by years................................................67

2-42 Clearwell isotopic content.......................................................................................67

2-43 Burn Pit isotopic content.........................................................................................67

2-44 Abbreviated history timeline and remediation and environmental restoration activities since 

1990...................................................................................................................................68

2-45 Uranium compounds in the former production area......................................................71

2-46 Relative activity of uranium mixtures.........................................................................72

2-47 Physical properties of radon and its progeny...............................................................74

2-48 Physical properties of thoron and its progeny..............................................................75

2-49 Conversion factors and isotopic and radionuclide activity fractions for recycled uranium.......76

LIST OF FIGURES

Figure Page

2-1 The FEMP Site...........................................................................................................9
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RECORD OF ISSUE/REVISIONS

ISSUE 

AUTHORIZATION 

DATE

EFFECTIVE 

DATE

REV. 

NO. DESCRIPTION

01/29/2004 01/29/2004 00-A New technical basis document for the Fernald 

Environmental Management Project (FEMP) – 

Site Description. Initiated by Samuel Chu.

03/23/2004 03/23/2004 00-B Incorporates comments from OCAS and Task 5 

reviewers. Initiated by Samuel L. T. Chu.

04/28/2004 04/28/2004 00-C Incorporates additional comments from reviewers. 

Initiated by Samuel L. T. Chu.

05/13/2004 05/13/2004 00-D Incorporates minor comments from OCAS 

reviewers. Initiated by Samuel L. T. Chu.

05/20/2004 05/20/2004 00 First approved issue. Initiated by Samuel L. T. 

Chu.
Effective Date: 05/20/2004 Revision No. 00 Document No. ORAUT-TKBS-0017-2 Page 6 of 83

ACRONYMS AND ABBREVIATIONS

AEC Atomic Energy Commission

ALAP as low as practicable

AMS air monitoring station

BZ breathing zone

DOE U.S. Department of Energy

DOT U.S. Department of Transportation

EEOICPA Energy Employees Occupational Illness Compensation Program Act of 2000

FEMP Fernald Environmental Management Project

FEMP Feed Material Production Center

GA general area

MAC maximum allowable concentration

MCW Mallinckrodt Chemical Works

MTU metric tons uranium

NIOSH National Institute for Occupational Safety and Health

NPR New Production Reactor

NTS Nevada Test Site

OEPA Ohio Environmental Protection Agency

PCB polychlorinated biphenyl

PHA preliminary hazard analysis

POOS plutonium out of specification

RMI Reactive Metals, Inc.

RTS Radon Treatment System

R-O-R reduction-oxidation-reduction

SERF Slightly Enriched Recovery Facility

STC strong tight container

TBD technical basis document

TRU transuranic

USC United States Code

WL working level
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2.0

This Technical Basis Document (TBD) of the Fernald Environmental Management Project (FEMP) site 

profile describes the facilities and processes that occurred at FEMP over the life of operations and the 

radioactive sources associated with those facilities and processes. This TBD provides the technical 

basis of the FEMP site, facilities and processes for the purpose of supporting estimates of 

occupational worker radiological dose based on exposure to the conditions and materials in the 

facilities.

SITE DESCRIPTION

Technical Basis Documents and Site Profile Documents are general working documents that provide 

guidance concerning the preparation of dose reconstructions at particular sites or categories of sites. 

They will be revised in the event additional relevant information is obtained about the affected site(s). 

These documents may be used to assist NIOSH in the completion of the individual work required for 

each dose reconstruction.

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

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

employer facility” or a “Department of Energy facility” as defined in the Energy Employee 

Occupational Illness Compensation Program Act of 2000 [42 U.S.C. § 7384l (5) and (12)]. 

For the National Institute for Occupational Safety and Health (NIOSH) dose reconstruction program 

and this TBD, the following criteria are applicable:

• TBDs are being developed to support the radiation dose reconstruction efforts. Therefore, the 

only information included in this TBD is that which will aid the dose reconstructor in that effort.

• This TBD is structured to support dose reconstructions that require research and analyses that 

might include determining or assuming specific characteristics of the monitoring procedures; 

identifying events or processes that were unmonitored; identifying types and quantities of 

radioactive materials involved; evaluating production processes and safety procedures 

employed; and identifying potential locations and activities of exposed persons.

2.1 INTRODUCTION

The FEMP, formally known as the Feed Materials Production Center (FEMP), was a large-scale, 

integrated facility that produced uranium metal products (e.g., derbies, ingots, billets, and fuel cores) 

used as feed materials in U.S. Department of Energy (DOE) defense programs. The site production 

mission has ended and the site has been undergoing remediation and cleanup since 1989. The 

facility officially closed as a production facility in 1991 and DOE changed its name to Fernald 

Environmental Management Project (FEMP) to reflect the current mission of site cleanup and 

remediation. FEMP will be used throughout this TBD regardless of any reference to operational 

period of the Site. The descriptions in this TBD provide technical bases that the dose reconstructor 

can use to support historic dose reconstruction analyses, including situations for which monitoring 

data are unavailable and/or other methods might not be amenable for claimant dose reconstruction. 

For this purpose, this TBD compiles available and obtainable relevant operational radiological 

characterization data for FEMP facilities and processes that can be used as technical bases to 

describe specific radiological conditions that workers may have experienced during their tenure at the 

FEMP.
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2.1.1

The FEMP began operations at the Pilot Plant in 1951, and was fully operational by the end of 1954. 

Its primary function was to convert uranium ore concentrates and recycled materials to either uranium 

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

assorted uranium enrichment. These products were prepared for use as production reactor fuel cores 

and target fuel elements. In addition, small amounts of thorium were processed. 

Site Background and History

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

1964 began to decline to a low in 1975 of approximately 1,230 MTU. During the 1970s, consideration 

was given to closing the FEMP. From 1972 through 1979, the staffing level, which peaked at 2,891 in 

1956, slowly declined from 662 to 538. In FY 1981, FEMP received direction to plan the restoration of 

the site to accommodate projected product requirements approaching the original production capacity. 

Significantly increased production levels, rapid staff buildup in many areas and implementation of a 

major facilities restoration program followed. Accomplishments since that time included production 

output of three times the 1979 level and a staff increase from 538 to more than 1,000.

The production of uranium metal at the FEMP ended in July 1989 to concentrate on site cleanup 

efforts. Current operations are completely dedicated to site remediation and restoration. 

2.1.2

The FEMP is near the unincorporated village of Fernald in the Great Miami River Valley, about 20 

miles northwest of Cincinnati in southwestern Ohio. The FEMP site covers an area of 1,050 acres. 

The production area encompasses approximately 136 acres in the center of the site.

Site Location and Facilities

Figure 2-1 shows the FEMP site in 2002. Historically, production area facilities included nine separate 

plants, the pilot plant, ancillary buildings, and administrative buildings that are connected with a 

network of roadways. These facilities, along with concrete storage pads, gravel ground cover, railroad 

access, sanitary landfill, and metal scrap piles, are surrounded by security fencing. Outside the 136-

acre fenced production area, the Waste Storage Area includes six low-level radioactive waste storage 

pits, two earthen-bermed concrete silos containing K-65 residues (high specific activity radiumbearing residues), one concrete silo containing metal oxides, and all affected adjoining areas. This 

area also includes two fly ash piles approximately 3,000 feet south-southwest of the waste storage 

area, as well as the burn pit between Pits 3 and 4. Production operations were handled in Plants 1 

through 9 and the Pilot Plant. Some of the plants and buildings that existed while the site was in 

operation no longer exist. As of 2002, Plants 1, 4, 5, 6, 7, and 9 have been dismantled, and others 

are in various phases of dismantlement or planning. The following sections provide a historical 

overview of the facilities and processes at the FEMP with respect to information and data to support 

the dose reconstruction efforts.

2.1.3

The FEMP used a wide variety of chemical and metallurgical process steps to support the primary 

mission of supplying metallic fuel cores for production reactors at Richland, Washington, and 

Savannah River, South Carolina. In addition, some metal was shipped directly to DOE facilities at 

Oak Ridge, Tennessee, and Rocky Flats, Colorado.

Site Operations

The first step in the FEMP production process was the purification of uranium. In the early years, the 

site processed uranium ore, including pitchblende ore from the Belgian Congo, through a series of 

chemical processes. Later, it extracted uranium from scrap metal or recycled material (e.g., floor 
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sweepings, dust collector and production residues) received from onsite operations and other DOE 

complex sites.

Figure 2-1. The FEMP Site (as of 2002).

FEMP uranium production began with ore concentrates, recycled uranium from spent reactor fuel, or 

various uranium compounds as feed materials. Impure starting material was dissolved in nitric acid to 

produce a crude uranyl nitrate solution (UNH) for solvent extraction purification. Purified UNH was 

concentrated by evaporation and thermally denitrated to uranium trioxide (UO3), or orange oxide. 

Orange oxide was converted to uranium tetrafluoride (UF4), or green salt, for reduction to metal. In 

addition, green salt was produced from uranium hexafluoride (UF6) received from other DOE sites. 

Uranium metal was produced by reacting UF4 and magnesium metal in a refractory-lined reduction 

vessel. The uranium metal was shaped in forms called derbies, each weighing 136 to 168 kg. This 

primary uranium metal was remelted with scrap uranium metal to yield a purified uranium ingot that 

was extruded to form rods or tubes. Sections were cut and machined to final dimensions. These 

machined cores were shipped to other DOE sites for canning and final assembly into reactor fuel 

elements. Since the late 1960s, all cylindrical ingots were center drilled at FEMP and sent off the site 

for extrusion. Most of the extruded tubes were returned to FEMP for heat treatment and final 

machining before they were shipped for use at other DOE sites.

Periodically, small amounts of thorium were processed. Thorium production steps, in general, were 

similar to those followed in uranium production. Final products were purified thorium nitrate solution, 
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solid thorium compounds, or metal. Since 1972, the FEMP served as the thorium materials repository 

for DOE. Approximately two-thirds of the material in the repository was processed at the FEMP. The 

remainder originated at other DOE facilities.

FEMP supported the Richland N-Reactor by processing enriched uranium scrap residues generated 

during the metal production steps plus Purex UO3 recycles. Enriched UF6 was required to replenish 

the 235U consumed in N-Reactor operations. All three material types were converted to UF4, which 

was the starting point for metal production operations. Production of depleted uranium metal for the 

Savannah River Mark 31 stream, the Y-12 Plant, and Rocky Flats began with UF4 from inventories.

The uranium processed was depleted, natural, or slightly enriched. The average content was close to 

natural uranium isotopic ratios. 235U enrichments in raw materials ranged up to approximately 20%. 

However, the maximum product enrichment was 1.25% and the average product was slightly 

depleted. In general, 2% enriched uranium was the highest enrichment processed in significant 

quantities. Small quantities of fission products (e.g. 90Sr, 137Cs, and 99Tc) and transuranics (e.g., 

neptunium, plutonium, and americium) were present in some recycled fuel. Raw feed materials 

containing very low concentrations of plutonium were used until as late as May 1986.

2.1.4

The following sections present description of each individual FEMP facilities and their associated 

processes in a manner that aids the historic dose reconstruction analyses, making use of available 

and obtainable site data and information. Each facility description contains 1) the physical description 

of the facility and its associated processes and activities and 2) the description of the radiation 

sources that include the technical basis for internal dose and external dose reconstruction analyses 

specifically for the described facility. Section 2.5, “FEMP Radioactive Materials and Characteristics” 

provides the discussions, radiological data, and technical basis germane to all of the FEMP facilities. 

(i.e., the specific activity and isotopic composition of uranium, uranium compounds, and recycled 

uranium, along with the principal gamma and beta emissions from uranium and processed uranium 

compounds measured in field studies and isotopic information on thorium and thorium compounds 

and radon and thoron).

Presentation for Dose Reconstruction Purposes

2.2 PRODUCTION FACILITIES ACTIVITIES AND PROCESSES

The FEMP was designed as a large-scale, integrated facility capable of converting uranium ore and 

recycled material into uranium metal through a series of chemical and metallurgical conversions. 

Production operations were conducted in Plants 1 through 9 and the Pilot Plant.

2.2.1

The Pilot Plant was located in the southwest corner of the fenced production area and was completed 

during 1951 and had a ground floor area of 23,500 ft2

. Its principal function was to convert UF6 to UF4

for use in the metal production process. In addition, the Pilot Plant had the capability of coating by 

plasma spray the crucibles used for casting uranium metal. The principal functions of the Pilot Plant 

were:

Pilot Plant – UF6 to UF4 Reduction Plant

• Reduction of UF6 to UF4 assaying up to 2.5% 235U;

• Purification and conversion of thorium nitrate solution to various thorium compounds; and
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• Miscellaneous operations for shot blasting uranium derby metal and plasma spray coating 

graphite crucibles.

This unit of the FEMP production complex had a wide range of chemical-metallurgical process 

equipment for the production of small quantities of uranium and thorium products. A variety of 

operations and assorted tasks occurred in this plant.

A considerable part of the metal produced at the FEMP was depleted uranium. The principal source 

for depleted uranium was the UF6 tails, the byproduct from uranium enrichment at the gaseous 

diffusion plants.

2.2.1.1 Processes and Activities

The Pilot Plant converted gaseous UF6 to solid UF4 green salt. The UF6 arrived in solid form in large 

cylinders. To produce UF4, the UF6 was first heated to form a gaseous compound and subsequently 

reduced to UF4. Three autoclaves and twin tube reactors were operated using a distributed control 

system. The UF6 flow rate from the autoclave to the reactor tubes was approximately 2,200 lb/hour 

(50 psig and 230oF). Both reactor tubes operated in the ranges of 900°-1200oF and 5-20 psig. Under 

these conditions, the facility was designed to produce 16 MTU per day at 100% on-stream factor. 

Enriched UF6 assaying as much as 2.5% 235U could be processed.

Other equipment was available for such operations as metallic shot preparation; plasma-spray coating 

of casting crucibles for controlling carbon pickup; shot blasting uranium derbies for removal of slag 

material; and salt-bath heat treating. Single uranium melts as large as 3 tons could be cast in the 

special 6 ft. diameter vacuum casting furnace.

Thorium ores, crushed thoria (powdery form of thorium dioxide) pellets, thorium oxalate, and other 

thorium materials were processed through the Pilot Plant thorium digestion and/or extraction system 

from about 1964 through 1980. These systems were used to produce a purified thorium nitrate 

solution that could be used as feed for the production of high-quality thorium compounds (thorium 

hydroxide) or metal. In addition, purified thorium nitrate solution was shipped as product. The 

digestion and extraction systems were not used continuously during the period they were in service. 

During some years the systems achieved throughput rates as high as 1 ton per day of purified thorium 

nitrate solution; in other years they operated periodically or not at all. Available production records 

indicate that 790.4 metric tons of thorium in the form of purified thorium nitrate solution were produced 

from 1966 to 1973. No production records have been located for other years. No production data are 

available for the digestion system.

The solvent extraction refining was followed by a multitank system for precipitation, several filters, and 

an oven-drying system to produce thorium oxalate, thoria gel (hydrated gel oxide), or thorium 

tetrafluoride. Metal was produced by further processing ThF4 through atmospheric furnaces for 

dehydration and metal reduction, vacuum furnaces for dezincing, and all auxiliary systems. A portion 

of the integrated extraction-precipitation system converted impure thorium nitrate solution to a 

storable dried thoria gel product for future DOE requirements. Production records indicate that 809 

thorium derbies weighing approximately 51 metric tons [total] of thorium metal were produced by this 

process between 1969 and 1971. 

Production records also indicate that 492 metric tons of thorium as thoria gel were produced from 

1966 to 1970. Production for 1964 and 1965 was estimated based on a linear extrapolation of the 

quantity produced in 1966 through 1970. The estimated total production from this process is 686 

metric tons assuming linear production from 1964 to 1970. From 1977 through 1979, production 
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records indicate that 350 metric tons of thorium as thoria gel were produced by similar process for 

storage. 

From 1971 through 1976, 153 metric tons of thorium as thorium oxalate were produced by the 

precipitation process at the Pilot Plant. 

In 1966, thorium nitrate tetrahydrate crystals were produced during a short-term test by use of a 

Sandvik Water Bed Conveyor. Purified thorium nitrate tetrahydrate solution was boiled in a tank to a 

defined concentration. The molten salt was fed onto a water-cooled conveyor where it crystallized, 

and product was drummed at the end of the conveyor. A total of 425 kilograms of thorium as thorium 

nitrate tetrahydrate crystals was produced during this test.

2.2.1.2 Radiation Sources

Radioactive sources in the Pilot Plant were primarily uranium and “recycle uranium” that consisted of 

UF6 and UF4 powder with mixed enrichment from depleted to 2.5% 235U by mass. Other radioactive 

sources included thorium metal and compounds. 

Table 2-1a lists the measured airborne radioactive material concentrations for operations and 

locations at the Pilot Plant. Table 2-1b and 2-1c provide the estimated potential annual uranium and 

thorium intakes by Pilot Plant operation personnel by the years as projected based on the measured 

data, respectively. 

Table 2-1a. Pilot Plant airborne radioactivity concentration by operation or location (year data taken 

as indicated). 

Operation or location

Measured Avg. 

conc.

α dpm/m3

Conc. 

w/Respiratorsa

Bq/ m3

Pilot Plant – wet area (uranium) (1968 data)

Breaking up UO2 tubes and putting metal into 5-gal cans 643.3 2.14E-01

Pump 6 ft west of Declad tank where decladding was conducted 49.7 8.28E-01

Pump 4 ft east of tank 51.3 8.55E-01

Unloading UO2 from declad tank (no respirator worn) 50.8 8.47E-01

Furnace Area (uranium) (1973 data)

General area on top of furnace 2’ south of cooling door 303.0 5.05E+00

General area south side of window to dumping tray 223.3 7.44E-02

BZ complete dumping cycle 670.0 2.23E-01

BZ dumping to take-out can 510.0 1.70E-01

BZ taking can out and putting empty can in 506.0 1.69E-01

BZ dumping tray removing full and replacing with empty 2030.0 6.77E-01

BZ changing trays from furnace to cooling 640.0 2.13E-01

Average uranium concentrations 516.0 7.77E-01

Thorium blending (thorium) (1968 data)

Separating thorium material from filter press (respirator worn) 400 1.33E-01

Pump 4 ft north of press 22.2 3.70E-01

Pump 5 ft west of press 80.0 1.33E+00

Pump 7 ft west of blender 49 8.17E-01

Unloading thorium from blender 61 2.03E-02

Unloading thorium from blender into small press, pump 9 ft southeast 

of blender 200 6.67E-02

Continuous precipitation operation, reactor tank vented by 4-in. duct

BZ small Oliver drum filter 48 8.00E-01
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Operation or location

Measured Avg. 

conc.

α dpm/m3

Conc. 

w/Respiratorsa

Bq/ m3

Pump 5 ft north of reactor tank 15.1 2.52E-01

4 ft south of thorium filter press 23.8 3.97E-01

Operator churning precipitate through funnel with broom handle 82 2.73E-02

Dumping drums into digest tank

Area background level 14.8 2.47E-01

Dumping two drums into port hole of tank (GA) 15.8 2.63E-01

Dumping thorium drums into tank (BZ) 76.2 1.27E+00

Removing dried thorium from oven

Removing from oven #2, respirator worn (BZ) 84.5 2.82E-02

Pump above #2 oven doors (GA) 14.5 2.42E-01

Unloading oven #1, respirator worn (BZ) 218.1 7.27E-02

Pump 6 in. in front open door of 07 dryer oven as it is loaded with 

wet thorium gel 542.7 1.81E-01

Pump 6 in. in front open door of 07 oven as dried thorium gel is 

removed 1,032.0 3.44E-01

General area – Stokes furnace platform 478.2 7.97E+00

Extraction area 

Operators desk 6.7 1.12E-01

Northwest corner 175 2.92E+00

West wall 150 2.50E+00

General area(1969 data)

D1-Platform 15 2.50E-01

Thorium dryer 19 3.17E-01

Mezzanine 15.5 2.58E-01

Chip burning in oxidation furnace

Desk in west of furnace 135.5 2.26E+00

Pipe 4 ft east of enclosure 507 8.45E+00

Packing chip into furnace crucible 116.7 3.89E-02

Average thorium concentration 170.2 1.07E+00 a

 A Decontamination Factor (DF) of 50 for wearing respirator was applied to Breathing Zone (BZ) concentrations for 

selected process activities.

Table 2-1b. Projected annual uranium intake estimates for Pilot Plant workers. 

Years

Adjustment Factorsa Average Uranium Conc.

Bq/m3

Average Annual 

Uranium Intakes

Bq/yrb 1968 Data 1973 Data

1951-1954 0.43 0.3 2.87E-01 6.89E+02

1955-1964 1.42 1.0 9.52E-01 2.28E+03

1965-1966 1.29 0.9 8.60E-01 2.06E+03

1967-1968 1.0 0.7 6.68E-01 1.60E+03

1969-1972 0.0 0.0 0.0 0.0

1973 1.42 1.0 9.52E-01 2.28E+03

1974 0.0 0.0 0.0 0.0

1975 0.35 0.25 2.37E-01 5.69E+02

1976 0.0 0.0 0.0 0.0

1977-1978 0.35 0.25 2.37E-01 5.69E+02

1979 0.0 0.0 0.0 0.0

1980 0.35 0.25 2.37E-01 5.69E+02

1981-1983 0.0 0.0 0.0 0.0
Effective Date: 05/20/2004 Revision No. 00 Document No. ORAUT-TKBS-0017-2 Page 14 of 83

Years

Adjustment Factorsa

Average 

Uranium Conc.

Bq/m3

Average Annual 

Uranium Intakes

Bq/yrb 1968 Data 1973 Data

1984 1.06 0.74 7.07E-01 1.70E+03 a

 The measured airborne radioactivity concentrations are adjusted for other years than the year in 

which the measurements are made, in accordance with the percentage of process system 

throughput capacity in one shift. In absence of specific throughput information, Appendix 2A is 

used as the default model for FEMP production rates for the derivation of the adjustment factors. 

The “ FEMP uranium emissions summary by plant” in Appendix 2B is used to identify operating 

years. 1973 data were taken at the end of a campaign, since then, it is assumed that the 

measured concentrations were resulting from full system capacity. b

 Radionuclide (e.g., Pu, Np, Tc and U isotopes) radioactivity fractions for recycled uranium is 

given in Table 2-54. 

 Table 2-1c. Projected annual thorium intake estimates for Pilot Plant workers. 

Years

Adjustment Factora

Average 

Thorium Conc.

Bq/m3

Average Annual 

Thorium Intakes

1968 Data 1969 Data Bq/yr

1964-1968 1.0 0.87 8.98E-01 2.16E+03

1969 1.14 1.0 1.03E+00 2.47E+03

1970 1.45 1.27 1.31E+00 3.14E+03

1971 0.18 0.16 1.63E-01 3.91E+02

1972 0.41 0.36 3.69E-01 8.86E+02

1973 0.15 0.13 1.35E-01 3.24E+02

1974 0.29 0.25 2.60E-01 6.24E+02

1975 0.01 0.01 9.16E-03 2.20E+01

1976 0.0 0.0 0.0 0.0

1977-1979 1.19 1.04 1.07E+00 2.57E+03 a

 The thorium concentrations were obtained by multiplying the uranium concentration estimates for 

a particular year by the thorium/uranium emission ratio using the emission rate estimates in 

Appendix 2B.

2.2.2

The Sampling Plant was completed in 1953 with a ground floor area of 22,040 sq. ft. This plant 

supported operations throughout the site. Its principal capabilities were:

Plant 1 – Sampling Plant

• Shipping, receiving, sampling and storing large amounts of depleted, normal, and enriched 

uranium materials in open and covered storage areas.

• Drying, crushing, milling, grinding and classifying feed materials for processing.

• Digesting enriched residues assaying 5% to 20% 235U in geometrically safe equipment.

• Opening unirradiated fuel pins containing enriched uranium dioxide pellets.

• Reconditioning steel drums for reuse and baling deteriorated drums for salvage.

The Sampling Plant weighed and sampled FEMP-generated and offsite recycle materials using 

various equipment to establish nuclear materials control data for accountability and control of 

fissionable materials processed at the FEMP. All incoming receipts were verified using 

nondestructive testing in the isotopic verification facility. The isotopic verification of incoming enriched 

uranium materials began in April 1972. At one time, the principal function of the Sampling Plant was 
Effective Date: 05/20/2004 Revision No. 00 Document No. ORAUT-TKBS-0017-2 Page 15 of 83

to obtain representative samples of large quantities of incoming ore concentrates. In 1988, however, 

the primary function was to store recycle materials until they were required by other production plants. 

The available storage area was 43,100 sq. ft. of open pads in a fenced area.

2.2.2.1 Processes and Activities

The sampling operation included a number of supporting operations. Several large-scale systems 

existed for crushing, grinding, and blending solid materials. These systems had a combined capacity 

of more than 10 tons per hour. Major equipment included hammer mills, ring-roll mills, and fallingstream samplers. Some of this equipment was shielded for handling radioactive materials. Special 

dust collecting and ventilating equipment permitted the processing of toxic and radioactive materials. 

Enriched uranium slag and selected recycle materials were processed through a ring-roller mill for 

reuse in the production of uranium derby metal or for chemical processing to UO3 in the refinery. This 

equipment could reduce particulate size to 95% minus 325-mesh at a rate of up to 9.1 tons per day.

A safe geometry digestion system provided a capability of safely processing enriched uranium 

materials assaying from 5% to 20% 235U at a rate of about 100 kg per shift.

Other facilities were available for opening fuel rods containing enriched UO2 pellets and powder, 

reconditioning steel drums, and scrap metal baling.

2.2.2.2 Radiation Sources

The principal types of radioactive materials received and sampled at the outset of Sampling Plant 

operations were:

• Q-11 material (code name for pitchblende), most of which was mined in the Belgian Congo. 

Because no processing was performed on this material before its receipt at the FEMP, it 

contained the equilibrium quantity of radium, thus requiring shielding against radiation during 

processing of the radium-bearing streams.

• Uranium in magnesium precipitate (MgX) resulting from a leaching process. 

• Q-11 containing less than 25% uranium. This low grade Q-11 was processed in the Belgian 

Congo and production was discontinued about the time the refinery was ready for operation. 

Only small quantities were received.

• Black oxide derived from concentrating low-grade uranium ores of U.S. and Canadian origin.

• Uranium in sodium salt (NaX) resulting from precipitation of leach liquors (from low-grade U.S. 

or Canadian source uranium ores) with sodium hydroxide.

• Canadian ores that contained thorium. These ores started arriving in 1956 and reached largequantity levels by June 1957.

• Uranium compounds of up to 5% enrichment. In 1965, the FEMP became the official receiving 

station for uranium compounds of up to 5% 235U furnished by licensees. With the startup of 

enriched uranium operations in the refinery in 1966, more than 1,500 safe mass batches of up 

to 10% 235U feed materials were prepared for drum digestion. This recycled uranium was 

known to contain traces of 237Np, 238Pu, and 239Pu.
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• Higher enrichment uranium compounds. In 1967, the isotopic level for the digestion operation 

was reduced to a maximum assay of 5% 235U as a further nuclear safety measure. Feed 

materials in the 5% to 10% assay range were set aside for the Safe Geometry Digestion 

System that started up in 1970 to provide the capability of safely processing enriched uranium 

materials up to 10% 235U. In subsequent years the range was increased to accept up to (but 

less than) 20% enriched material.

• Concentrate and residue drums. After the Mallinckrodt Chemical Works (MCW) Refinery was 

shut down in 1966, more than 20,000 drums of concentrates and residues were shipped to the 

FEMP and stored on the Plant 1 pad.

• Grand Junction, Colorado, shipments. Shipments of concentrates from Grand Junction, 

Colorado, began arriving in mid-1972. Shipments were resumed in 1974, and use of this 

material in the refinery was given priority to avoid storage.

Air samples were collected and reported in terms of the “maximum allowable concentration” (MAC). 

Early in the program the MAC was set at 70 dpm alpha/cubic meter. Table 2-2a lists the resulting 

daily weighted airborne alpha concentration in the vicinity of various worker groups and various job

locations from a 1955 exposure study of Plant 1 personnel. The calculated concentrations do not 

include any factor for reduction of exposure through the use of respirators. Table 2-2b provides the 

estimated potential annual uranium intakes by Plant 1 operation personnel by the years.

Table 2-2a. Plant 1 1955 airborne radioactivity concentration by worker 

groups. 

Groups × MACa

U Conc.

α dpm/m3

Conc. 

w/Respirator b 

Bq/m3

Sample Preparation Operators 23.4 1.64E+03 5.46E-01

General Plant Operators 6.7 4.69E+02 7.82E+00

Leadman 1.4 9.80E+01 1.63E+00

Production Foreman 0.53 3.71E+01 6.18E-01

Sample Preparation Foreman 0.4 2.80E+01 4.67E-01

General Foreman 0.32 2.24E+01 3.73E-01

Superintendent 0.21 1.47E+01 2.45E-01

Storage Pad Foreman 0.16 1.12E+01 1.87E-01

Storage Pad Operator 0.15 1.05E+01 1.75E-01

Storage Pad Laborer 0.14 9.80E+00 1.63E-01

Production Records Clerks 0.11 7.70E+00 1.28E-01

Administrative Clerk 0.07 4.90E+00 8.17E-02 a

 Maximum Allowable Concentration (MAC)- 70 α dpm/m3

b

 A Decontamination Factor (DF) of 50 for wearing respirator was applied to Breathing 

Zone (BZ) concentrations for selected process activities.

Table 2-2b. Projected annual uranium intake estimates by Plant 1 worker groups.

Groups

Annual Uranium Intakes Bq/yr b

1953-1954 

0.6a

1955-1964

1.0 a

1965-1970

0.7 a

1971-1975

0.6 a

1976-1981

0.25 a

1982-1984

0.74 a

Sample Preparation 

Operators 7.86E+02 1.31E+03 9.17E+02 7.86E+02 3.28E+02 9.70E+02

General Plant Operators 1.13E+04 1.88E+04 1.31E+04 1.13E+04 4.69E+03 1.39E+04

Leadman 2.35E+03 3.92E+03 2.74E+03 2.35E+03 9.80E+02 2.90E+03
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Groups

Annual Uranium Intakes Bq/yr b

1953-1954 

0.6a 1955-1964

1.0 a 1965-19700.7 a 1971-19750.6 a 1976-19810.25 a 1982-19840.74 a

Production Foreman 8.90E+02 1.48E+03 1.04E+03 8.90E+02 3.71E+02 1.10E+03

Sample Preparation Foreman 6.72E+02 1.12E+03 7.84E+02 6.72E+02 2.80E+02 8.29E+02

General Foreman 5.38E+02 8.96E+02 6.27E+02 5.38E+02 2.24E+02 6.63E+02

Superintendent 3.53E+02 5.88E+02 4.12E+02 3.53E+02 1.47E+02 4.35E+02

Storage Pad Foreman 2.69E+02 4.48E+02 3.14E+02 2.69E+02 1.12E+02 3.32E+02

Storage Pad Operator 2.52E+02 4.20E+02 2.94E+02 2.52E+02 1.05E+02 3.11E+02

Storage Pad Laborer 2.35E+02 3.92E+02 2.74E+02 2.35E+02 9.80E+01 2.90E+02

Production Records Clerks 1.85E+02 3.08E+02 2.16E+02 1.85E+02 7.70E+01 2.28E+02

Administrative Clerk 1.18E+02 1.96E+02 1.37E+02 1.18E+02 4.90E+01 1.45E+02 a

 Adjustment factors: The measured airborne radioactivity concentrations are adjusted for other years than the year in 

which the measurements are made, in accordance with the percentage of process system throughput capacity in one 

shift. In absence of specific throughput information, Appendix 2A is used as the default model for FEMP production rates 

for the derivation of the adjustment factors. b

 Radionuclide (e.g., Pu, Np, Tc and U isotopes) radioactivity fractions for recycled uranium is given in Table 2-54.

Table 2-3a lists the maximum measured alpha air concentrations by operation or location in Plant 1. 

Table 2-3b provides the estimated potential annual uranium intakes by Plant 1 operation personnel by 

the years. 

Table 2-3a. Plant 1 1955 airborne alpha concentrations by operation or location. 

Operation or location × MACa

U Conc.

Α dpm/m3

Conc. 

w/Respiratorb

Bq/m3

Outside operation

Deheading operation 9.0 6.30E+02 1.05E+01

Burning off battered drum lids 13.0 9.10E+02 1.52E+01

Pipe sampling operations 17.0 1.19E+03 1.98E+01

INX operations & locations

Deheading control room 0.63 4.41E+01 7.35E-01

Drum washing station 5.78 4.05E+02 6.74E+00

2nd level 9.90 6.93E+02 1.16E+01

3rd level 5.00 3.50E+02 5.83E+00

4th level 1.70 1.19E+02 1.98E+00

Packaging control operation 33.00 2.31E+03 7.70E-01

Packaging scale operation 11.57 8.10E+02 2.70E-01

Lidding drums operation 62.5 4.38E+03 1.46E+00

Q-11 operations & location

Hammering cake loose from drums at wash station 1.5 1.05E+02 1.75E+00

Pump area 0.46 3.22E+01 5.37E-01

Dryer area 0.40 2.80E+01 4.67E-01

Panel board area 0.31 2.17E+01 3.62E-01

2nd floor area 6.59 4.61E+02 7.69E+00

3rd floor area 1.2 8.40E+01 1.40E+00

4th floor area 4.09 2.86E+02 4.77E+00

Putting drums on conveyor for packaging 0.9 6.30E+01 1.05E+00

Packaging control operator 2.00 1.40E+02 2.33E+00

Lidding drums operation 120.00 8.40E+03 2.80E+00

Thaw tunnel control room 3.34 2.34E+02 3.90E+00

Sample preparation room
Effective Date: 05/20/2004 Revision No. 00 Document No. ORAUT-TKBS-0017-2 Page 18 of 83

Operation or location × MACa

U Conc.

Α dpm/m3

Conc. 

w/Respiratorb

Bq/m3

Spooning thorium nitrate into blender 10.3 7.21E+02 1.20E+01

Emptying Texas City ore out of blender 2.7 1.89E+02 3.15E+00

Pulverizing Texas City ore 0.54 3.78E+01 6.30E-01

Blending black oxide inside K-65 dry box 0.80 5.60E+01 9.33E-01

Blending black oxide outside (no longer done outside 

hood) 566.00 3.96E+04 1.32E+01

Pulverizing Q-11 inside hood 0.63 4.41E+01 7.35E-01

Cleaning Q-11 from screens outside hood on table 309.00 2.16E+04 7.21E+00

Blending Q-11 inside hood 2.00 1.40E+02 2.33E+00

Rolling Q-11 outside hood 6.00 4.20E+02 7.00E+00

Canning Q-11 outside hood 3.2 2.24E+02 3.73E+00

Putting Q-11 into screening pans outside hood 63.00 4.41E+03 1.47E+00

Riffling South African ore outside hood 96.00 6.72E+03 2.24E+00

Pulverizing African ore in DFC pulverizer inside hood 3.20 2.24E+02 3.73E+00

Cleaning African ore from DFC pulverizer inside hood 2.9 2.03E+02 3.38E+00

Blending and canning ammonium diuranate outside of 

hood 10.60 7.42E+02 1.24E+01

Pipe sampling ammonium diuranate outside hood 3.70 2.59E+02 4.32E+00

Grinding ammonium diuranate inside hood 2.5 1.75E+02 2.92E+00

Pipe sampling green salt outside hood 53.00 3.71E+03 1.24E+00

Blending green salt and canning under hood 126.00 8.82E+03 2.94E+00

Miscellaneous areas

Maintenance shop 0.63 4.41E+01 7.35E-01

Washroom 1.04 7.28E+01 1.21E+00

Average over Plant 1 areas and operations 36.59 2.34E+03 4.03E+00 a

 Maximum Allowable Concentration (MAC)- 70 α dpm/m3. b

 A Decontamination Factor (DF) of 50 for wearing respirator was applied to Breathing Zone (BZ) concentrations for 

selected process activities.

2.2.3

The Refinery was built in 1953 and had a ground floor area of 36,604 sq. feet. The refinery converted 

the Government stockpile of natural uranium ore concentrates to UO3. In 1988, the refinery operated 

intermittently on a sequential campaign mode to convert enriched recycled materials to oxide. The 

principal capabilities of the refinery were:

Plant 2/3 – Refinery

• Digesting enriched uranium recycled materials in nitric acid and storing blended feed solutions 

in stainless-steel tanks;

• Performing solvent extraction operations in stainless-steel, perforated-plate pulse columns to 

purify the uranium solution;

Table 2-3b. Projected annual uranium intake estimates 

for Plant 1 workers. 

Years

Adjustment 

Factora

1955 Data

Average Annual

Uranium Intakes

Bq/yr b

1953-1954 0.7 6.77E+03

1955-1964 1.0 9.67E+03
Effective Date: 05/20/2004 Revision No. 00 Document No. ORAUT-TKBS-0017-2 Page 19 of 83

Years

Adjustment 

Factora

1955 Data

Average Annual

Uranium Intakes

Bq/yr b

1965-1968 0.8 7.74E+03

1969-1971 1.0 9.67E+03

1972-1975 0.6 5.80E+03

1976-1981 0.25 2.42E+03

1982-1984 0.74 7.16E+03 a

 The measured airborne radioactivity concentrations are adjusted for 

other years than the year in which the measurements are made, in 

accordance with the percentage of process system throughput 

capacity in one shift. In absence of specific throughput information, 

Appendix 2A is used as the default model for FEMP production rates 

for the derivation of the adjustment factors. b

 Radionuclide (e.g., Pu, Np, Tc and U isotopes) radioactivity fractions 

for recycled uranium is given in Table 2-54. 

• Concentrating pure uranium solution by evaporation, and converting the uranium to UO3;

• Recovering nitric acid from NOx discharges from the digestion and denitration operations; and

• Recovering uranium from internal process waste solutions.

The three steps in the process of converting recycled materials to UO3 were digestion, extraction, and 

denitration. Auxiliary operations included the recovery of nitric acid from nitrogen oxides and uranium 

from internal process liquors.

2.2.3.1 Processes and Activities

Recycle materials from various sources were conveyed into agitated tanks for digestion using nitric 

acid. The resulting slurry consisted of acid insolubles and a digest liquor of impure uranyl nitrate and 

excess nitric acid. Adjustments were made to 235U and nitric acid concentrations. Following a check 

for proper solution concentrations, the blended feed was pumped to the extraction system. Lowgrade uranium slurries, primarily from the leaching of enriched uranium slag materials, required 

filtration and evaporation.

In the extraction columns, two immiscible liquids entered from opposite ends and were pumped 

counter-current to each other in a pulsating manner through a large number of perforated plates. The 

combined actions of the pulsations, the counter-current flow, and the turbulence created by perforated 

plates caused the two liquids to become intimately mixed. In the primary extraction columns, the two 

liquids brought into contact were the aqueous feed slurry and an organic solvent – a mixture of tributyl 

phosphate and kerosene. The uranyl nitrate in the feed slurry was preferentially attracted to the 

organic solvent in the presence of nitric acid. Most of the nitric acid and impurities were left behind in 

the aqueous raffinate, which excluded uranium.

A raffinate mixer-settler was used in series with the primary extraction columns to further reduce the 

uranium content of the aqueous waste stream leaving the columns. Extraction raffinate was sent to 

the general sump for neutralization with slaked lime. Neutralized raffinate was sent to interim waste 

processing facilities for conversion to a dry solid for offsite burial.

Additional purification of the uranium in the extract stream was achieved by scrubbing with a small 

counterflow of water. The aqueous stream from the scrubbing operation was recycled to the digestion 

operation for use in preparing feed slurries.
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Purified uranyl nitrate was recovered from the organic solvent stream by re-extraction with deionized 

water in perforated-plate pulse columns. In the absence of nitric acid, uranyl nitrate in the solvent was 

preferentially attracted to the water phase. After treatment with a sodium carbonate solution for 

removal of degradation products, the stripped solvent stream was reacidified and recycled to the 

primary extraction columns. The aqueous uranium nitrate product was sampled and analyzed to 

ensure conformance with strict chemical purity specifications before introduction to the denitration 

system. This aqueous uranyl nitrate product, at a concentration of approximately 100 g/l U, was sent 

to the denitration process.

In the denitration process, pure aqueous uranyl nitrate solution was concentrated by forced 

convective evaporation and high-pressure steam boildown to approximately 1,350 g/l U. It was 

calcined in batches in nominal 1,900-liter agitated denitration pots to yield UO3, the end product of 

refinery operations. After milling, the uranium trioxide was packaged in hoppers, with each hopper 

containing approximately 4.5 metric uranium tons, or into 55-gallon drums.

Nitric acid and uranium were recovered in auxiliary operations. The nitric acid recovery system 

consisted of two bubble cap absorption towers operating in series at atmospheric pressure. This 

system operated in conjunction with the refinery to recover nitric acid from the NOx fumes generated in 

the digestion and denitration areas and from other minor sources. The acid was returned to the 

digestion area for reuse in preparing feed slurries. The acid recovery operation reduced operating 

costs by saving acid, and reduced discharges to the environment.

Uranium in aqueous waste streams from the solvent treatment and cleanout operations was 

precipitated with magnesia and recycled to the digestion area.

In 1968, Plant 2/3 was used to process thorium as a thorium production test for a short duration. Few 

details are available regarding this process. Thorium nitrate crystals were produced in a denitration 

pot in Plant 2/3. Interviews with long-time employees indicated that this was a short-term operation; 

probably one pot of crystals was produced. Other records discuss the production of thorium oxide in 

Plant 2/3 by a process of denitration, redigestion, and drying.

2.2.3.2 Radiation Sources

The radioactive materials that were handled and processed in Plant 2/3 included:

• Enriched recycled uranium. This recycled uranium was known to contain traces of 237Np, 238Pu, 

and 239Pu. 

• Uranyl nitrate solution with variable concentrations. The solvent extraction product was at a 

concentration of 100 g/l U, and the evaporative concentrate was at 1,350 g/l U. 

• Solid powder of UO3. 

• Extraction raffinate (K-65 material). 

In March 1984, radiological measurements were made to characterize the radiation fields in the 

FEMP. The characterization showed a mixed photon and beta field with both low- and high-energy 

components. The low-energy beta component quickly became attenuated with distance from its 

source. The dose rate from the photon component was minor compared to the beta dose rate. 

Results of the gamma fluxes recorded at various stations applicable to Plant 2/3 from the Radiation 
Effective Date: 05/20/2004 Revision No. 00 Document No. ORAUT-TKBS-0017-2 Page 21 of 83

Field Characterization and Radiological Equipment Evaluation at the FEMP (Alvarez et al. 1984) are 

listed in Table 2-4:

Table 2-4. Plant 2/3 gamma fluxes. 

Spectrum 

number Location

Integrated gamma flux (photons/cm2-sec)

30-225 keV 675-1050 keV Ratio

358405 Beside UO3 barrel 538 159 3.4

378401 Open UO3 barrel 919 232 4.0

378411 Open U barrel; at 40 cm 891 218 4.1

A study of Plant 2/3 airborne alpha concentrations in late1955 indicated the daily weighted 

concentrations in the vicinity of the various groups and for the various jobs and locations in Table 2-

5a. The calculated concentrations do not include any factor for reduction of exposure through the use 

of respirators. Table 2-5b provides the estimated potential annual uranium intakes by Plant 2/3 worker 

groups by the years.

Table 2-5a. Plant 2/3 airborne radioactivity concentrations in late 1955 

by worker groups. 

Groups × MACa

U Conc.

α dpm/m3

Conc. b

w/Respiratorb 

Bq/m3

Recovery operator and helper 0.51 3.57E+01 5.95E-01

Leaderman 0.44 3.08E+01 5.13E-01

Sump operator 0.09 6.30E+00 1.05E-01

Foreman 0.53 3.71E+01 6.18E-01 a

 Maximum Allowable Concentration (MAC)—70 α dpm/m3. b

 A Decontamination Factor (DF) of 50 for wearing respirator was applied to Breathing Zone 

(BZ) concentrations for selected process activities.

Table 2-5b. Projected annual uranium intake estimates by Plant 2/3 worker groups.

Groups

Annual Uranium Intakes Bq/yr b

1953-1954

0.6a

1955-1964

1.0 a

1965-1970

0.7 a

1971-1975

0.6 a

1976-1981

0.25 a

1982-1984

0.74 a

Recovery operator and helper 8.57E+02 1.43E+03 1.00E+03 8.57E+02 3.57E+02 1.06E+03

Leaderman 7.39E+02 1.23E+03 8.62E+02 7.39E+02 3.08E+02 9.12E+02

Sump operator 1.51E+02 2.52E+02 1.76E+02 1.51E+02 6.30E+01 1.86E+02

Foreman 8.90E+02 1.48E+03 1.04E+03 8.90E+02 3.71E+02 1.10E+03 a

 Adjustment factors: The measured airborne radioactivity concentrations are adjusted for other years than the year in 

which the measurements are made, in accordance with the percentage of process system throughput capacity in one shift. 

In absence of specific throughput information, Appendix 2A is used as the default model for FEMP production rates for the 

derivation of the adjustment factors. b

 Radionuclide (e.g., Pu, Np, Tc and U isotopes) radioactivity fractions for recycled uranium is given in Table 2-54. 

Table 2-6a lists the maximum measured alpha concentration of radioactive dust by operation or 

location in Plant 2/3. Table 2-6b provides the estimated potential annual uranium intakes by Plant 2/3 

workers by the years.
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Table 2-6a. Plant 2/3 airborne radioactivity concentration by operation 

or location in late 1955. 

Operation or location

Ave. Conc.

dpm/m3

Conc. 

w/Respiratora 

Bq/m3

Hot Raffinate

Hot Tank Room 8 1.33E-01

Control Board, main floor 60 1.00E+00

K-65 Slurry Pump Area 70 1.17E+00

2nd level 5 8.33E-02

Oliver Filter Area 9 1.50E-01

Control Panel Area 46 7.67E-01

Raking Filter Trough 90 1.50E+00

Combined Raffinate

Main Level, Near Calciner 16 2.67E-01

Second Level 89 1.48E+00

Third Level 15 2.50E-01

Fourth Level 6 1.00E-01

Operating Batch Conveyor 50 8.33E-01

Nitric Recovery

Main Level 39 6.50E-01

2ndLevel 21 3.50E-01

3rd Level 8 1.33E-01

4th Level 8 1.33E-01

5th Level 5 8.33E-02

6th Level 15 2.50E-01

Control Room 62 1.03E+00

Wash Room 37 6.17E-01

Ozone Room 4 6.67E-02

Maintenance Office 17 2.83E-01

General Sump Tank Area, Ground Level 4 6.67E-02

General Sump Tank Area, Second Level 6 1.00E-01

General Sump Building, First Level 7 1.17E-01

General Sump Building, Second Floor 6 1.00E-01

Average over Plant 2/3 areas and 

operations 24.5 3.18E-01 a A Decontamination Factor (DF) of 50 for wearing respirator was applied to Breathing 

Zone (BZ) concentrations for selected process activities.

2.2.4

In October 1953, the Green Salt Plant began to produce green salt (UF4) from UO3. The ground floor 

area was 26,500 sq. ft. The UO3 was either produced in the refinery or recycled from other DOE 

sites. The primary functions of the Green Salt Plant are:

Plant 4 – Green Salt Plant

• Processing UO3 to uranium tetrafluoride, UF4, or green salt, in continuous-flow reactor banks 

designed and staged for gas-solid reactions

• Blending and packaging depleted UF4 for the Metal Production Plant (Plant 5)

• Operating the tank farm to supply production plants with bulk quantities of required liquid 

chemical agents
Effective Date: 05/20/2004 Revision No. 00 Document No. ORAUT-TKBS-0017-2 Page 23 of 83

Table 2-6b. Projected annual uranium intake estimates 

by Plant 2/3 workers.

Years

Adjustment 

Factora

1955 Data

Average Annual 

Uranium Intakes b

Bq/yr

1953-1954 0.7 5.34E+02

1955-1964 1.0 7.63E+02

1965-1968 0.8 6.11E+02

1969-1971 1.0 7.63E+02

1972-1975 0.6 4.58E+02

1976-1981 0.25 1.91E+02

1982-1984 0.74 5.65E+02

1985-1989 0.4 3.05E+02 a

 Adjustment factors: The measured airborne radioactivity 

concentrations are adjusted for other years than the year in which 

the measurements are made, in accordance with the percentage of 

process system throughput capacity in one shift. In absence of 

specific throughput information, Appendix 2A is used as the default 

model for FEMP production rates for the derivation of the adjustment 

factors. b

 Radionuclide (e.g., Pu, Np, Tc and U isotopes) radioactivity fractions 

for recycled uranium is given in Table 2-54. 

In the Green Salt Plant UO3 was converted to uranium dioxide, UO2, or brown oxide, by reducing it 

with hydrogen. The UO2 was converted to UF4 in a reaction with anhydrous hydrogen fluoride. The 

UO3 recycled from the reactor sites underwent double pass Reduction-Oxidation-Reduction (R-O-R) 

processing to achieve high yields of product UF4. 

2.2.4.1 Processes and Activities

Mobile hoppers delivered the orange oxide to stainless-steel fluid bed reactors, which were heated to 

approximately 1,100oF. Dissociated ammonia entered the bottom of the reactors through a gas 

diffuser. The hydrogen and nitrogen held the UO3 powder in suspension. Partially converted UO3

overflowed from the first fluid bed reactor into the second, where the reaction with hydrogen was 

completed. 

Hydrofluorination took place in groups of three heated horizontal ribbon-screw reactors, arranged in 

vertical stacks. Uranium dioxide entered at one end of the top reactor, and was conveyed slowly to 

the other end and stirred by a power-driven ribbon screw. The operating temperature was 

progressively higher for each reactor, starting at approximately 300oF at the first and ranging up to 

1,200o F at the third. Anhydrous hydrogen fluoride gas entered at the discharge end of the first and 

third reactors and flowed countercurrent to the powder flow. The UF4 product was weighed, blended, 

sampled for chemical analysis, and packaged in 38-liter cans for transportation to Plant 5.

Initial production throughput from Plant 4 did not progress as rapidly as expected. It was not until 

1956, when 5,029 MTU was achieved, that the design rate of 5,193 MTU/yr was approached.

Normal and enriched UF4 production decreased during 1967 and 1970 because of declining demand. 

Plant 4 operated only 4 months in 1971, and 2 months in 1972, on enriched UF4 production. The 

plant was idle during 1973.

In April 1974, production began on enriched UF4 for the New Production Reactor (NPR) Stream and 

continued through June. Another enriched campaign occurred from January through April 1975.
Effective Date: 05/20/2004 Revision No. 00 Document No. ORAUT-TKBS-0017-2 Page 24 of 83

The Plant 4 process was idle from the 1975 enriched campaign through 1977. A small campaign 

occurred in 1978 and the plant was again shut down in 1979. In 1980, the process was restarted and 

operated packaging depleted UF4 from T-hoppers into cans for the Mark 31 stream, processing UO3

and UF4 T-hoppers, and operating the Tank Farm.

In 1954, Plant 4 was used for a short campaign to produce dry ThF4 from the ThO2 dried and calcined 

in Plant 9 in hydrofluorination Bank 7. The ThF4 was returned to Plant 9 and used to produce thorium 

metal. This was a short-duration process due to mechanical difficulties in Bank 7. Production 

quantities are not available for ThF4 production in Plant 4.

2.2.4.2 Radiation Sources

The following is a chronology the history of radioactive sources in the feed materials; the primary 

radioactive sources are UO3 (orange oxide), UO2 (brown oxide), and UF6 (green salt) powders:

1. Initially it was planned that all UO3 feed would come from the FEMP refinery. However, in 

1957, large quantities of UO3 from the Port Hope Refinery in Canada became available for 

process, in addition to FEMP refinery output.

2. Recycling of low 236U (“300”) and high 236U (“500”) enriched materials began in 1960, and 

required complete segregation from standard UO3 and other enriched UO3 generation.

3. After the shutdown of the FEMP refinery in 1962, all UO3 feed for Plant 4 came from MCW, 

Port Hope, Hanford, and the Savannah River Site.

4. Beginning in 1965, enriched UO3 produced from the FEMP Slightly Enriched Recovery Facility 

(SERF) process was used as feed. Other UO3 feed that year included MCW normal pot, MCW 

normal fluid bed, Port Hope normal and recycle UO3 from Hanford and Savannah River Site. 

Receipt of these materials continued through 1971.

Table 2-7a lists the 1965 airborne radioactive material concentrations for various operations and 

locations at Plant 4. Table 2-7b provides the estimated potential annual uranium intakes by Plant 4 

workers by the years.

Table 2-7a. Plant 4 airborne radioactivity concentrations by operation or location in 1965. 

Operation or location

Ave. 

conc.

dpm/m3

Conc. w/Respirator

a

Bq/m3

Level 629

General southwest area 125.2 2.09E+00

General northwest area 103.3 1.72E+00

General northeast area 180.8 3.01E+00

General southeast area 200.3 3.34E+00

“Enriched” packaging station (no respirator worn)

Weighing empty cans before charging with UF4 154.3 5.14E-02

Charging cans 313.3 1.04E-01

Adjusting weight of packaged UF4 283.3 9.44E-02

Hanford orange oxide

Charging “T” hopper at Bank 2 east side (respirator worn) 74 2.47E-02

Charging “T” hopper at Bank 3 east side (respirator worn) 400 1.33E-01

Two operators changing mobile hopper Bank 1 175 5.83E-02

5 ft north of Bank 1 while changing hopper 86.5 2.88E-02
Effective Date: 05/20/2004 Revision No. 00 Document No. ORAUT-TKBS-0017-2 Page 25 of 83

Operation or location

Ave. 

conc.

dpm/m3

Conc. w/Respirator

a

Bq/m3

10 ft north of mobile hopper while being changed 100 3.33E-02

10 ft south of mobile hopper while being changed 160 5.33E-02

5 ft south of Bank 1 while changing hopper 185 6.17E-02

Two operators removing empty “T” hopper from Bank 3 130 4.33E-02

7 ft south of Bank 3 while changing hopper 350 1.17E-01

7 ft north of Bank 3 while changing hopper 62 2.07E-02

General level at Bank 3 slightly south 53.5 1.78E-02

On rail in front of Bank 2 205 6.83E-02

South end of east rail 14.5 2.42E-01

Removing empty T hopper from Bank 7 84 2.80E-02

Replacing full T hopper on Bank 7 280 9.33E-02

7 ft north of Bank 7 270 9.00E-02

5 ft south of Bank 7 170 5.67E-02

Removing T hopper from Bank 9 (no respirator) 510 1.70E-01

Putting full T hopper on Bank 9 77 2.57E-02

10 ft south of Bank 9 120 2.00E+00

10 ft north of Bank 9 63 1.05E+00

Two operators changing mobile hopper Bank 1 45 1.50E-02

10 ft north of Bank 1 while hopper was being changed 71 1.18E+00

10 ft south of Bank 1 while hopper was being changed 15.5 2.58E-01

East Reblend Dumping Station

Dumping UF4 69.57 2.32E-02

5 ft north of dumping station 62 1.03E+00

5 ft south of dumping station 25 4.17E-01

West Side Dumping Station- guillotine door open while dumping

Operator dumping 10-gal cans of UF4 150.3 5.01E-02

5 ft south of dumping station 67 1.12E+00

10 ft west of dumping station 10.3 1.72E-01

East Side Dumping Station

Operator dumping 10-gal cans of UF4 138 4.60E-02

5 ft north of dumping station 16.7 2.78E-01

10 ft south of dumping station 37.5 6.25E-01

East H2 Burnoff

7 ft northwest of East Burnoff 356.7 1.19E-01

15 ft west of East Burnoff 215.3 7.18E-02

7 ft southwest of East Burnoff 293.3 9.78E-02

2 ft from tempering unit East Burnoff (process) 15,775 5.26E+00

Level 619 (5th Level)

General northeast area 137.3 2.29E+00

General northwest area 212.8 3.55E+00

General southeast area 166.7 2.78E+00

General southwest area 114.3 1.91E+00

Smoking area 83.8 1.40E+00

Maintenance area 12.9 2.15E-01

HF Recovery area 9.7 1.62E-01

Enriched material Dumping Station

Dumping UF4 at dumping station north side 101 3.37E-02

7 ft south of dumping station 25.3 4.22E-01

7 ft north of dumping station 29.7 4.95E-01

Weighing empty cans at west charging station 35.3 5.88E-01

Dust Collector Drumming Station 0.00E+00
Effective Date: 05/20/2004 Revision No. 00 Document No. ORAUT-TKBS-0017-2 Page 26 of 83

Operation or location

Ave. 

conc.

dpm/m3

Conc. w/Respirator

a

Bq/m3

Checking and changing of drums (respirator worn) 1,090 3.63E-01

10 ft east during changing and checking of drums at drumming 

station 13.7 2.28E-01

10 ft west during changing and checking of drums 122.3 2.04E+00

“Enriched” UO3 dumping station (respirator worn)

Operator sampling reactor 1,493.3 4.98E-01

Delipping, pumping and relipping drum 669 1.12E+01

Enriched uranium packaging station 

5 ft southwest of UF4 charging station 27.3 4.55E-01

10 ft north of UF4 adjusting weigh station 133 2.22E+00

Level 608 (4th Level)

General southwest area (605 level) 47.2 7.87E-01

General northwest area (605 level) 259.5 4.33E+00

General southwest area (610 level) 21.2 3.53E-01

General northwest area (610 level) 28.2 4.70E-01

HF Recovery Area 12 2.00E-01

Level 597 (3rd Level)

General southeast area 13.7 2.28E-01

General northeast area 22 3.67E-01

General northwest area 15.3 2.55E-01

General southwest area 14.3 2.38E-01

Smoking area 24.3 4.05E-01

HF Recovery area 8.7 1.45E-01

Operator sampling reactor 1 22 7.33E-03

Operator sampling reactor 2 82 2.73E-02

Operator sampling reactor 7 100 3.33E-02

Operator sampling reactor 8 110 3.67E-02

Operator sampling reactor 9 13 4.33E-03

Level 588 (Mezzanine)

General southwest area 63 1.05E+00

General northwest area 32 5.33E-01

General southeast area 15.7 2.62E-01

General north east area 16 2.67E-01

HF Recovery Area 10.7 1.78E-01

Level 580 

General west side near water fountain area 60 1.00E+00

General north side near door area 127 2.12E+00

General east side near packaging station 49 8.17E-01

General south side near scale area 12.7 2.12E-01

East Packaging Station (enriched material)

Weighing empty can before charging with UF4 18 3.00E-01

Charging cans of UF4 42.3 1.41E-02

Adjusting weigh of package UF4 and wiping can 4.3 7.17E-02

8 ft south of UF4 canning station 5.0 8.33E-02

10 ft southwest of UF4 canning station 4.7 7.83E-02

West Canning Station 

Weighing empty can before charging with UF4 103.7 3.46E-02

Charging cans of UF4 23 7.67E-03

Adjusting weigh of package UF4 and wiping can 12.3 4.10E-03

8 ft north of UF4 packaging station 27.3 4.55E-01

10 ft northeast of UF4 packaging station 32.3 5.38E-01
Effective Date: 05/20/2004 Revision No. 00 Document No. ORAUT-TKBS-0017-2 Page 27 of 83

Operation or location

Ave. 

conc.

dpm/m3

Conc. w/Respirator

a

Bq/m3

Enriched uranium packaging station

10 ft north of UF4 adjusting weigh station 75.7 1.26E+00

5 ft southwest of UF4 charging station 42.3 7.05E-01

Charging cans 36 1.20E-02

Adjusting weight of packaged UF4 103.3 3.44E-02

Weighing empty cans before charging with UF4 56.7 1.89E-02

Blender operations

Operator collecting and changing sample bottles at 3 blenderenriched 48.5 1.62E-02

Operator collecting and changing sample bottles at 2 blender 133 4.43E-02

First Level

Ammonia Disassociator Area 21 3.50E-01

Administrative and offices 14.7 2.45E-01

HF Recovery Area 11.7 1.95E-01

Sump pump 2.7 4.50E-02

West pad 12 2.00E-01

East pad 6.0 1.00E-01

Average over Plant 4 areas and operations 261.5 6.80E-01 a

 A Decontamination Factor (DF) of 50 for wearing respirator was applied to Breathing Zone (BZ) 

concentrations for selected process activities.

2.2.5

Construction of this facility was complete in 1953. The ground floor area was 58,620 ft2

. At the Metal 

Production Plant, UF4, the product of Plant 4 and the Pilot Plant, underwent a thermite-type reaction 

with magnesium to produce uranium metal. The principal functions of the Metals Production Plant 

were:

Plant 5 – Metals Production Plant

• Reducing uranium tetrafluoride to high-purity depleted, normal, and enriched uranium derby 

metal with magnesium in electric resistance furnaces.

• Casting depleted uranium derbies and recycle metal scrap into ingots using vacuum melt 

induction furnaces.

• Sawing ingots to size.

• Sampling metal products for quality.

• Machining graphite into almost any shape using lathes, saws, milling machines, routers, and 

grinders

• Milling MgF2 slag byproduct for reuse in lining reduction pots
Effective Date: 05/20/2004 Revision No. 00 Document No. ORAUT-TKBS-0017-2 Page 28 of 83

Table 2-7b. Projected annual uranium intake estimates by Plant 4 

workers.

Years

Adjustment Factora

1965 Data

Average Annual Uranium

Intakes

Bq/yr f

1953-1954 0.8 1.31E+03

1955-1964 1.1 1.80E+03

1965-1968 1.0 1.63E+03

1969-1970 1.1 1.80E+03

1971 1.0 5.44E+02b

1972 1.0 2.72E+02c

1973 0.0 0.0 d

1974 1.0 2.72E+02c

1975 1.0 5.44E+02b

1976-1977 0.0 0.0 d

1978 1.0 2.72E+02e

1979 0.0 0.0 d

1980-1988 0.82 1.34E+03 a

 Adjustment factors: The measured airborne radioactivity concentrations are 

adjusted for other years than the year in which the measurements are made, 

in accordance with the percentage of process system throughput capacity in 

one shift. In absence of specific throughput information, Appendix 2A is 

used as the default model for FEMP production rates for the derivation of the 

adjustment factors. B

 Plant 4 operated 4 months in 1971 and 1975. c

 Plant 4 operated 2 months in 1972 and 1974. d

 Plant 4 idled in 1973, 1976 through 1977 and 1979. e

 Plant 4 processed a small campaign in 1978. f

 Radionuclide (e.g., Pu, Np, Tc and U isotopes) radioactivity fractions for recycled 

uranium is given in Table 2-54.

2.2.5.1 Processes and Activities

To begin the reduction process that produces uranium metal, UF4 and magnesium granules, totaling 

about 500 pounds, were blended and charged into a steel pot lined with magnesium fluoride slag. 

The pot was capped with slag to protect it from the intense heat of the reaction. Materials handling 

systems were included in the Productivity Retention Program to reduce worker contact with uranium 

materials in this area. The pot was fitted with a steel cover and heated in a resistance furnace to a 

temperature range of 1200° to 1500oF for 3 to 4 hours, until the contents reacted spontaneously. At 

this point, the internal temperature of the pot might have reached 3,000oF. Approximately 5 minutes 

after the reaction occurred, the pot was removed from the furnace and stored in an air-cooling tank for 

at least 1 hour, and then transferred to a water-cooling tank for several hours. After cooling, the 

contents were removed and the uranium mass, called a derby, was separated and cleaned by slag 

chipping hammers.

Standard depleted production derbies weighed approximately 167 kg U. Enriched derbies weighed 

approximately 152 kg U. Most of the derbies were transferred to the Metals Production Plant casting 

area, some were sent to the Special Products Plant casting area, and others were sent to other DOE 

sites. The MgF2 slag resulting from the thermite reaction was milled for reuse as liner material.

In the casting process, cleaned depleted or normal derbies together with recycle metal scrap were 

charged into a graphite crucible. The loaded crucible was placed in a vacuum induction furnace and 

heated for about 95 minutes at 130 Kw input to approximately 2,700oF, at which temperature it is 

ready to pour. A shear plug in the bottom of the crucible was broken to permit the molten metal to 
Effective Date: 05/20/2004 Revision No. 00 Document No. ORAUT-TKBS-0017-2 Page 29 of 83

flow into a heated graphite mold directly under the crucible. After cooling, the mold was separated 

from the ingot, cleaned, and prepared for reuse. Crucibles were flame-treated to oxidize residual 

uranium. On occasion, metal spills occurred from mold breakage. All product ingots cast in the 

Metals Production Plant were for supplying target element core requirements for reactors at the 

Savannah River Site. Slab billets were cast from derby metal for the Rocky Flats Plant.

As-cast ingots were cropped by sawing approximately 2 inches from the top section to remove 

shrinkage cavities and impurities that rose to the top of the melt after pouring and during solidification. 

Ingot dimensions ranged from 23 to 40 inches in length. Longer ingots were sawed in half, producing 

two billets for the extrusion step. After sampling, cropped ingots were transferred to the Special 

Products Plant for center drilling and surface machining. Before machining, as-cast ingots weighed 

up to 653 kg U. Machined billets weighed as much as 520 kg, and were heat-treated in Plant 6, the 

Metals Fabrication Plant, before being moved to another site.

2.2.5.2 Radiation Sources

Radioactive sources in Plant 5 consisted of “uranium” and “recycle uranium” of UF4 granules, uranium 

metal derbies, billets, and ingots, and uranium oxide in MgF2. 

Results of gamma fluxes recorded at various stations applicable to Plant 5 from the radiation field 

characterization and radiological equipment evaluation efforts conducted at the FEMP are listed in 

Table 2-8.

Table 2-8. Plant 5 gamma fluxes. 

Spectrum 

number Location

Integrated gamma flux (photons/cm2 sec)

30-225 keV 675-1050 keV Ratio

358402 Graphite crucible (3898) at 30 cm 310 18 17.2

378411 Graphite crucible (G-8010)) at 30 cm 183 11 16.6

368407 Crucible Load Station (Scrap Can) 962 261 3.6

368408 Crucible Load Station (crucible) 453 76 5.9

368409 Crucible Load Station (can of briquettes) 638 163 3.9

368411 Crucible Load Station (55 gal. Drum) 990 348 2.8

368413 Crucible Burn Out Station 776 69 11.2

368414 NE Saw (Machined U ingot at 0.5 cm) 1032 358 3.6

368416 NE Saw (Black U ingot at 0.5 cm) 1130 290 3.9

378427 Box of black crop tops at 25 cm 848 154 5.5

378424 Black U ingot at 1 cm

Outdoors 989 163 6.1

378417 Background just outside Building 3045 35 5 7.0

378418 Near old thorium hopper 424 58 7.3

378420 Background 75 ft. from Building 3045 25 2 12.5

An exposure study of Plant 5 production personnel to airborne radioactive dust provided the data 

listed in Table 2-9a. Table 2-9b provides the estimated potential annual uranium intakes by Plant 5 

workers by the years. 
Effective Date: 05/20/2004 Revision No. 00 Document No. ORAUT-TKBS-0017-2 Page 30 of 83

Table 2-9a. Plant 5 airborne radioactivity concentration by operation or location (year data taken as 

indicated). 

Operation or location

Ave. Conc.

α dpm/m3

Conc. 

w/Respiratora 

Bq/m3

Plant 5 –A Area(1965 data)

1st floor administrative and office areas 10 1.67E-01

Furnace pot cooling tank area east side 6 1.00E-01

Rockwell area east of Rockwell 16 9.3 1.55E-01

Rockwell area west of Rockwell 29 27 4.50E-01

Approximately 7 ft. west of 35 Rockwell area (GA) 43.3 7.22E-01

Furnace pot lipping area 5 ft west of pots 45 7.50E-01

Furnace pot lipping area 5 ft east of pots 18.3 3.05E-01

Furnace pot cooling tank area west side 15.3 2.55E-01

West side smoking area 23 3.83E-01

Dumping UF4 and magnesium into No. 4 blender 74 2.47E-02

3rd floor storage area while dumping magnesium 8.77 1.46E-01

Jolters Area

Jolter operator breathing zone 163.3 5.44E-02

General area 103.7 1.73E+00

East jolters A, B, C (General Area) 257 4.28E+00

East jolters D, F (General Area) 23.3 3.88E-01

East jolters A, B, C (Breathing zone) 52.6 1.75E-02

West jolters (General Area) 93 1.55E+00

West jolters between J and K (GA) 17.7 2.95E-01

West jolters (BZ) 45.7 7.62E-01

East side of jolters D, E, F (GA) 87.2 1.45E+00

East side of jolters D, E, F (BZ) 191 6.37E-02

West side of jolters H, I, J, K (GA) 112.3 1.87E+00

West side of jolters H, I, J, K (BZ) 260 8.67E-02

Removing pot from jolter 716.7 2.39E-01

Dumping bad MgF2 liner west side north of jolters 316 1.05E-01

7 ft south of dumping station during dumping of broken liner 183.3 6.11E-02

7 ft north of dumping station during dumping of broken liner 273.3 9.11E-02

Pot charging

GA during pot charging at bottom F machine 102 1.70E+00

BZ during pot charging at bottom F machine 4 44.8 1.49E-02

Operator placing MgF2 liner into pot with scoop; taming with scoop; 

scraping pot with back of saw blade; vacuuming and removing pots 

(Breathing Zone) (1969 data) 91.5 3.05E-02

Capping Stations

BZ Capping pots at No. 3 station 32 1.07E-02

GA Capping pots at No. 3 station 10.8 1.80E-01

Capping pots at No. 1 station 65.7 2.19E-02

No. 1 capping station 33.5 5.58E-01

East Breakout Area

Smoking area 41.3 6.88E-01

Breakout platform 56.7 9.45E-01

Breakout operation at the platform (BZ) 89.25 2.98E-02

Delipping furnace pots and breaking crust 301.2 1.00E-01

Breakout area desk 40.3 6.72E-01

Cleaning derbies and breaking slag 280.9 9.36E-02

Weighing and marking derbies 68 2.27E-02
Effective Date: 05/20/2004 Revision No. 00 Document No. ORAUT-TKBS-0017-2 Page 31 of 83

Operation or location

Ave. Conc.

α dpm/m3

Conc. 

w/Respiratora 

Bq/m3

West Breakout station

Delidding furnace pots and breaking crust 346.7 1.16E-01

GA at breakout platform 47 7.83E-01

BZ- Breakout operation on the platform 67.3 2.24E-02

Cleaning derbies and breaking slag 30 1.00E-02

Weighing and marking derbies 17.3 5.77E-03

Breakout area 30.3 5.05E-01

Breakout smoking area 5 8.33E-02

Changing drums and lipping at the west slag drumming station 39 1.30E-02

5 ft southeast of drumming station 12.7 2.12E-01

Stenciling empty drums at drumming station (BZ) 35.7 1.19E-02

Stenciling empty drums at drumming station (GA) 153.7 2.56E+00

Hoffman Canning Station

Checking Hoffmans and shaking bags 2,033.3 6.78E-01

Changing drums 4,983 1.66E+00

Emptying collectors 638.4 2.13E-01

5 ft north of Hoffman 248, 253, 241 131 4.37E-02

5 ft south of Hoffman 248, 253, 247 32.8 1.09E-02

3 ft south of Hoffman canning station 261 1,280 4.27E-01

3 ft north of Hoffman canning station 261 1,787 5.96E-01

Residue Area

Weighing and marking drums of residue at scale 30 5.00E-01

During operation at the residue scale 22 3.67E-01

Plant 5 B Area

Administrative and office areas 19.8 3.30E-01

Graphite shop 5.3 8.83E-02

East remelt furnaces area north end 71.3 1.19E+00

East remelt furnaces area south end 61 1.02E+00

West furnace area north end 31 5.17E-01

West furnace area south end 53 8.83E-01

Midway between east and west charge station 72.7 1.21E+00

Scrap storage area 2nd floor north end 43.7 7.28E-01

Ingot storage area 1st floor north 56 9.33E-01

Ingot storage area smoking area 60.3 1.01E+00

Shipping area 1st floor north 51 8.50E-01

Charging Station

Station desk at east charging station 117.3 1.96E+00

Operator charging material at east charging station 150 5.00E-02

Station desk at west charging station 58.3 9.72E-01

Operator charging material at west charging station 46.8 1.56E-02

Lathe Area

Putting cut in lathe cutting cup and removing cup from lathe 72.5 2.42E-02

4 ft north of lathe during operation 31 5.17E-01

Operator putting contaminated mold in lathe 178 5.93E-02

Working controls and operating contaminated lathe 75 2.50E-02

Removing contaminated mold from lathe 51.3 1.71E-02

4 ft southwest of lathe 16.7 2.78E-01

East Bottom Remelt Furnaces

Removing cup and uncovering furnace, placing cup and mold inside 

furnace 10.6 3.56E-03

Lowering mold tank from furnace, and transporting to cooling tunnel 396.8 1.32E-01
Effective Date: 05/20/2004 Revision No. 00 Document No. ORAUT-TKBS-0017-2 Page 32 of 83

Operation or location

Ave. Conc.

α dpm/m3

Conc. 

w/Respiratora 

Bq/m3

Raising and lowering ram inside of buggy 101.3 3.38E-02

Raising and lowering ram without use of buggy 331 1.10E-01

Placing mold in open furnace 127 4.23E-02

Attaching mold tank to bottom furnace 386.3 1.29E-01

East furnace area 38 6.33E-01

Smoking area 69.3 1.16E+00

North of furnace 14 123.3 2.06E+00

West Bottom Remelt Furnaces

Removing cup and vacuuming furnace 1,392.5 4.64E-01

Placing mold in mold tank 435 1.45E-01

West furnace area 35.7 5.95E-01

Lowering mold tank from furnace, and transporting to cooling tunnel 121.3 4.04E-02

Raising and lowering ram inside buggy 123 4.10E-02

Placing mold in open furnace 196.7 6.56E-02

Smoking area 42 7.00E-01

North of 28 furnace north end 220 3.67E+00

South of 15 furnace north end 41.3 6.88E-01

Separation Booths

Vacuuming top of ingot and separating ingot and mold at west booth 43.7 1.46E-02

West separation booth general area 46.7 7.78E-01

Vacuuming top of ingot and separating ingot and mold at east booth 74.7 2.49E-02

East separation booth general area 38.8 6.47E-01

East Saw Area

East saw general area 68.3 1.14E+00

During operation of east circular saw 32 5.33E-01

Operating east circular saw 58 1.93E-02

Setting up ingot 813.3 2.71E-01

Removing ingot from saw weighing and stamping 480 1.60E-01

Removing sample from saw and stamping 693.3 2.31E-01

Cropping 10-in depleted ingots(1976 data)

Millipore filter sampling (BZ) 27.6 4.60E-01

East Saw operator’s desk (GA) 59.3 9.88E-01

Column southwest of saw 225 3.75E+00

Electric box northwest of saw 16 2.67E-01

10 ft south of saw blade at conveyor 77 1.28E+00

Mold reconditioning area 56.7 9.45E-01

Cleaning and coating molds inside enclosure 55.7 9.28E-01

Coating cups 101 1.68E+00

West Saw Area

West saw general area 57 9.50E-01

During operation of east circular saw 32 5.33E-01

Operating east circular saw 58 9.67E-01

Setting up ingot 235.3 7.84E-02

Removing ingot from saw weighing and stamping 292 9.73E-02

Removing sample from saw and stamping 354.3 1.18E-01

North Saw Area

Operating saw sharpener 36.3 6.05E-01

Saw sharpening area 12.7 2.12E-01

Operating north circular saw (BZ) 54.8 9.13E-01

During saw operation 12.7 2.12E-01
Effective Date: 05/20/2004 Revision No. 00 Document No. ORAUT-TKBS-0017-2 Page 33 of 83

Operation or location

Ave. Conc.

α dpm/m3

Conc. 

w/Respiratora 

Bq/m3

Crucible Assembling Area

Knocking out plug, putting new plug in crucible and painting inside of 

crucible 110.3 3.68E-02

5 ft south of crucible assembling station 143 4.77E-02

5 ft north of crucible assembling station 154.5 5.15E-02

Working controls at west burnout station 2nd floor 67 1.12E+00

Working controls at east burnout station 2nd floor 44 7.33E-01

Graphite crusher

Breaking and charging graphite to crusher 2nd floor 823.3 2.74E-01

GA during charging crusher operation 1,533 5.11E-01

Changing and weighing drums at drumming station 1st floor 3,233 1.08E+00

7 ft northwest of crusher while charging 147.8 4.93E-02

East Burnout Canning Station

Changing drum and lipping 613.3 2.04E-01

Cleaning dirty pot lips at east pot lip cleaning station 33.8 1.13E-02

Filter Press

5 ft north of press 19.3 6.43E-03

5 ft south of press 529.4 1.76E-01

Operator cleaning filter press 216 7.20E-02 a

 A Decontamination Factor (DF) of 50 for wearing respirator was applied to Breathing Zone (BZ) concentrations for 

selected process activities.

Table 2-9b. Projected annual uranium intake estimates by Plant 5 workers. 

Years

Adjustment Factora

Average 

Uranium Conc.

Bq/m3

Average Annual 

Uranium Intakes

Bq/yr b 1965 Data 1976 Data

1953-1954 0.5 2.0 4.09E-01 9.82E+02

1955-1964 1.1 4.1 8.65E-01 2.08E+03

1965-1968 1.0 3.7 7.83E-01 1.88E+03

1969-1971 1.1 4.1 8.65E-01 2.08E+03

1972-1975 0.67 2.4 6.64E-01 1.59E+03

1976-1981 0.15 1.0 1.67E-01 4.02E+02

1982-1984 0.28 3.0 4.42E-01 1.06E+03

1985-1989 0.82 1.0 4.03E-01 9.68E+02 a

 Adjustment factors: The measured airborne radioactivity concentrations are adjusted for other 

years than the year in which the measurements are made, in accordance with the percentage of 

process system throughput capacity in one shift. In absence of specific throughput information, 

Appendix 2A is used as the default model for FEMP production rates for the derivation of the 

adjustment factors. b

 Radionuclide (e.g., Pu, Np, Tc and U isotopes) radioactivity fractions for recycled uranium is 

given in Table 2-54. 

2.2.6

The ground floor area of the Metals Fabrication Plant was 206,270 sq. ft. At this plant, uranium metal 

products were heat-treated to improve their strength and grain structure. Some of these products 

were shipped off the site for extrusion. The extruded tubes were returned to the Metals Fabrication 

Plant to be machined into final products. After being inspected to ensure the highest quality, the 

products were shipped to other DOE sites. The principal capabilities of the Metals Fabrication Plant 

were:

Plant 6 – Metals Fabrication Plant
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• Heat treating machined ingots for extrusion using neutral salt (NuSal) furnace and salt-oil 

baths.

• Heat-treating and final machining of target elements or fuel cores.

• Cropping, surface milling, and inspecting products for shipment.

• Metal pickling and chip briquetting.

• Providing standby capabilities for rolling as-cast ingots into rods that had close dimensional 

tolerances.

• Inspecting finished target element cores and final products before shipment.

2.2.6.1 Processes and Activities

All center drilled and machined ingots and billets received from the Special Products Plant were heattreated in a molten salt bath (50% NaCl, 50% KCl) at 1,350o F and water-quenched to achieve the 

beta phase grain structure for improving extrusion and final machining yields. Heat-treated ingots and 

billets were extruded into tubular forms having various inside and outside diameter dimensions. 

Extruded billets for N-Reactor were fabricated at Hanford, while extruded tubes produced for the 

Savannah River Site were returned to the FEMP for target element core fabrication.

Extruded tubes were cut into 8-inch lengths using a cut-off lathe. After salt heating in the beta range 

and oil quenching, final machining was performed automatically on the Cross Transformatic machine. 

The machined elements were individually stamped for identification and processed through a 

degreasing step. Machining chips were recrushed, pickled, rinsed, dried, briquetted, and recycled to 

casting operations. NOx destruction equipment was installed on the pickling system exhaust. 

Cleaned finished pieces were conveyed to the Final Inspection Area for testing by instrumental 

nondestructive techniques.

Depleted uranium nonbriquettable turnings were oxidized in a Recovery Plant furnace; nonremeltable 

metal spills and secondary top crops were stored on an outdoor concrete pad for potential sale to the 

commercial nuclear industry. Enriched uranium nonbriquettable turnings were sent to the Recovery 

Plant for oxidation; the nonremeltable metal spills and top crops are recovered by dissolution in a 

refinery metal dissolver tank, using nitric acid.

The standby rolling mill facility consisted of an ingot preheat furnace, blooming mill, crop shear, 

equalizing furnace, flying shear and related conveying, and cooling and rod straightening equipment. 

The rolling capacity for oval billets and round rod was approximately 36 metric tons/day based on 

one-shift operation.

2.2.6.2 Radiation Sources 

Radioactive sources included uranium and recycle uranium metal.

Results of the gamma fluxes recorded at various stations applicable to Plant 6 from the radiation field 

characterization and radiological equipment evaluation at the FEMP are listed in Table 2-10. 
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Table 2-10. Plant 6 gamma fluxes. 

Spectrum 

number Location

Integrated gamma flux (photons/cm sec)

30-225 keV 675-1050 keV Ratio

388403 Tube cutting work station 253 58 4.4

388406 Tube packing work station 155 34 4.6

388407 Box of packed tubes at 1 cm 1,273 348 3.7

The results of airborne radioactivity surveillance data for Plant 6 are listed in Table 2-11a. Table 2-

11b provides the estimated potential annual uranium intakes by Plant 6 workers by the years. 

Table 2-11a. Plant 6 airborne radioactivity concentration by operation or location.(year 

data taken as indicated)

Operation or Location

Ave. Conc.

α dpm/m3

Conc. 

w/Respiratora 

Bq/m3

Beta slug furnace and rolling mill (1959 data)

Sample results obtained before modification of 

furnace

Taken 2 ft from furnace 99 1.65E+00

Taken 4 ft from furnace 276 4.60E+00

Rod stamping area about 25 ft south of furnace 375 6.25E+00

Aisle 15 ft west of furnace. 158 2.63E+00

Sample results obtained after modification of furnace

Taken 2 ft from west side of furnace 50 8.33E-01

South end of furnace, ≈ 18 in. from salt bath. 13 2.17E-01

North end of furnace, ≈ 18 in. from salt bath. 33 5.50E-01

Rod stamping area about 25 ft south of furnace 8 1.33E-01

Top of electrical bay, 4 ft east of cooling bed 24 4.00E-01

West side of furnace 38 6.33E-01

Slug loading and unloading area 27 4.50E-01

East side of furnace where operator works crane 30 5.00E-01

Air dust evaluation of NuSal Treatment (1958 data)

Pump hanging on crane hooks 3-4 ft directly above 

furnace, slugs in furnace 56 9.33E-01

Adding slugs to furnace 276 4.60E+00

Pump 12 ft west of furnace. Slugs being treated. 

Visible smoke. Cause eye burning 297 4.95E+00

Pump in smoking area situated 15-20 ft east of 

furnace. Slugs being treated. 104 1.73E+00

Pump 15 ft east of crop shear pulpit in aisle at face 

level. Slugs being treated 87 1.45E+00

Pump 15 ft east of blooming mill pulpit in aisle at 

face level. Slugs being treated 60 1.00E+00

Air dust evaluation of turret lathes (1956 data)

No. 1 turret lathe operator 43 7.17E-01

No. 2 turret lathe operator 150 2.50E+00

No. 3 turret lathe operator 417 6.95E+00

No. 4 turret lathe operator 199 3.32E+00

Lathe area 313 5.22E+00 a

 A Decontamination Factor (DF) of 50 for wearing respirator was applied to Breathing Zone (BZ) 

concentrations for selected process activities.
Effective Date: 05/20/2004 Revision No. 00 Document No. ORAUT-TKBS-0017-2 Page 36 of 83

Table 2-11b. Projected annual estimated uranium intake by Plant 6 workers. 

Years

Adjustment 

Factora

1956, 1958,1959 

Data

Average 

Uranium Conc.

Bq/m3

Average Annual 

Uranium Intakes b

Bq/yr

1953-1954 0.5 1.01E+00 2.41E+03

1955-1964 1.0 2.01E+00 4.82E+03

1965-1968 0.9 1.81E+00 4.34E+03

1969-1971 0.6 1.21E+00 2.89E+03

1972-1975 0.6 1.21E+00 2.89E+03

1976-1981 0.25 5.03E-01 1.21E+03

1982-1984 0.74 1.49E+00 3.57E+03

1985-1988 0.2 4.02E-01 9.65E+02 a

 The measured airborne radioactivity concentrations are adjusted for other years 

than the year in which the measurements are made, in accordance with the

percentage of process system throughput capacity in one shift. In absence of 

specific throughput information, Appendix 2A is used as the default model for 

FEMP production rates for the derivation of the adjustment factors. b

 Radionuclide (e.g., Pu, Np, Tc and U isotopes) radioactivity fractions for recycled 

uranium is given in Table 2-54. 

2.2.7

Plant 7 operations began in June 1954. The plant converted uranium hexafluoride to green salt using 

the same process as the Pilot Plant. The green salt was used in Plant 5 to produce uranium metal. 

After only 2 years of operation, the U.S. Atomic Energy Commission (AEC) directed the shutdown of 

Plant 7 because a similar processing plant was operating in Paducah, Kentucky. For the next 13 

years, site management considered several proposals for the idle facility, but none were accepted. In 

1969, all equipment was excessed and sold, and the building was used to store drums of green salt 

and empty containers.

Plant 7 – Hexafluoride Reduction Plant

2.2.7.1 Processes and Activities

Plant 7 was designed for the conversion of UF6 to UF4 in a gas-gas reaction. UF6 gas vaporized from 

storage cylinders was mixed with hydrogen gas produced by ammonia dissociation in a specially 

designed nozzle inside a tapered vertical reaction tower maintained at approximately 100°F with 100 

kW heaters and forced-air cooling.

Most of the solid UF4 formed in the reaction settled downward in the tower as a fine powder, and was 

collected in a surge hopper. Some of the product was carried out of the reactor with byproduct HF 

and excess hydrogen and collected. The product UF4 was conveyed through a micropulverizer, 

batch-blended, packaged into 10-gallon cans, and weighed.

The filtered gases, HF, H2, and N2, leaving the solid collection system were passed through a Monel 

wool filter, cooled to 150°F, passed through a CaF2 chemical trap, and refrigerated to minus 90°F. 

The condensed anhydrous liquid HF was sent to the Tank Farm for in-plant storage prior to use or 

sale. The uncondensed HF was absorbed and neutralized in a lime slurry system. Hydrogen leaving 

the scrubbers was burned and the combustion gases vented to the atmosphere.

The design capacity of this plant was 12 tons per day of uranium as uranium tetrafluoride, split evenly 

between depleted and normal uranium from two sets of four reactors each. One on-stream spare 

reactor per set was included.
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At the completion of plant construction, the capability was determined to be a UF6 feed rate of 300 

pounds per hour. Based on a 75% on-stream factor for the eight-reactor capacity, the UF4 production 

rate was calculated at 14.6 tons per day. By December 1955, the UF6 feed rate was at 550 pounds 

per hour and the UF4 rate was at 26.76 tons per day.

2.2.7.2 Radiation Sources

The sources of radioactivity during the limited years of operation for Plant 7 were UF6 and UF4. 

An exposure study of Plant 7 production personnel to airborne radioactive dust provided the data 

listed in Table 2-12. The last two columns of Table 2-12 provide the estimated potential annual 

uranium intakes by Plant 7 worker groups by the years.

Table 2-12. Plant 7 airborne radioactivity concentration and uranium intakes by worker 

groups. 

Job description

Ave. 1955 and 1956 Data

Annual Uranium

Intakes d

Bq/yr

× MACa

Conc.

α dpm/m3

Conc. 

Bq/m3 1954b 1955-1956c

Packaging helpers 1.56 109.2 1.82E+00 2.18E+03 4.37E+03

Packing laborers 2.20 154 2.57E+00 3.08E+03 6.16E+03

Transfer operators 3.56 248.85 4.15E+00 4.98E+03 9.95E+03

Packaging operators 0.69 48.3 8.05E-01 9.66E+02 1.93E+03

General foreman 0.45 31.15 5.19E-01 6.23E+02 1.25E+03

Reactor operators & helpers 0.36 25.2 4.20E-01 5.04E+02 1.01E+03

Vaporizer operators 0.23 15.75 2.63E-01 3.15E+02 6.30E+02

Laborers, general 0.31 21.7 3.62E-01 4.34E+02 8.68E+02

Shift foreman & lead man 0.26 18.2 3.03E-01 3.64E+02 7.28E+02

Forklift operators 0.26 18.2 3.03E-01 3.64E+02 7.28E+02

Control room operators 0.19 13.3 2.22E-01 2.66E+02 5.32E+02

Utility helpers 0.24 16.8 2.80E-01 3.36E+02 6.72E+02

Area foreman 0.19 13.3 2.22E-01 2.66E+02 5.32E+02

Superintendent 0.17 11.9 1.98E-01 2.38E+02 4.76E+02

Lime slurry operators 0.18 12.25 2.04E-01 2.45E+02 4.90E+02

Office personnel 0.10 6.65 1.11E-01 1.33E+02 2.66E+02

Chemists 0.09 109.2 1.82E+00 2.18E+03 4.37E+03 a

 Maximum Allowable Concentration (MAC) - 70 α dpm/m3. b Plant 7 operated 6 months in 1954. c

 Maximum capacity assumed for these years. d

 Radionuclide (e.g., Pu, Np, Tc and U isotopes) radioactivity fractions for recycled uranium is given in 

Table 2-54. 

The results of airborne radioactive dust surveillance data for Plant 7 by operation or location are listed 

in Table 2-13a. Table 2-13b provides the estimated potential annual uranium intakes by Plant 7 

worker by the years.
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Table 2-13a. Plant 7 airborne radioactivity concentration by operation or location. 

Operation or location

Type of 

sample

Ave.1955-1956 Data

× MACa Conc.α dpm/m3

Conc. 

Bq/m3

Emptying dust collectors Breathing zone 25.55 1.79E+03 2.98E+01

Adjusting weights and lidding Breathing zone 3.05 2.14E+02 3.56E+00

Cleaning cans after packaging Breathing zone 3.90 2.73E+02 4.55E+00

Packaging green salt Breathing zone 1.15 8.05E+01 1.34E+00

Changing cylinders of UF6 General air 0.70 4.90E+01 8.17E-01

First-floor area General air 0.54 3.75E+01 6.24E-01

Lime slurry area, 1st floor General air 0.29 2.00E+01 3.33E-01

Smoking area, 1st floor General air 0.39 2.73E+01 4.55E-01

Second-floor area General air 0.09 5.95E+00 9.92E-02

Washroom, 2nd floor General air 0.09 6.30E+00 1.05E-01

Lime slurry area, 2nd floor General air 0.09 6.30E+00 1.05E-01

Third-floor area General air 0.41 2.87E+01 4.78E-01

Fourth-floor area General air 0.23 1.61E+01 2.68E-01

Fifth-floor area General air 0.29 2.00E+01 3.33E-01

Smoking area, 5th floor General air 0.14 9.80E+00 1.63E-01

Washroom, 5th floor General air 0.13 9.10E+00 1.52E-01

Sixth-floor area General air 0.27 1.86E+01 3.09E-01

Seventh floor area General air 0.75 5.25E+01 8.75E-01

Vaporizer area General air 0.21 1.44E+01 2.39E-01

Control room General air 0.20 1.40E+01 2.33E-01

Area foreman’s office General air 0.07 4.55E+00 7.58E-02

Superintendent’s office General air 0.08 5.25E+00 8.75E-02

Laboratory General air 0.08 5.60E+00 9.33E-02

STORAGE WAREHOUSE General air 0.46 3.22E+01 5.37E-01

Storage tank area General air 0.04 2.80E+00 4.67E-02

Average over Plant 7 areas and 

operations 1.6 1.1E+02 1.8E+00 a

 Maximum Allowable Concentration (MAC) - 70 α d/m per meter3. 

Table 2-13b. Projected annual estimated uranium intake 

by Plant 7 workers. 

Years

Adjustment 

Factora

1955-1956 Data

Average Annual 

Uranium Intakes

Bq/yr

1954 1.0 2.16E+03b

1955-1956 1.0 4.32E+03 a Maximum capacity assumed for these years. b Plant 7 operated 6 months in 1954.

2.2.8

Plant 8 construction was complete in 1953 and production began in November of that year. The 

ground floor area was 25,500 sq. ft. In the Scrap Recovery Plant, recycled residues and scrap from 

uranium processing were upgraded before being sent to the refinery for uranium extraction. In 

addition to recycle functions, the plant equipment was used to treat enriched uranium residues. The 

primary functions in the Scrap Recovery Plant included:

Plant 8 – Scrap Recovery Plant
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• Furnacing various enriched residues from onsite generation and offsite receipt to prepare 

refinery feed by removing moisture, oil, graphite and metallic impurities;

• Filtering large volumes of low-level radioactive waste slurries using rotary vacuum precoat 

filters;

• Drying and heating in a furnace waste materials for offsite disposal, including contaminated 

filter cakes from neutralized process waste streams and Refinery slag leach operations; and

• Drum washing operations.

2.2.8.1 Processes and Activities

The Scrap Recovery Plant process primarily involved upgrading enriched uranium recycle materials 

from FEMP and offsite operations to produce feed materials for head-end processing in the refinery. 

Recycle materials containing metallic uranium or magnesium, oil, graphite, or other troublesome 

agents were roasted to oxidize the components. Oversize pieces were screened and milled. Wet 

materials were dried; several furnaces were used for these purposes: a rotary kiln, multiple-hearth 

vertical furnaces, and small single-hearth furnaces.

All roasted materials were used as feed in the refinery. Recycle materials of significantly different 

isotopic assays were segregated throughout processing and in storage. The size of production 

equipment limited processing of enriched materials to 1.25% 235U. Filtration equipment was used for 

reclaiming low-level solid waste for drying, packaging, and shipment off the site for burial.

Various thorium residues were processed through the Plant 8 oxalate system to thorium hydroxide to 

return them to the production stream. Residues were digested in hydrochloric acid and the solution 

was filtered on a rotary vacuum filter. The filtrate was pumped to a precipitation tank and mixed with 

oxalic acid. Thorium oxalate precipitated out of the solution. The thorium oxalate precipitate was 

filtered, slurried, and reverted with sodium hydroxide solution to thorium hydroxide. The thorium 

hydroxide was filtered on a rotary vacuum filter and the solution was recycled back to the precipitation 

tank. The filter cake was calcined in a multiple hearth furnace called the uranium ammonium 

phosphate furnace. The dried thorium hydroxide was digested in the Pilot Plant digestion system. 

Approximately 310 metric tons of thorium as thorium hydroxide were produced through the Plant 8 

oxalate system.

Thorium hydroxide was produced in Plant 8 for 6 months in 1966 in the reversion system. Thorium 

tetrafluoride was reverted to thorium hydroxide by heating it in a reverter tube with hydrofluoric acid. 

The offgas from the reverter was neutralized in a caustic scrubber. Approximately 59 metric tons of 

thorium in the form of thorium hydroxide were produced by this system.

2.2.8.2 Radiation Sources

Radioactive sources in Plant 8 included uranium, recycled uranium metal, and thorium compounds. 

Radioactive thorium compounds are contained in one silo and in two sections of a common bin 

structure at Plant 8. The silo contents are primarily thorium oxide, thorium hydroxide, and from 25% 

to 50% diatomaceous earth. Other thorium compounds in small quantities might be present. In 

addition, some residual uranium compounds are in the silo.
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Radiation measured at the surface of the thorium silo ranges from 3 to 40 mR/hr. The area above 

and immediately around the silo is an airborne radioactivity area. The radiation level of the silo 

gradually increases from ground level (3 mR/hr) to about 20 mR/hr at the bottom conical section. The 

radiation level gradually increases to about 37 mR/hr at the top of the thorium fill level in the silo, and 

then decreases to 7 mR/hr at the top of the silo. The maximum surface radiation level is 55 mR/hr at 

the cone/cylinder interface; however, radiation levels as high as 70 mR/hr have been observed.

A spot check of the thorium bin shows maximum surface radiation of 40 mR/hr. The surface gamma 

radiation measures 11 mR/hr at the bottom of the bin and 13 mR/hr at the top of the bin. The 

maximum surface radiation is 45 mR/hr at the bin/hopper junction.

Results of an exposure study of Plant 8 personnel to airborne radioactivity by operation or location 

during 1956 and 1957 are listed in Table 2-14a. Table 2-14b provides the estimated potential annual 

uranium intakes by Plant 8 worker by years. Table 2-14c provides the estimated thorium intakes.

Table 2-14a. Plant 8 average airborne radioactivity concentrations by operations or locations. 

Operation or location

Type 

sampleb

Ave. 1956 and 1957 Data

× MACa Conc.

α dpm/m3

Conc. 

w/Respiratorc 

Bq/m3

Outside

Outside Williams Mill GA 44.30 3101.00 1.03E+00

Breaking salt at outside mill BZ 30.80 2156.00 7.19E-01

Shoveling onto conveyor at outside mill BZ 137.80 9646.00 3.22E+00

Changing drums at outside mill BZ 122.90 8603.00 2.87E+00

C-liner flow

Juice hopper area GA 0.65 45.50 7.58E-01

Conveyor silo to calciner area GA 1.48 103.60 1.73E+00

Conveyor silo to bucket elevator GA 3.48 243.60 4.06E+00

Williams Mill, 2nd floor GA 0.65 45.50 7.58E-01

Calciner furnace, drum dumping area GA 2.12 148.05 2.47E+00

Rotex dumping station area GA 0.65 45.50 7.58E-01

Rotex drumming station area GA 0.74 51.80 8.63E-01

Dumping drum of C-liner BZ 8.21 574.35 1.91E-01

Feeding material to Rotex BZ 9.71 679.35 2.26E-01

Changing drums at Rotex drumming station BZ 8.02 561.40 1.87E-01

Checking Rotex drum for fullness BZ 4.57 319.90 1.07E-01

Box furnace

Box furnace area GA 1.56 109.20 1.82E+00

Feeding box furnace BZ 1.09 75.95 1.27E+00

Checking drum for fullness BZ 6.99 489.30 1.63E-01

Changing drums at box furnace BZ 4.95 346.50 1.16E-01

Oxidation furnace

Oxidation furnace area, 2nd floor GA 1.56 109.20 1.82E+00

Smoking area, 1st floor GA 1.48 103.60 1.73E+00

Spreading material on feed tray & removing foreign 

objects

BZ

1.77 123.90 4.13E-02

Feeding oxidation furnace BZ 2.05 143.15 4.77E-02

Checking drum for fullness BZ 2.85 199.50 6.65E-02

Changing drums at oxidation furnace BZ 4.45 311.15 1.04E-01

Muffle furnace

Muffle furnace area, 1st floor GA 1.03 72.10 1.20E+00

Muffle furnace area, 2nd floor GA 0.60 41.65 6.94E-01
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Operation or location

Type 

sampleb

Ave. 1956 and 1957 Data

× MACa Conc.

α dpm/m3

Conc. 

w/Respiratorc 

Bq/m3

Charging muffle furnace BZ 29.04 2032.80 6.78E-01

Hoeing material in muffle furnace BZ 2.98 208.25 6.94E-02

Checking drum for fullness BZ 32.05 2243.15 7.48E-01

Changing drums at canning station BZ 121.25 8487.50 2.83E+00

Graphite furnace

Graphite furnace area, 2nd floor GA 0.84 58.80 9.80E-01

Dumping feed material on tray BZ 4.60 322.00 1.07E-01

Charging (chips and turnings) BZ 7.20 504.00 1.68E-01

Pushing chips into furnace, south side BZ 1.08 75.60 2.52E-02

Hoeing material in furnace BZ 0.92 64.05 2.14E-02

Checking drums for fullness BZ 1.36 95.20 3.17E-02

Changing drums BZ 24.44 1710.45 5.70E-01

Rotary kiln

Rotary kiln area, 1st floor GA 2.80 196.00 3.27E+00

Smoking area by panel board for kiln GA 0.67 46.90 7.82E-01

Feed tray area GA 0.18 12.60 4.20E-03

Feeding kiln BZ 1.20 83.65 2.79E-02

Checking drum for fullness at “scalping” canning station BZ 5.83 407.75 1.36E-01

Changing drum at kiln scalping station BZ 24.05 1683.50 5.61E-01

Checking drum for fullness at product canning station BZ 9.90 693.00 2.31E-01

Changing drums at product canning station BZ 18.01 1260.70 4.20E-01

Dumping through kiln drum dumper BZ 4.59 321.30 1.07E-01

Green salt reverter

Dumping area GA 1.83 127.75 2.13E+00

Mezzanine area GA 1.98 138.25 2.30E+00

Canning station area GA 2.17 151.90 2.53E+00

Feeding at reverter dumping station BZ 9.04 632.80 2.11E-01

Checking drum for fullness at canning station BZ 32.51 2275.70 7.59E-01

Sampling oxide at canning station BZ 12.43 870.10 2.90E-01

Changing drums at reverter canning station BZ 155.37 10875.90 3.63E+00

Wet area, 1st floor

Dump trailer area GA 0.20 14.00 2.33E-01

Foreman’s office GA 0.14 9.45 1.58E-01

Control room GA 0.12 8.40 1.40E-01

Washrooms GA 0.31 21.70 3.62E-01

Moyno pump area GA 0.71 49.70 8.28E-01

Denver crusher canning station area GA 0.17 11.90 1.98E-01

Digester area, 1st floor GA 0.32 22.40 3.73E-01

Sampling dump trailer BZ 0.33 23.10 3.85E-01

Wet area, 2nd floor

Surge and weigh hopper area GA 0.17 11.90 1.98E-01

Conveyor area, surge hopper to digest tanks GA 0.97 67.90 1.13E+00

Denver crusher area GA 0.34 23.80 3.97E-01

Oliver filter area GA 0.50 34.65 5.78E-01

Nimco filter area GA 0.17 11.55 1.93E-01

Smoking areas: 0.00 0.00E+00

By weigh hopper GA 1.29 89.95 1.50E+00

By Oliver filter GA 0.70 48.65 8.11E-01

In front of generator room GA 0.21 14.35 2.39E-01

Raking material on Oliver filter BZ 0.63 44.10 7.35E-01
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Operation or location

Type 

sampleb

Ave. 1956 and 1957 Data

× MACa Conc.

α dpm/m3

Conc. 

w/Respiratorc 

Bq/m3

Diuranate furnace 0.00 0.00E+00

Canning station area GA 0.42 29.40 4.90E-01

Smoking area panel board GA 0.07 4.90 8.17E-02

Second floor GA 1.35 94.50 1.58E+00

Pushing material off top hearth BZ 0.63 44.10 1.47E-02

Checking drum for fullness BZ 3.84 268.45 8.95E-02

Scoop sampling oxide at canning station BZ 21.44 1500.45 5.00E-01

Changing drum BZ 49.31 3451.70 1.15E+00

Plate-frame filter press

General area, 1st floor GA 0.30 21.00 3.50E-01

Scraping cake from filter leaves BZ 1.52 106.40 3.55E-02

Shoveling cake into drums BZ 1.26 87.85 2.93E-02

Reassembling press BZ 0.34 23.80 3.97E-01

Maintenance Area, 2nd floor GA 0.09 6.30 1.05E-01

Lime and Precoat Make-up Tank Area GA 0.11 7.70 1.28E-01

Centrifuge

General area, 1st floor GA 0.47 32.90 5.48E-01

Charging area, 2nd floor GA 1.83 128.10 2.14E+00

Charging with contaminated oil BZ 5.81 406.70 1.36E-01

Operating controls, 1st floor BZ

Leaching tank

General area, 1st floor GA 0.17 11.90 1.98E-01

General area, 2nd floor GA 0.10 7.00 1.17E-01

Charging leach tank BZ 0.31 21.70 3.62E-01

Average over Plant 8 areas and operations 11.23 674.20 7.56E-01 a

 Maximum Allowable Concentration (MAC) - 70 α d/m per meter3 

b

 GA = General Area BZ = Breathing Zone c

 A Decontamination Factor (DF) of 50 for wearing respirator was applied to Breathing Zone (BZ) concentrations for 

selected process activities.

Air samples from the top of the thorium silo indicate thoron daughter concentrations as high as 71 

WLs (WL defined as 1.3 x 105 MeV of potential alpha radiation per liter of air from radon daughter 

products without regard to the radon and daughter equilibrium). Air samples taken at the top of the 

silo indicate a thoron daughter (220Rn) concentration of 71 times the Acceptable Working 

Concentration or 71 WL. The area above and immediately around the thorium silo is an airborne 

radioactivity area.

Air samples at the surface of the thorium bin indicated a thoron daughter concentration of 0.44 WL. 

The airborne concentration of thoron daughters in air samples taken from the top of the bin is 0.44 

WL. Concentrations above 0.3 WL cause the area to be declared an airborne radiation area.

Table 2-14b provides a summary of the projected annual uranium intake that was estimated based on 

exposures to Plant 8 workers.
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Table 2-14b. Projected annual estimated uranium intake 

by Plant 8 workers. 

Years

Adjustment 

Factor

a

1956 and 1957

Data

Average Annual 

Uranium Intakes b

Bq/yr

1953-1954 0.7 1.27E+03

1955-1964 1.0 1.81E+03

1965-1968 0.8 1.45E+03

1969-1971 1.0 1.81E+03

1972-1975 0.6 1.09E+03

1976-1981 0.25 4.54E+02

1982-1984 0.74 1.34E+03 a

 The measured airborne radioactivity concentrations are adjusted for 

other years than the year in which the measurements are made, in 

accordance with the percentage of process system throughput capacity 

in one shift. In absence of specific throughput information, Appendix 

2A is used as the default model for FEMP production rates for the 

derivation of the adjustment factors.

b

 Radionuclide (e.g., Pu, Np, Tc and U isotopes) radioactivity fractions 

for recycled uranium is given in Table 2-54. 

Table 2-14c. Projected annual estimated thorium intake by Plant 8 workers. 

Years Adjustment Factora

1956 and 1957 U Data

Average 

Thorium Conc.

Bq/m3

Average Annual 

Thorium Intakes

Bq/yr

1966 3.39E-03 2.56E-03 6.14E+00

1969 1.32E-02 9.97E-03 2.39E+01

1970 2.54E-02 1.92E-02 4.62E+01

1971 3.05E-02 2.31E-02 5.55E+01 a

 The estimated uranium concentrations are adjusted to obtain thorium concentrations in 

accordance with the ratio of thorium/uranium emission estimates. Emission estimates are in 

Appendix 2B.

2.2.9

Construction of the Special Products Plant was complete in 1954; the facility had a ground floor area 

of 48,500 sq. ft. This facility was originally designed and constructed as a thorium metal production 

plant. The two basic processes, hydrofluoric acid precipitation of thorium fluoride and induction 

dezincing and melting, which were used to start the plant, were not able to produce a pure metal. 

However, improvement in production techniques permitted the eventual development of an oxalate 

precipitation process capable of producing pure thorium metal. The process began with the 

dissolution of solid thorium nitrate tetrahydrate in nitric acid. Hydrofluoric acid was added to the 

solution to precipitate a wet thorium tetrafluoride (ThF4). The ThF4 was dried in a predryer and a 

retort dryer and then was pulverized. The pulverized ThF4 was blended with calcium metal and zinc 

chloride and placed in a refractory lined furnace pot. Thorium tetrafluoride was coreduced by the 

calcium to form a zinc-thorium derby, from which the zinc was removed and remelted in a vacuum 

furnace. The resulting 500-kilogram thorium metal ingot was machined to produce the final thorium 

metal product. An estimated 380 metric tons of thorium metal were produced in Plant 9 from 1954 

through 1955. 

Plant 9 – Special Products Plant
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2.2.9.1 Processes and Activities

Because thorium processing was infrequent and in small amounts, the primary function of the Special 

Products Plant became casting larger ingots than those produced in Plant 5 and machining uranium 

metal pieces for extrusion. The principal capabilities of the Special Products Plant were:

• Casting enriched uranium derbies and recycle metal scrap into large-diameter ingots for NReactor

• Machining as-cast ingots and billets for extrusion

• Cleaning depleted uranium derbies for offsite shipment

• Decladding unirradiated copper-zirconium fuel cores

Operations in the Special Products Plant primarily involved machining uranium metal pieces and 

casting enriched uranium ingots up to 13 inches in diameter and 25 inches in length, and weighing up 

to 900 kg. Shrinkage cavities and impurities were removed by cropping the top section, as was done 

with smaller ingots. Cropped ingots from both plants were center-drilled on a unique LeBlondCarlstadt Rapid Boring machine. Surface machining was performed on standard lathes.

The ingot bore process (12.8 min. per ingot) involved the following operations: 

1. Operator presses BORE button to lower loader, then COOLANT START button. Check for 

good seal, then presses SPINDLE ON, checks RPM, and starts bore by pressing FEED 

forward button (all < 1 minute). During bore, operator continually monitors gauges for tool 

breakage and coolant pressure levels, and performs the following tasks:

2. Move the bored, inspected ingot from “inspection” stand to skid (45 sec).

3. Move a new ingot to “preparation” stand and knocks off “fins,” which might prevent proper 

chucking (1 min).

4. Replace chip drum when full (after every two ingots, 1.5 min).

5. Measure and mark lengths, removes isotopic labels, and record identity numbers of skid of 

new ingots (3 min/4 ingots).

The ingot exchange process (5.7 min. per ingot) involved the following operations:

1. By observing pressure dials, operator determines that boring is completed.

2. STOP the machine, RETRACT tool, press LOAD button, and slide back the splash guard. 

Unclamp and unchuck ingot, check core clearance, move hoist in, and clamp and lift ingot.

3. Use a rod to push a rag through the bore to wipe out oil and move ingot to inspection stand. 

Inspect tool and remove core to place in storage drum.

4. A new ingot is hoisted from “preparation” stand and loaded into position. Operator clamps in, 

tightens chuck, and closes splash guard.
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The NPR Saw operation (performed only on the second shift) consisted of five steps, namely 

positioning, sawing, stamping, weighing, and crating or storing of ingot material. After the ingots were 

separated within the separation booth, they were positioned manually on a conveyor extending from 

the separation booth to the saw. The ingots were pushed along the conveyor to a point near the saw 

before they could be automatically positioned for sawing off the “crop” end. After the “crop” end was 

sawed, at least two additional cuts, and sometimes three, were made on each ingot. The total time 

for each ingot in the saw operation was approximately 1 hour. The “crop” end and other sample cuts 

were placed manually in a wooden box. The ingots were then hoisted onto a skid for storage nearby.

All ingots cast in the Special Products Plant were for supplying the enriched N-Reactor fuel core 

requirement for the reactors at the Hanford Site. Machined ingots weighed as much as 520 kg and 

underwent heat-treating in the Metals Fabrication Plant.

A chemical decladding process, called Zirnlo, was performed in the Special Products Plant. Reject 

coextrusion sections from the cladding operation at Hanford were immersed in dilute nitric acid to 

remove the outer copper layer that served as a lubricant during coextrusion of Zircaloy-2 and uranium. 

The decoppered coextrusion sections were treated with dilute HF to remove the Zircaloy-2 jacket that 

encased the uranium metal core. The totally declad bare uranium metal cores were recycled to the 

casting operation for remelting to ingot form.

2.2.9.2 Radiation Sources

Radioactive sources in the Special Products Plant were recycled uranium and uranium metal, and 

thorium metal and compounds.

Results of the gamma fluxes recorded at various stations applicable to Plant 9 from the radiation field 

characterization and radiological equipment evaluation at the FEMP are listed in Table 2-15. 

Table 2-15. Plant 9 gamma fluxes. 

Spectrum 

number Location

Integrated gamma flux (photons/cm2 sec)

30-225 keV 675-1050 keV Ratio

388411 Just inside south entrance 253 43 5.9

388412 Lathe work station northeast corner 424 76 5.6

388415 West wing, southwest hot area 708 72 9.8

388416 Just outside south entrance by exhaust fan 776 165 4.7

388402 HP change room 5 <0.4 12.5

388419 HP change room 5 <0.5 10.0

LeBlonde Rapid Bore Operation Doses

An investigation was performed early in 1981 to determine elements of the Plant 9 LeBlonde Rapid 

Bore operation that contributed to operator external (beta-gamma) radiation exposure and to make 

quantitative estimates of these elements to recommend means for exposure reduction under ALAP 

goals. The results (skin doses) are listed in Table 2-16. 

Table 2-16. Plant 9 LeBlonde Rapid Bore skin doses. 

Operation

Average dose rate 

(mR/hr)

Skin dose (mR)

Per ingot Per shift (22 ingot) Per month (21 shifts)

Bore 2.23 2.23 49.1 1,030

Exchange 14.6 1.39 30.6 642

Total 11.7 3.62 79.7 1,672
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NPR Saw Operation Doses

An external radiation survey for the NPR Saw Station in Plant 9 had resulted in the following 

measurements throughout the NPR saw area:

Contact – 200 mrad/hr β 

Conveyor from Separation Booth to Saw (4 NPR 11-in. ingots)

Side of ingots – 1,000 – 2,000 mrad/hr β, 5 mR/hr γ 

Crop end of ingots – 20 – 30 mR/hr γ

At 1 ft. away – 100 – 300 mrad/hr β, 3.0 –5.0 mR/hr γ

Behind shield – 0.5 mrads/hr β and 0.5 mR/hr γ

Saw Station (11-in. NPR ingot within Saw; metal shield approximately 4 ft high and 2 ft was directly 

in front of saw)

Above shield (working distance – chest level) – 25 mrads/hr β and 0.8 mR/hr γ

Saw area – 5-20 mrads/hr β; 2.0 – 3.0 mR/hr γ

General Saw and Storage Area 

Storage area – 20-30 mrads/hr β; 3.0-4.0 mR/hr γ

The following table provides the resulting radiation exposure data for several work assignments 

performed by the operator:

Sawing extrusions made from 11-inch ingots 11 mR (3 days avg.)

Sawing 11 inch NPR ingots 33.4 mR (5 days avg.)

Crushing – briq. area 10 mR

Pickling - briq. area 5 mR

Total 59.4 mR γ and 

259 mrads β (24 days avg.)

Dose Rates at Plant 9 Locations

Surveys of external radiation levels at the Plant 9 remelt, ingot separation, charging and burnout 

areas were conducted in February 1963. Results are as follows:

General work areas - 10 mrad/hr.

Remelt Area

Working distance in the bottom remelt areas – 15 mrad/hr

Working distance while vacuum cleaning the NPR bottom mold tank sections – 40 mrad/hr 

The crucible cleaning and coating operation – 100 mrad/hr

Charging Station

Radiation inside the burnout – 40 mrad/hr

Burnout Station

Bottom north side of the burnout enclosure general background – 15 mrad/hr.

Table 2-17a lists airborne radioactive material concentrations for various operations and locations at 

Plant 9. Table 2-17b provides the estimated potential annual uranium intakes by Plant 9 worker by 

years. Table 2-17c provides the estimated thorium intakes. 
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Table 2-17a. Plant 9 airborne radioactive dust concentration by operation or location. 

(year data taken as indicated)

Operation or location

Ave. 

Conc. 

(α 

dpm/m3)

Conc. 

w/Respiratora

Bq/m3

Wet area (thorium): (1955 data) 

Operator hand scooping white salt from 30-gal drum into pre-dryer

1,394 4.65E-01

Emptying cellophane bag of white salt into predryer 59 1.97E-02

Over drum of wet thorium oxalate while operator hand scoops from 

drum to predryer

1,379

4.60E-01

In northwest corner of area while hand-filling predryer 1,056 3.52E-01

Over opening to predryer. Steam and fumes evolving from 

predryer. No hand loading of predryer 4,811 1.60E+00

North end of the building next to sump area: (1959 data) 

Operator removing slugs from box placing on angle iron rest, 

deburring and placing slugs back in box. No respirator worn 277 4.62E+00

Mold cleaning station: (1960 data)

Operator cleaning mold and coating with MgZrO3. Respirator worn

143 4.77E-02

Ingot stamping station: (1960 data) 0.00E+00

Pushing ingot down conveyor 140 4.67E-02

Stamping crop end 6,840 2.28E+00

Breaking sample piece off ingot, stamping, and breaking in two. 1,976 6.59E-01

Pot delidding station: (1960 data) 0.00E+00

Delidding pot (prior to ventilated) 1,190 3.97E-01

Delidding pot (ventilated) 281 9.37E-02

East Pad, south of the maintenance shop—mortar and brick removal 

from remelt furnace coil (done twice each week)

Breaking brick and mortar with hammer and chisel, and picking 

loose brick from inside coil and tossing into drum. (Respirator 

worn) 

24,932 8.31E+00

Laborer shoveling brick and mortar residue from bottom of tank. 

(Respirator worn)

67,039 2.23E+01

Furnace charging and capping area: (1963 data) 

Furnace pot capping---operator scoop MgF2 from the drum into 

furnace pot.

150 5.00E-02

Capping operator exposes to 620 2.07E-01

Plant 9 Zirnlo operation: 270 pounds of crushed fuel elements were 

manually transferred from 55-gal drum to Zirnlo basket and back. 

This operation was infrequent and of short duration (1980 data) 

Drum unloading (General Area) 6.73 1.12E-01

 (Breathing Zone)

Drum loading (General Area) 21.88 3.65E-01

 (Breathing Zone)

Gisholt “5L” ingot sampling lathe located near the Zirnlo: (1981 data) 

Before modification to reduce cutting fumes

Inventory table northwest of lathe (general area) 59 9.83E-01

End of wall at south end of lathe 56.7 9.45E-01

Bench & chair west of lathe where operator rests 39.3 6.55E-01

On lathe operator (breathing zone) 11/20 695 2.32E-01

On lathe tool bench 35 5.83E-01

Behind lathe near chip pile 73 1.22E+00
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Operation or location

Ave. 

Conc. 

(α 

dpm/m3)

Conc. 

w/Respiratora

Bq/m3

On lathe operator (breathing zone) 12/1 270 9.00E-02

After modification to reduce cutting fumes

On operator 191 6.37E-02

On ticket table north of lathe 42.4 1.41E-02

On lathe tool feed bench near where operator stands when 

checking tool and cleaning chips off tool 154.8 5.16E-02

On vacuum barrel 25 ft southwest of lathe tool rest near south aisle

33 5.50E-01

On south end of ingot conveyor east lathe 58.8 9.80E-01 a

 A Decontamination Factor (DF) of 50 for wearing respirator was applied to Breathing Zone (BZ) concentrations 

for selected process activities.

Table 2-17b. Projected annual estimated uranium intake by Plant 9 workers.

Years

Adjustment Factorsa Average 

Uranium 

Conc.

Bq/m3

Average Annual 

Uranium Intakesb

Bq/yrb 1955-1963

Data 1981 Data

1957-1964 1.0 4.0 1.78E+00 4.26E+03

1965-1968 0.9 3.6 1.60E+00 3.84E+03

1969-1971 1.0 4.0 1.78E+00 4.26E+03

1972-1975 0.6 2.4 1.07E+00 2.56E+03

1976-1981 0.25 1.0 4.44E-01 1.07E+03

1982-1984 0.74 2.9 1.30E+00 3.13E+03

1985-1988 0.5 1.0 7.01E-01 1.68E+03 a

 The measured airborne radioactivity concentrations are adjusted for other years than the year in 

which the measurements are made, in accordance with the percentage of process system 

throughput capacity in one shift. In absence of specific throughput information, Appendix 2A is 

used as the default model for FEMP production rates for the derivation of the adjustment 

factors. The “ FEMP uranium emissions summary by plant” in Appendix 2B is used to identify 

operating years. 1973 data were taken at the end of a campaign, since then, it is assumed that 

the measured concentrations were resulting from full system capacity. b

 Radionuclide (e.g., Pu, Np, Tc and U isotopes) radioactivity fractions for recycled uranium is 

given in Table 2-54. 

Table 2-17c. Projected annual estimated thorium intake by Plant 9 workers.

Years Adjustment Factora

Average 

Thorium Conc.b

Bq/m3

Average Annual 

Thorium Intakes

Bq/yr

1954 1.0 7.28E-02 1.75E+02

1955 1.0 7.28E-02 1.75E+02 a

 The thorium concentrations for the casting operations are assumed to be the same in weight/vol.

as those estimated for uranium in 1954 and 1955. b

 The thorium concentrations in Bq/m3 are obtained by multiplying U concentrations by the Th/U 

specific activity ratio.

2.3 WASTE MANAGEMENT FACILITIES ACTIVITIES AND PROCESSES

Figure 2-1 shows locations of waste treatment and storage facilities in and outside the fenced 

production area of the FEMP. The Waste Storage Area outside the production area includes six lowlevel radioactive waste storage pits, two earthen-bermed concrete silos containing K-65 residues 

(high specific activity, low-level radium-bearing residues), one concrete silo containing metal oxides, 
Effective Date: 05/20/2004 Revision No. 00 Document No. ORAUT-TKBS-0017-2 Page 49 of 83

and affected adjoining areas. It also includes two fly ash piles approximately 3,000 feet southsouthwest of the waste storage area, as well as the burn pit between Pits 3 and 4.

2.3.1

A wide variety of chemical and metallurgical processes were used at FEMP to produce high-purity 

uranium metal. As a result, various waste forms were generated. Both solid and liquid wastes were 

produced. Radioactive and mixed wastes from FEMP production processes included wet filter cake, 

sludge, neutralized raffinate, dry slag, ash, metallic uranium fines, oxides, and miscellaneous trash. 

Production Wastes

Dry filter cakes from the Recovery Plant, resulting from the filtration of raffinate and sludge that 

accumulated in the general sump, are stored in drums for disposal at the Nevada Test Site (NTS). 

Wet filter cakes were drummed and stored on the FEMP and later (after 1986), the wet filter cakes 

were further dried in a rotary hearth furnace and drummed for disposal at NTS.

Combustible residues, sewage sludge, graphite, and oils that contain low-level radioactive wastes

were treated as process residues and incinerated in various facilities. Uranium was recovered from 

the generated ash in the Recovery Plant of the refinery.

Noncombustible solid wastes, including clothing and gloves, generated at the FEMP have been boxed 

or drummed for offsite shipment and disposal at NTS. The uranium content of these wastes is low 

and uranium recovery is not feasible.

The chemical composition of the production wastes is quite varied, but major components are metal 

oxides and nitrates of copper, aluminum and iron; calcium oxide; free uranium and uranium in various 

oxidation states; magnesium fluoride; and traces of free magnesium. 

Several radionuclides including uranium and thorium, small quantities of fission products (90Sr, 137Cs, 

and 99Tc), transuranics (neptunium and plutonium), and non-uranics are present in the radioactive and 

mixed wastes. 

The uranium content of the various waste streams varies from about 0.0005 grams U/gram of solids in 

the neutralized refinery raffinate and slag leach filter cake to 80% to 100% U in discarded metal and 

oxides. The percent enrichment in 235U is 0.2% to 1.1% in the waste materials.

Since 1972, the FEMP has served as the thorium materials repository for DOE. Approximately twothirds of the material in the repository was processed at the FEMP. The remainder originated at other 

DOE facilities. An inventory of thorium is stored in the production areas of the site, and some thorium 

is in relatively high-traffic areas (Plant 8 silo). Table 2-18 lists the FEMP thorium inventory as of 1988. 

Table 2-18. Location and number of thorium 

drums*. 

Location Thorium containers

Building 65 5,836 drums

Pilot Plant 1,313 drums

Building 67 382 drums, 5,500 cans

Building 64 180 cans

TOTAL 13,211

* As of 1988. 
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Many containers have been removed from the site during the remediation process. The remaining 

thorium (approximately 500,000 lb) is stored in approximately 200 metal boxes. The thorium is 

primarily a mixture of thorium metal, thorium oxides, and process residues.

2.3.2

The waste treatment, storage, and management facilities directly associated with FEMP production 

operations include the Thorium Overpacking Operation, Ferrous Metal Scrap Pile, Copper Scrap Pile, 

Waste Oil Storage Pad, Oil Burner, Plant 1 Storage Pad, Solid Waste Incinerator, Pilot Plant Tank 

Farm, Pilot Plant Warehouse, KC-2 Warehouse, and Refinery Storage Pad.

Production Area Waste Treatment and Storage Facilities

2.3.2.1 Thorium Buildings and Overpacking Operations

The thorium buildings (Buildings 64, 65, 67, and 68) were constructed with preengineered 

warehouse/storage buildings. During the 1990s, drums of thorium stored in various plant locations, 

including those in Building 65 and 67, were overpacked and transferred to Buildings 64 and 78 for 

interim storage and eventual offsite shipment. Structurally unsound drums were placed inside larger 

containers by overpacking. The overpacked and structurally sound drums were transferred to 

Buildings 64 and 78 and subsequently placed in strong tight containers (STCs) suitable for offsite 

shipment. 

The stored thorium compounds (thorium hydroxide, oxide, and oxalates) came from two separate time 

periods. In the 1950s, thorium metal was produced for the nuclear aircraft program that was 

cancelled in 1961. In the mid-1960s, thorium oxides were produced for the Light Water Breeder 

Reactor Program. The stored materials are out-of-specification production runs, process byproducts, 

or excess materials returned from downstream production facilities when the materials were no longer 

required. 

To assess exposures to radiation associated with thorium operations (overpacking or storage), all of 

the stored thorium is assumed to be 232Th. It is considered to be 35 years old for the purpose of 

determining daughter isotope concentrations. The average total thorium inventory in a typical drum is 

approximately 70 kg, and the dispersible component of the thorium inventory in a drum is 

approximately 57 kg.

External gamma radiation exposure can be as great as 70 mR/hr from a drum. Table 2-19 lists 

external radiation by job and the overpacking operational activities.

Table 2-19. Dose estimates for thorium overpacking operations. 

Operation/personnel

Dose rates (mR/hr) 

w/o shieldg/w/shieldg/bkgdd Handling time (min.)

Unshielded dose 

(mR/drum)

Shielded dose 

(mR/drum)

Container handling 50 / 3.0 / 7.0 10a

Fork-truck operator 9.5 0.4

Overpacking material identification. 60 / 10.0 / 7.0 30b

Overpacking operator 26.6 1.1

Laborer 7.1 0.3

Th overpack handling. 40 / 1.0 / 7.0 10c

Fork-truck operator 2.6 0.06

a. Time for handling per drum, that is, time from initial drum pickup to placement on the conveyor. Pickup 5 min, Transport 2 min, 

Conveyor Placement 2 min.

b. If a drum does not require repackaging or sampling, identification/placement into six-packs is estimated to take 20 min. The estimated 

majority (85%) of the drums would be in this condition. Drums requiring sampling and/or repackaging are estimated to require 40 min. 

An estimated 10% of the drums would be in this condition. Potentially pyrophoric drums require sampling and special handling. The 

estimated average handling time for the remaining 10% of the drums is 50 min per drum. The average drum handling time is estimated 

to take 26 min. to place into six-packs. To this average handling time is added 4 min to account for urethane foam application, foam 

trimming, and six-pack lid emplacement. Therefore, the total average handling time is 30 min.
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c. The thorium overpack handling time is the estimated elapsed time from six-pack pickup by the fork-truck at the portable shield area to 

placement of the six-pack in Building 64.

d. Shielding provided at the fork-truck operator cage and in the jib crane area for worker to stand behind. Background dose rate is 

referred to as ambient dose rate in the building. 

In relation to gamma exposure to other workers of the thorium overpacking operation, the exposure 

data from1992 job-specific thermoluminescent dosimeters for overpacking operations in Buildings 68 

and 67 (800 drums) in Table 2-20 can be used for dose estimation. 

Table 2-20. Average annual exposure for Building 68 and 67 overpacking staff (800 

drums). 

Total 

staff

Mean 

dose

Dose (mrem)

0-10 >10-50 >50-100 >100-500 >500-1000

93 (100%) 73 mrem 31% 40% 13% 13% 3%

The source terms for the estimation of internal dose due to normal thorium dust inhalation are based 

on expected events that could release dispersible thorium dust in operating areas. Table 2-21 is a 

summary of dispersible thorium in a typical thorium drum. 

Table 2-22 lists expected airborne thorium as a result of the overpacking operation on an annual 

basis.

Table 2-21. Summary of dispersible thorium in a typical thorium drum. 

Locations No. of drums Thorium (grams) gm/druma

Building 64 180 1.23E+05 6.83E+02

Building 65 5,599 3.22E+08 5.75E+04

Building 64 overpacked 240 1.90E+07 7.91E+04

Total 6,019 3.41E+08 5.66E+04 a Average g/drum is the total grams divided by total drums.

Table 2-22. Expected airborne thorium during the overpacking operation. 

Event

Frequency 

(event/year)

Th released 

(gm/event) a

Annual release 

(gm/year)

Fork impact 17 1.62E+01 2.75E+02

Fork drop 17 3.23E+01 5.50E+02

Physical degradation 42 6.90E-01 2.91E+01

Miscellaneous 1,358 1.62E-01 2.19E+02

Total 1.07E+03 a This includes the applicable ARFs

Based on ventilation rate and reduction of airborne dust by the filtration systems, the estimated 

average thorium dust concentration is 1.97 x 10-5 gm/m3

. 

2.3.2.2 Ferrous Metal Scrap Pile

The FEMP had approximately 5,000 metric tons of metallic scrap containing above-background levels 

of uranium. This material is stored on a controlled curbed pad on the northeast corner of the site. 

The scrap pile consists primarily of ferrous material; the remainder is a mixture of aluminum, stainless 

steel, copper, brass, and nickel. The scrap includes, but is not limited to, vessels, wiring, cable, duct, 

pipe, tubing, valves, grating, sheets, plates, and miscellaneous abandoned equipment. Some 

materials at the facility are classified as low-level radioactive waste. 
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2.3.2.3 Pilot Plant Tank Farm

The Pilot Plant Tank Farm is the location of the FEMP hazardous waste tank container facility. The 

facility consists of two 10,000-gallon stainless-steel storage tanks (T-5 and T-6) on a concrete pad 

surrounded by a concrete containment dike. 

The FEMP inventory of these tanks is 13,000 gallons or 136,874 lb of spent degreasing solvents. 

Annual generation averages approximately 110 gallons (1300 lbs.) of RMI1

2.3.2.4 Pilot Plant Warehouse

degreasing solvents (45% 

methylene chloride, 55% perchloroethylene) and 110 gallons (1110 lbs.) of 1,1,1-trichloroethylene

The Pilot Plant Warehouse is a totally enclosed and roofed warehouse facility with a poured concrete 

floor and side walls, so no rain falls directly on the storage area. The storage area for hazardous 

waste at the Pilot Plant Warehouse is 7 ft wide by 62 ft long.

The Pilot Plant Warehouse container storage facility was used as the storage area for waste barium 

chloride salts shipped to the FEMP from RMI in drum containers with U.S. Department of 

Transportation (DOT) Specification numbers 17H. Waste salts are generated as a result of RMI heattreatment operations. They are a spent eutectic salt mixture containing barium chloride (45%), 

potassium chloride (32.5%), and sodium chloride (22.5%). One-hundred and sixty-seven drums of 

barium chloride are stored at this facility: 70 drums of barium chloride, 13 drums of trash 

contaminated with barium chloride, 26 drums of floor sweepings with barium chloride, and 58 drums 

of salt brick.

2.3.2.5 KC-2 Warehouse

The KC-2 Warehouse container facility is a curbed storage facility 42 ft long by 7 ft wide in Bay No. 5 

of the warehouse. The warehouse is roofed, with a poured concrete floor, concrete block divided 

(side) walls, and chain-link front and back walls. 

Spent solvents generated on the site (1,1,1-trichloroethane), as well as spent solvents generated off 

the site (45% methylene chloride/55% perchloroethylene) at the RMI facility, were generated as a 

result of degreasing operations. These solvents might temporarily be stored at the KC-2 Warehouse 

container storage facility until the liquid waste can be transferred to the bulk storage tanks at the Pilot 

Plant Tank Farm.

The KC-2 Warehouse (Bay 5) is used to store polychlorinated biphenyls (PCBs). Although PCBs are 

regulated by the Toxic Substance Control Act, PCB wastes at the FEMP also contain uranium and are 

classified as mixed wastes. Solvent still bottoms and sludges, and PCB-containing capacitors are the 

two types of PCB-contaminated wastes stored at the KC-2 Warehouse.

The inventory of PCB still bottoms and sludges is approximately 20,000 lb (37 55-gallon drums) and 

two 55-gallon drums of absorbent from cleanup operations. In addition, 19 55-gallon drums of 

capacitors, one drum of rags from capacitor cleanup, one drum of PCB-containing water (30 gallons) 

and one drum of laboratory samples are stored at this warehouse; this totals 51 drums of PCBcontaminated materials.

 

1

 From the Reactive Metals, Inc., DOE support site in Ashtabula, Ohio.
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2.3.2.6 Refinery Storage Pad

In addition to the other production storage areas, a controlled pad at the refinery (Plants 2/3) is used 

for storage of waste oil and oil sludges. The FEMP inventory at the Refinery Storage Pad is 900 55-

gal drums of waste cutting and cooling oils. Approximately one 55-gallon drum of waste oil is 

generated per day at the FEMP. In addition to the waste oil, 48 drums of oil sludges are stored at the 

Refinery Storage Pad.

2.3.3

Waste management facilities outside the FEMP Production Area include the Storage Silos, Low-Level 

Radioactive Waste Pits, Underground Storage Tanks, and Sanitary Landfill.

Waste Facilities Outside Production Area

2.3.3.1 Storage Silos

Four above ground silos (Silos 1 to 4) used for storage of radium-bearing residues from the uranium 

ore processing are on the west side of the FEMP.

Silos 1 & 2 (K-65 Silos)

The two above-ground K-65 silos were constructed in 1955. They were used for storage of radiumbearing residues, a byproduct of uranium ore processing. The silos are of cylindrical concrete 

construction, 88 ft in diameter and approximately 27 ft high. 

The silos were designed to load with metal oxides in slurry form at a maximum rate of 8,000 gal per 

day. The radioactive residues were allowed to settle and the water was decanted, leaving sludge with 

a density of 100 lb/ft3

. The maximum allowable height of solid material was 23 ft and the water level 

was limited to a maximum height of 25 ft.

In 1963, berms were constructed around each silo to provide lateral support to the silo walls and, as a 

secondary benefit, to provide shielding. In 1990, a protective coating was applied to the silos to 

minimize concrete deterioration and reduce radon emissions. 222Rn (3.8 day half-life, and 5.5 day 

mean-life) is generated at a rate in secular equilibrium with its 226Ra parent. The nature of radon, 

being an inert radioactive gas, results in the continual release of the radionuclide from K-65 residues 

into the headspace of the silos. The actual quantity of radon present in the silo headspace is 

determined by the production rate (secular equilibrium) and the loss rate. There are essentially two 

loss mechanisms: the natural decay of radon gas and the escape of the gas from cracks and 

openings in the silo structure. After a period of time, there is a steady-state quantity of radon in the 

headspace. A Radon Treatment System (RTS) was added in 1987 to reduce the level of radon gas in 

the headspace above the residues. This system was operated only when access to the silo domes 

for sampling or maintenance was required. 

Radon treatment involving a carbon collection system similar to the existing RTS results in a highly 

concentrated radon inventory with the carbon and a significant external radiation hazard. A person in 

proximity to the collection system without adequate controls would receive radiation exposures above 

allowable limits in a short period. Maintenance activities involving system entry must consider worker 

exposure to gamma radiation and radon and its immediate daughter products. 

The K-65 silo area represented one of the most prevalent sources of gamma exposure at the FEMP. 

Past projects involving significant stay times in the proximity of K-65 have produced some of the 

larger personnel whole-body exposures at the FEMP since 1990.
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From 1953 through 1955, the FEMP refinery processed pitchblende ore from the Belgian Congo. No 

chemical separation or purification had been performed prior to the receipt of the ore at the FEMP. 

The generated residues were not mixed with other site wastes but were placed in two dedicated silos. 

Pitchblende residues were first added to the concrete storage silos in 1953. The residues were batchpumped from the FEMP refinery to the silos as aqueous slurry. The supernatant liquid was withdrawn 

and pumped back to the refinery to be used in the slurrying step. Additions to the silo ended in 1955 

when the last pitchblende was processed at the FEMP. Also in 1955, pitchblende residues from 

another site were added to the silos. Core sampling of silo contents in 1972 showed a dry freeflowing powder at the surface and 40% moisture in the samples from the bottom.

A radiochemical analysis of the material in each silo material was performed in the past. Core 

sampling was performed on both silos and isotopic analyses reported for uranium, thorium, radium, 

and other radionuclides. Using the upper 95% confidence interval of the mean value, Tables 2-23 

and 2-24 list the activity per gram of each radionuclide in Silos 1 and 2, respectively. The uranium 

appears to be 1% enriched, based on the relative activity of 238U and 234U to the total activity.

Table 2-23. Isotopic composition of K-65 Silo 1 (on a per gram 

of waste basis). 

Isotope Specific Activity (nCi/gm) Activity fraction

Uranium (total) 1.68 1.61 E-3

Ac-227 7.67 17.36E-3

Pb-210 202 1.94E-1

Po-210 281 2.70E-1

Ra-226 477 4.58 E –1

Th-228 2.28 2.19 E-3

Th-230 68.9 6.62 E –2

Th-232 1.11 1.07 E -3

Table 2-24. Isotopic composition of K-65 Silo 2 (on a per gram 

of waste basis). 

Isotope Specific Activity (nCi/gm) Activity fraction

Uranium (total) 2.37 3.04 E-3

Ac-227 6.64 8.50E-3

Pa-231 4.04 5.17 E -3

Pb-210 190 2.43 E-1

Po-210 231 2.96 E-1

Ra-226 263 3.36 E -1

Th-228 7.36 9.42 E-3

Th-230 76.2 9.75 E -2

Th-232 0.99 1.26 E -3

External dose characteristics of the silos depend on the radon and radon daughter activity in the silo 

headspaces that emit gamma radiation. The actual quantity of radon present in the silo headspace is 

determined by the production rate (secular equilibrium) and the loss rate. After a period of time, there 

is a steady-state quantity of radon and radon daughters in the silo headspace. The Preliminary 

Hazard Analysis Report for OU4 Silos (FEMP 1998) used 10 curies (each silo) for the release 

calculation and projected a releasable radon inventory per silo headspace after clay covering dried 

and cracked may exceed 30 curies per silo. FEMP (1998) stated that the general area exposure rate 

at either silo dome is about 25 mrem/hr.
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Radon treatment involving a carbon collection system will result in a highly concentrated radon 

inventory in the carbon and a significant external radiation hazard. Maintenance activities involving 

system entry must consider worker exposures to gamma radiation, radon, its immediate decay 

daughters, and 210Pb, which builds up with system operation.

Silo 3

Silo 3 contains approximately 138,000 cu ft of waste residues, known as cold metal oxides, which 

were generated during extraction operations in the 1950s involving uranium concentrates. The 

concentrates were received from a variety of uranium mills in the United Stated and abroad. The 

residues in Silo 3 are substantially different from those in Silos 1 and 2. First, Silo 3 residues are dry 

(~10% moisture) while K-65 material is wet (~30% moisture). The material was calcined prior to 

placement in Silo 3 and is now a fully oxidized fine powder. Second, while the radiological 

constituents are similar to those in Silos 1 and 2, some radionuclides such as radium are present in 

much lower concentrations. The major fraction of metal impurities, including radium, has been 

removed by the uranium mills prior to transfer to the FEMP site. Without the radium contribution, Silo 

3 exhibits a much lower direct radiation field and radon emanation rate than those of Silos 1and 2. 

The immediate concern involving Silo 3 is the risk of inhalation dose from suspended material 

following a release. The radionuclide having both the largest specific activity (60 nCi/gm) and the 

greatest fraction of relative dose (83%) in Silo 3 is Th-230, which is produced from the natural decay 

of 238

U.

The isotopic composition of Silo 3 can be derived from a past radiochemical analysis of the silo 

material. Table 2-25 lists the activity per gram of waste of each radionuclide in Silo 3. The uranium 

appears to be natural, based on the relative activity of 238U and 234U to the total activity.

Table 2-25. Isotopic composition of K-65 Silo 3 

(on a per gram of waste basis). 

Isotope

Specific Activity 

(nCi/gm)

Activity 

fraction

Uranium (total) 3.63 4.83 E-2

Ac-227 0.93 1.23 E-2

Pa-231 0.63 8.35 E-3

Pb-210 3.48 4.63 E-2

Ra-224 0.37 4.89 E-3

Ra-226 3.87 5.15 E-2

Ra-228 0.41 5.42 E-3

Th-228 0.75 9.95 E-3

Th-230 60.2 8.02 E-1

Th-232 0.84 1.12 E-2

The external dose characteristics for Silo 3 are based on the radium content. While the material 

stored in Silo 3 is similar to that on Silos 1 and 2, much of the radium had been removed by uranium 

mills prior to transfer to the FEMP site, and is present in much lower concentrations than in Silos 1 

and 2. Without the radium contribution, Silo 3 exhibits a much lower direct radiation field and radon 

emanation rate than Silos 1 and 2. The primary hazard for workers is the inhalation of the fine powder 

in the silo.
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Silo 4

As with Silo 3, Silo 4 was designed to receive cold metal oxides from uranium processing. Due to a

process change in 1957, storage of the waste cold metal oxide was no longer necessary; Silo 4 was 

never used and is empty.

2.3.3.2 Low-Level Radioactive Waste Pit Area

The Waste Pit Area is west of the former production area and covers approximately 37.7 acres. It 

consists of Waste Pits 1, 2, 3, 4, 5, and 6 and the Burn Pit (also used for the disposal and burning of 

waste).

Waste Pit 1

Waste Pit 1 was constructed in 1952. Its surface area is oval-shaped with dimensions at the bottom 

of approximately 165 ft wide by 347 ft long. Its depth averages 29.5 ft, including approximately 18 ft 

of waste, 11 ft of lining, and a 6-in. cover. Waste Pit 1 is a dry pit, because waste slurries were 

filtered or calcined to remove water before they were placed in the pit. This pit primarily received 

neutralized waste filter cakes, vacuum-filtered sludges from production activities, magnesium fluoride 

slag, scrap graphite, and contaminated brick. Waste Pit 1 was closed and covered with clean fill in 

1959. Waste inventory records indicate that Waste Pit 1 contains 1,075 metric tons of uranium. The 

radionuclide distribution for buried wastes in Waste Pit 1 are listed in Table 2-26. 

Table 2-26. Waste Pit 1 isotopic content. 

Radionuclides

Activity Concentration 

(pCi/gm of waste)

Avg Min Max

Cesium-137 0.43 0.2 1.1

Neptunium-237 -- -- --

Plutonium-238 -- -- --

Plutonium-239/240 -- -- --

Ruthenium-106 -- -- --

Strontium-90 -- -- --

Technetium-99 3.6 1 15

Thorium-228 20.14 6.1 52

Thorium-230 1,504.4 326 3,600

Thorium-232 21.62 7.7 55

Uranium-234 558.4 244 1180

Uranium-235 56 16 151

Uranium-238 1,950.6 360 6,980

The results of an airborne radon flux survey for Waste Pit 1 indicated an average radon flux of 9.1 

pCi/m2

/sec with a limited regional maximum radon-flux density of 75 pCi/m2/sec on the western 

boundary of the pit, next to the Clearwell. The observed radon flux pattern is probably an indication of 

localized areas that contain higher concentrations of 226Ra, are more permeable, and allow the radon 

gas to escape more readily. Assuming an air velocity of 3.2 m/sec at the Waste Pit 1, the average 

radon-flux for Pit 1 in Working Level (WL) is 0.013 WL. 

Airborne radioactive material concentration due to air particulate emissions at Waste Pit 1 for dose 

reconstruction purposes can be estimated using the fugitive uranium and thorium emissions from wind 
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erosion as estimated in the Addendum to FEMP-2082 (Appendix 2C). Table 2-27 presents the 

derivation and results. 

Table 2-27. Airborne radionuclide concentration at Waste Pit 1 by years.

Year

A

Exposed pit area 

m2

B

Uranium 

emission

kg/yr

C

Uranium fluxa

(B/A)×28.6 

Bq/m2-sec

D

Uranium 

Concentrationb,c

Bq/m3

1953 398.82 0.15 1.08E-02 3.36E-03

1954 638.04 0.25 1.12E-02 3.50E-03

1955 3,589.01 1.4 1.12E-02 3.49E-03

1956 5,184.10 2.0 1.10E-02 3.45E-03

1957 5,582.83 2.16 1.11E-02 3.46E-03

1958 5,582.83 2.16 1.11E-02 3.46E-03

1959 5,582.83 2.16 1.11E-02 3.46E-03

1960 5,582.83 2.16 1.11E-02 3.46E-03

1961 4,785.28 1.85 1.11E-02 3.46E-03

1962 3,987.73 1.54 1.10E-02 3.45E-03

1963 1,595.09 0.62 1.11E-02 3.47E-03

1964 797.55 0.31 1.11E-02 3.47E-03

1965 797.55 0.31 1.11E-02 3.47E-03

1966 797.55 0.31 1.11E-02 3.47E-03

1967 797.55 0.31 1.11E-02 3.47E-03

1968 1,196.37 0.46 1.10E-02 3.44E-03

1969 6,380.37 2.5 1.12E-02 3.50E-03

1970 5,582.83 2.15 1.10E-02 3.44E-03

1971 4,785.28 1.85 1.11E-02 3.46E-03

1972 3,987.73 1.54 1.10E-02 3.45E-03

1973 3,190.19 1.23 1.10E-02 3.45E-03

1974 2,392.64 0.92 1.10E-02 3.44E-03

1975 1,595.09 0.62 1.11E-02 3.47E-03

1976 797.55 0.31 1.11E-02 3.47E-03 a Based on 2% enriched uranium.

b Assumed an air velocity of 3.2 m/sec at Waste Pit 1. c

 Radionuclide (e.g., Pu, Np, Tc and U isotopes) radioactivity fractions for recycled uranium is given in Table 

2-54

Direct radiation for Waste Pit 1 is within the beta/gamma dose rate contour of 1.0 mRad/hr. 

Radiological analyses for soil samples taken in these areas indicate that uranium is the principal 

constituent for the elevated dose rates.

Waste Pit 2

Waste Pit 2 was constructed northeast of Waste Pit 1 in 1957. The surface area boundary of Waste 

Pit 2 resembles a six-sided polygon with dimensions at the top of approximately 19 ft wide by 270 ft 

long. It is approximately 23.5 ft deep including 15 ft of lining and 1 to 4 ft of cover. Waste Pit 2 is a 

dry pit that received primarily waste filter cakes, vacuum-filtered production sludge, magnesium 

fluoride slag, scrap graphite, contaminated brick, and concentrated raffinate residues. It was closed 

and covered with clean fill in 1964. Waste inventory records indicate that Waste Pit 2 contains 175 

MTU. The radionuclide distribution for buried wastes in Waste Pit 2 are listed in Table 2-28. 

The results of an airborne radon flux survey for Waste Pit 2 indicated an average radon flux of 6.4 

pCi/m2

/sec. The maximum radon flux measurement in Waste Pit 2 was 81.1 pCi/m2/sec in the 

southwestern part of the pit. The raffinate waste from Plant 2/3 disposed of in this pit contains 
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elevated concentrations of 226Ra that are the source of radon. The observed radon flux pattern 

probably reflects localized areas that contain higher concentrations of 226Ra that are more permeable, 

and allow the radon to escape more readily. Assuming an air velocity of 3.2 m/sec at the Waste Pit 2, 

the average radon-flux for Pit 2 in Working Level (WL) is 0.014 WL.

Table 2-28. Waste Pit 2 isotopic content. 

Radionuclides

Activity Concentration 

(pCi/gm of waste)

Avg Min Max

Cesium-137 1.15 0.2 3.6

Neptunium-237 -- -- --

Plutonium-238 0.1 0.07 0.1

Plutonium-239/240 0.16 0.05 0.6

Ruthenium-106 -- -- --

Strontium-90 -- -- --

Technetium-99 128.9 1 618

Thorium-228 62.6 11 164

Thorium-230 1,434.8 41 3,980

Thorium-232 42.6 11 88

Uranium-234 3,866.6 40 18,200

Uranium-235 1,793 2 8,780

Uranium-238 4,725.4 60 17,900

Tables 2-29 and 2-30 provide, respectively, the estimated airborne uranium and thorium particulate 

concentration due to the fugitive uranium and thorium emissions from wind erosion at Waste Pit 2. 

Table 2-29. Airborne uranium concentration at Waste Pit 2 by years.

Year

A

Exposed pit area 

m2

B

Uranium 

emission

kg/yr

C

Uranium fluxa

(B/A)×28.6 

Bq/m2-sec

D

Uranium 

Concentrationb,c

Bq/m3

1957 1,651.85 46.4 8.03E-01 2.51E-01

1958 3,303.71 92.7 8.02E-01 2.51E-01

1959 3,303.71 92.7 8.02E-01 2.51E-01

1960 3,716.74 104.3 8.03E-01 2.51E-01

1961 4,129.68 115.9 8.03E-01 2.51E-01

1962 4,129.68 115.9 8.03E-01 2.51E-01

1963 4,129.68 115.9 8.03E-01 2.51E-01

1964 4,129.68 115.9 8.03E-01 2.51E-01

1965 2,064.89 57.95 8.03E-01 2.51E-01

1966 1,238.91 34.8 8.03E-01 2.51E-01 a Based on 2% enriched uranium. b Assumed an air velocity of 3.2 m/sec at Waste Pit 2. c

 Radionuclide (e.g., Pu, Np, Tc and U isotopes) radioactivity fractions for recycled uranium is given in Table 2-

54. 

Table 2-30. Airborne thorium concentration at Waste Pit 2 by years.

Year

A

Exposed pit area 

m2

B

Thorium

emission

kg/yr

C

Thorium flux

(B/A)×1.29

Bq/m2-sec

D

Thorium 

Concentrationa

Bq/m3

1957 1,651.85 0.014 1.08E-05 3.39E-06

1958 3,303.71 0.032 1.24E-05 3.87E-06

1959 3,303.71 0.032 1.24E-05 3.87E-06
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Year

A

Exposed pit area 

m2

B

Thorium

emission

kg/yr

C

Thorium flux

(B/A)×1.29

Bq/m2-sec

D

Thorium 

Concentrationa

Bq/m3

1960 3,716.74 0.032 1.10E-05 3.44E-06

1961 4,129.68 0.036 1.12E-05 3.49E-06

1962 4,129.68 0.036 1.12E-05 3.49E-06

1963 4,129.68 0.036 1.12E-05 3.49E-06

1964 4,129.68 0.036 1.12E-05 3.49E-06

1965 2,064.89 0.018 1.12E-05 3.49E-06

1966 1,238.91 0.01 1.03E-05 3.23E-06 a Assumed an air velocity of 3.2 m/sec at Waste Pit 2. 

Direct radiation for Waste Pit 2 is within the beta gamma dose rate contour of 1.2 mRad/hr. 

Radiological analyses for soil samples taken in these areas indicate that uranium is the principal 

constituent for the elevated dose rates.

Waste Pit 3

Waste Pit 3 was placed in service in December 1958. The surface area boundary of the pit is ovalshaped and has dimensions of approximately 450 ft wide by 720 ft long. This pit is approximately 42 

ft deep, which includes a maximum of 14 ft of cover material, 27 ft of wastes, and an estimated 1 ft of 

native clay. Waste Pit 3 was the first waste pit built specifically for settling solids from liquid waste 

streams. Primarily lime-neutralized raffinate slurries and contaminated surface-water runoff were 

pumped to this pit. After Waste Pit 2 was filled, Waste Pit 3 received vacuum-filtered production 

sludge, neutralized liquid from process systems, neutralized refinery sludges, and cooling water from 

heat treatment operations. During the late 1960s, large quantities of neutralized residues from acid 

leaching of uranium-bearing magnesium fluoride slag were pumped to Waste Pit 3. In 1973, fill 

material (including filter cake, slag leach residue, lime sludge, and flyash) was placed in Waste Pit 3 

and construction activities were initiated to cover this waste with soil. Covering activities were 

complete in 1977. Waste inventory records indicate that Waste Pit 3 contains 846 MTU. The 

radionuclide distribution for buried wastes in Waste Pit 3 are listed in Table 2-31. 

Table 2-31. Waste Pit 3 isotopic content. 

Radionuclides

Activity Concentration (pCi/gm 

of waste)

Avg Min Max

Cesium-137 -- -- --

Neptunium-237 0.53 0.2 2.1

Plutonium-238 0.23 0.08 1

Plutonium-239/240 2.13 0.05 14

Ruthenium-106 -- -- --

Strontium-90 1.31 0.6 5.2

Technetium-99 233.4 1.2 1,110

Thorium-228 11.79 1 40

Thorium-230 4,635.7 15 1,1680

Thorium-232 15.74 1 59

Uranium-234 207.71 27 475

Uranium-235 11.7 2.5 21

Uranium-238 442 134 1,380

The results of airborne radon flux survey for Waste Pit 3 indicated an average radon flux of 2.6 

pCi/m2

/sec. The maximum radon-flux reading in Waste Pit 3 was 48 pCi/m2/sec located on the south 
Effective Date: 05/20/2004 Revision No. 00 Document No. ORAUT-TKBS-0017-2 Page 60 of 83

part of the pit. The raffinate waste from Plant 2/3 disposed of in this pit contains elevated 

concentrations of 226Ra that is the source of radon. The observed radon-flux pattern probably reflects 

localized areas that contain higher concentrations of 226Ra, are more permeable, and allow the radon 

gas to escape more readily. Assuming an air velocity of 3.2 m/sec at the Waste Pit 2, the average 

radon-flux for Pit 2 in Working Level (WL) is 0.008 WL.

Table 2-32 provides the estimated airborne uranium particulate concentration and Table 2-33 

provides the estimated airborne thorium concentration resulting from the fugitive uranium and thorium 

emissions from wind erosion at Waste Pit 3.

Table 2-32. Airborne uranium concentration at Waste Pit 3 by years.

Year

A

Exposed pit area 

m2

B

Uranium 

emission

kg/yr

C

Uranium fluxa

(B/A)×28.6 

Bq/m2-sec

D

Uranium 

Concentrationb,c

Bq/m3

1962 4,127.64 0.63 4.37E-03 1.36E-03

1963 4,127.64 0.63 4.37E-03 1.36E-03

1964 22,014.14 3.36 4.37E-03 1.36E-03

1965 24,765.93 3.78 4.37E-03 1.36E-03

1966 24,765.93 3.78 4.37E-03 1.36E-03

1967 0.00 0.0 0.0 0.0

1968 0.00 0.0 0.0 0.0

1969 9,631.22 1.47 4.37E-03 1.36E-03

1970 16,510.65 2.52 4.37E-03 1.36E-03

1971 16,510.65 2.52 4.37E-03 1.36E-03

1972 8,255.28 1.26 4.37E-03 1.36E-03

1973 8,255.28 1.26 4.37E-03 1.36E-03

1974 2,751.79 0.42 4.37E-03 1.36E-03

1975 22,014.14 3.35 4.35E-03 1.36E-03

1976 22,014.14 3.35 4.35E-03 1.36E-03

1977 22,014.14 3.35 4.35E-03 1.36E-03

1978 19,262.44 2.94 4.37E-03 1.36E-03

1979 16,510.65 2.52 4.37E-03 1.36E-03

1980 12,383.01 1.89 4.37E-03 1.36E-03

1981 9,631.22 1.47 4.37E-03 1.36E-03

1982 4,127.64 0.63 4.37E-03 1.36E-03 a Based on 2% enriched uranium.

b Assumed an air velocity of 3.2 m/sec at Waste Pit 3. c

 Radionuclide (e.g., Pu, Np, Tc and U isotopes) radioactivity fractions for recycled uranium is given in Table 2-54c 

Based on 2% enriched uranium

Table 2-33. Airborne thorium concentration at Waste Pit 3 by years.

Year

A

Exposed pit area 

m2

B

Thorium

emission

kg/yr

C

Thorium flux

(B/A)×1.29

Bq/m2-sec

D

Thorium 

Concentrationa

Bq/m3

1962 4,127.64 0.0 0.0 0.0

1963 4,127.64 0.0 0.0 0.0

1964 22,014.14 0.01 5.81E-07 1.82E-07

1965 24,765.93 0.014 7.24E-07 2.26E-07

1966 24,765.93 0.014 7.24E-07 2.26E-07

1967 0.00 0.0 0.0 0.0

1968 0.00 0.0 0.0 0.0
Effective Date: 05/20/2004 Revision No. 00 Document No. ORAUT-TKBS-0017-2 Page 61 of 83

Year

A

Exposed pit area 

m2

B

Thorium

emission

kg/yr

C

Thorium flux

(B/A)×1.29

Bq/m2-sec

D

Thorium 

Concentrationa

Bq/m3

1969 9,631.22 0.005 6.65E-07 2.08E-07

1970 16,510.65 0.01 7.75E-07 2.42E-07

1971 16,510.65 0.01 7.75E-07 2.42E-07

1972 8,255.28 0.005 7.75E-07 2.42E-07

1973 8,255.28 0.005 7.75E-07 2.42E-07

1974 2,751.79 0.0 0.0 0.0

1975 22,014.14 0.01 5.81E-07 1.82E-07

1976 22,014.14 0.01 5.81E-07 1.82E-07

1977 22,014.14 0.01 5.81E-07 1.82E-07

1978 19,262.44 0.01 6.65E-07 2.08E-07

1979 16,510.65 0.01 7.75E-07 2.42E-07

1980 12,383.01 0.005 5.17E-07 1.62E-07

1981 9,631.22 0.005 6.65E-07 2.08E-07

1982 4,127.64 0.0 0.0 0.0 a Assumed an air velocity of 3.2 m/sec at Waste Pit 3.

Direct radiation in the form of beta/gamma dose rates range geometrically from 1.2 mRad/hr to 2.7 

toward the middle of the pit. A high dose rate of 8.5 mRad/hr was measured at the northeastern 

boundary of the pit and a maximum beta gamma dose rate of 9.5 mRad/hr was measured in the 

northeast end of the pit. Radiological analyses for soil samples taken in these areas indicate that 

uranium is the principal constituent responsible for elevated dose rates. 

Waste Pit 4

Waste Pit 4 was constructed in 1960. The surface area boundary of the waste pit is trapezoidal in 

shape and has maximum dimensions of approximately 380 ft wide by 310 ft long. The pit is 

approximately 32 ft deep, including 25 ft of waste, 1 to 2 ft of liner, and 6 ft of cover. It received solid 

wastes that included process residues, scrap uranium metal, off-specification intermediate uranium 

products and residues, thorium metal and residues, and contaminated ceramics. Process residues 

included filter sludges, raffinates, graphite, magnesium fluoride slag, and pyrophoric uranium-bearing 

materials. Thorium metal and residues were placed in Waste Pit 4 when additional metal recovery 

was not economically feasible. Waste Pit 4 disposal activities were terminated in 1985. The pit was 

closed in 1986 and cover activities were started. Waste inventory records indicate that Waste Pit 4 

contains 2,203 MT of uranium and 74 MT of thorium. The radionuclide distribution for buried wastes 

in Waste Pit 4 are listed in Table 2-34. 

The airborne radon flux measurements for Waste Pit 4 were considerably less than those for Pits 1, 2 

or 3. All samples indicated an average radon flux of < 0.1 pCi/m2

/sec or < 3.13E-05 WL, in term of 

concentration.

Tables 2-35 and 2-36 provide, respectively, the estimated airborne uranium and thorium particulate 

concentration due to the fugitive uranium and thorium emissions from wind erosion at Waste Pit 4.

Table 2-34. Waste Pit 4 isotopic content. 

Radionuclides

Activity Concentration 

(pCi/gm of waste)

Avg Min Max

Cesium-137 -- -- --

Neptunium-237 0.18 0.1 0.4

Plutonium-238 0.25 0.1 0.5
Effective Date: 05/20/2004 Revision No. 00 Document No. ORAUT-TKBS-0017-2 Page 62 of 83

Radionuclides

Activity Concentration 

(pCi/gm of waste)

Avg Min Max

Plutonium-239/240 0.15 0.1 0.4

Strontium-90 -- -- --

Technetium-99 80.45 6.8 225

Thorium-228 114.75 18 395

Thorium-230 428.25 225 566

Thorium-232 37 17 92

Uranium-234 783.25 149 2,320

Uranium-235 152.25 35 426

Uranium-238 4,644 509 15,800

Table 2-35. Airborne uranium concentration at Waste Pit 4 by years.

Year

A

Exposed pit area 

m2

B

Uranium 

emission

kg/yr

C

Uranium fluxa

(B/A)×28.6 

Bq/m2-sec

D

Uranium 

Concentrationb,c

Bq/m3

1960 890.26 1.9 6.10E-02 1.91E-02

1961 1,780.52 3.79 6.09E-02 1.90E-02

1962 3,561.14 7.58 6.09E-02 1.90E-02

1963 4,006.22 8.52 6.08E-02 1.90E-02

1964 4,451.40 9.47 6.08E-02 1.90E-02

1965 4,451.40 9.47 6.08E-02 1.90E-02

1966 4,896.57 10.42 6.09E-02 1.90E-02

1967 4,896.57 10.42 6.09E-02 1.90E-02

1968 5,341.66 11.36 6.08E-02 1.90E-02

1969 5,341.66 11.36 6.08E-02 1.90E-02

1970 5,341.66 11.36 6.08E-02 1.90E-02

1971 5,341.66 11.36 6.08E-02 1.90E-02

1972 5,786.83 12.3 6.08E-02 1.90E-02

1973 5,786.83 12.3 6.08E-02 1.90E-02

1974 5,786.83 12.3 6.08E-02 1.90E-02

1975 6,231.92 13.26 6.09E-02 1.90E-02

1976 6,677.09 14.2 6.08E-02 1.90E-02

1977 7,567.36 16.1 6.08E-02 1.90E-02

1978 8,457.62 18.0 6.09E-02 1.90E-02

1979 8,902.79 18.94 6.08E-02 1.90E-02

1980 8,902.79 18.94 6.08E-02 1.90E-02

1981 8,902.79 18.94 6.08E-02 1.90E-02

1982 8,902.79 18.94 6.08E-02 1.90E-02

1983 8,902.79 18.94 6.08E-02 1.90E-02

1984 8,902.79 18.94 6.08E-02 1.90E-02

1985 8,902.79 18.94 6.08E-02 1.90E-02

1986 8,902.79 18.94 6.08E-02 1.90E-02

1987 8,902.79 18.94 6.08E-02 1.90E-02

1988 8,902.79 18.94 6.08E-02 1.90E-02 a Based on 2% enriched uranium.

b Assumed an air velocity of 3.2 m/sec at Waste Pit 4. c

 Radionuclide (e.g., Pu, Np, Tc and U isotopes) radioactivity fractions for recycled uranium is given in Table 2-54
Effective Date: 05/20/2004 Revision No. 00 Document No. ORAUT-TKBS-0017-2 Page 63 of 83

Table 2-36. Airborne thorium concentration at Waste Pit 4 by years.

Year

A

Exposed pit area 

m2

B

Thorium

emission

kg/yr

C

Thorium flux

(B/A)×28.6 

Bq/m2-sec

D

Thorium 

Concentrationa

Bq/m3

1960 890.26 0.38 5.46E-04 1.71E-04

1961 1,780.52 0.76 5.46E-04 1.71E-04

1962 3,561.14 1.53 5.50E-04 1.72E-04

1963 4,006.22 1.72 5.50E-04 1.72E-04

1964 4,451.40 1.91 5.49E-04 1.72E-04

1965 4,451.40 1.91 5.49E-04 1.72E-04

1966 4,896.57 2.1 5.49E-04 1.72E-04

1967 4,896.57 2.1 5.49E-04 1.72E-04

1968 5,341.66 2.3 5.51E-04 1.72E-04

1969 5,341.66 2.3 5.51E-04 1.72E-04

1970 5,341.66 2.3 5.51E-04 1.72E-04

1971 5,341.66 2.3 5.51E-04 1.72E-04

1972 5,786.83 2.49 5.51E-04 1.72E-04

1973 5,786.83 2.49 5.51E-04 1.72E-04

1974 5,786.83 2.49 5.51E-04 1.72E-04

1975 6,231.92 2.68 5.50E-04 1.72E-04

1976 6,677.09 2.87 5.50E-04 1.72E-04

1977 7,567.36 3.25 5.50E-04 1.72E-04

1978 8,457.62 3.64 5.51E-04 1.72E-04

1979 8,902.79 3.83 5.51E-04 1.72E-04

1980 8,902.79 3.83 5.51E-04 1.72E-04

1981 8,902.79 3.83 5.51E-04 1.72E-04

1982 8,902.79 3.83 5.51E-04 1.72E-04

1983 8,902.79 3.83 5.51E-04 1.72E-04

1984 8,902.79 3.83 5.51E-04 1.72E-04

1985 8,902.79 3.83 5.51E-04 1.72E-04

1986 8,902.79 3.83 5.51E-04 1.72E-04

1987 8,902.79 3.83 5.51E-04 1.72E-04

1988 8,902.79 3.83 5.51E-04 1.72E-04 a Assumed an air velocity of 3.2 m/sec at Waste Pit 4. 

Direct radiation for Waste Pit 4 is within the beta/gamma dose rate contour of 1.3 mRad/hr. A 

maximum beta/gamma dose rate reading of 11 mRad/hr occurred in the southeast corner of the pit. 

An area between the southeast corner of Waste Pit 4 and the south corner of Waste Pit 6 exhibit a 

relatively high dose rate of 35 mRad/hr. Radiological analyses for soil samples taken in these areas 

indicate that uranium is the principal constituent causing elevated dose rates.

Waste Pit 5

Waste Pit 5 was placed into service in 1968. The pit surface area boundary is rectangular in shape 

and is approximately 820 ft long by 240 ft wide. The waste pit is approximately 29 ft deep, including 

28 to 29 ft of wastes, with a membrane liner. Waste Pit 5 was a settling basin for slurries including 

neutralized raffinates, slurries from the acid leaching of uranium-bearing slags, and sump slurries that 

were generally filtered to remove solids. Supernatant and sludges, produced by the coprecipitation of 

thorium wastes with barium carbonate and aluminum sulfate and the precipitation uranium with 

calcium oxide, were disposed of in Waste Pit 5. The discharge of slurried waste materials into the pit 

ended in 1983, and use of this pit as a settling basin ended in 1987. Waste inventory records indicate 

that Waste Pit 5 contains 527 MT of uranium and 72 MT of thorium. The radionuclide distribution for 

buried wastes in Waste Pit 5 are listed in Table 2-37. 
Effective Date: 05/20/2004 Revision No. 00 Document No. ORAUT-TKBS-0017-2 Page 64 of 83

Table 2-37. Waste Pit 5 isotopic content. 

Radionuclides

Activity Concentration 

(pCi/gm of waste)

Avg Min Max

Cesium-137 24.42 2 76

Neptunium-237 8.61 0.3 19

Plutonium-238 1.99 0.1 4.4

Plutonium-239/240 5.03 0.1 13

Ruthenium-106 -- -- --

Strontium-90 9.95 1.4 31

Technetium-99 1,274.5 423 2,990

Thorium-228 31.13 3.8 44

Thorium-230 4,475 330 8,480

Thorium-232 30.47 3.8 55

Uranium-234 638.33 310 1,250

Uranium-235 34.33 14 79

Uranium-238 641.17 387 1,230

The airborne radon flux density for Waste Pit 5 is assumed to be less than 0.1 pCi/m2

/sec or 3.13E-05 

WL.

Tables 2-38 and 2-39 provide the estimated airborne uranium and thorium particulate concentration 

due to the fugitive uranium and thorium emissions from wind erosion at Waste Pit 5,respectively.

Table 2-38. Airborne uranium concentration at Waste Pit 5 by years.

Year

A

Exposed pit area 

m2

B

Uranium 

emission

kg/yr

C

Uranium fluxa

(B/A)×28.6 

Bq/m2-sec

D

Uranium 

Concentrationb,c

Bq/m3

1974 1,659.19 0.28 4.83E-03 1.51E-03

1975 4,147.89 0.71 4.90E-03 1.53E-03

1976 13,273.27 2.27 4.89E-03 1.53E-03

1977 4,977.49 0.85 4.88E-03 1.53E-03

1978 4,977.49 0.85 4.88E-03 1.53E-03

1979 4,977.49 0.85 4.88E-03 1.53E-03

1980 4,977.49 0.85 4.88E-03 1.53E-03

1981 4,977.49 0.85 4.88E-03 1.53E-03

1982 4,977.49 0.85 4.88E-03 1.53E-03

1983 4,977.49 0.85 4.88E-03 1.53E-03

1984 4,977.49 0.85 4.88E-03 1.53E-03

1985 4,977.49 0.85 4.88E-03 1.53E-03

1986 4,977.49 0.85 4.88E-03 1.53E-03

1987 8,295.60 1.42 4.90E-03 1.53E-03

1988 11,066.62 1.9 4.91E-03 1.53E-03 a Based on 2% enriched uranium.

b Assumed an air velocity of 3.2 m/sec at Waste Pit 5. c

 Radionuclide (e.g., Pu, Np, Tc and U isotopes) radioactivity fractions for recycled uranium is given in Table 2-54
Effective Date: 05/20/2004 Revision No. 00 Document No. ORAUT-TKBS-0017-2 Page 65 of 83

Table 2-39. Airborne thorium concentration at Waste Pit 5 by years.

Year

A

Exposed pit area 

m2

B

Thorium

emission

kg/yr

C

Thorium flux

(B/A)×28.6 

Bq/m2-sec

D

Thorium 

Concentrationa 

Bq/m3

1974 1,659.19 0.1 7.71E-05 2.41E-05

1975 4,147.89 0.24 7.41E-05 2.31E-05

1976 13,273.27 0.76 7.33E-05 2.29E-05

1977 4,977.49 0.29 7.46E-05 2.33E-05

1978 4,977.49 0.29 7.46E-05 2.33E-05

1979 4,977.49 0.29 7.46E-05 2.33E-05

1980 4,977.49 0.29 7.46E-05 2.33E-05

1981 4,977.49 0.29 7.46E-05 2.33E-05

1982 4,977.49 0.29 7.46E-05 2.33E-05

1983 4,977.49 0.29 7.46E-05 2.33E-05

1984 4,977.49 0.29 7.46E-05 2.33E-05

1985 4,977.49 0.29 7.46E-05 2.33E-05

1986 4,977.49 0.29 7.46E-05 2.33E-05

1987 8,295.60 0.48 7.41E-05 2.31E-05

1988 11,066.62 0.64 7.40E-05 2.31E-05 a Assumed an air velocity of 3.2 m/sec at Waste Pit 5. 

For dose reconstruction purposes, a beta/gamma direct radiation dose rate of less than 1 mRad/hr 

can be assumed for Waste Pit 5. 

Waste Pit 6

Waste Pit 6 was completed in 1979. It is square in shape with sides measuring approximately 210 ft. 

It is approximately 24 ft deep, measured from the top of the berm to the liner, but the depth of the 

wastes in the pit is only 20 ft. Waste Pit 6 received only noncoarse nonpyrophoric materials (this 

excludes uranium and thorium in metallic form), including magnesium fluoride slag, process residues, 

and filter cakes from vacuum filtering operations to protect the membrane liner. In addition, extrusion 

residue and heat treatment quench water were deposited in Waste Pit 6. Use of the pit ended in 

1985. Waste Pit 6 is currently covered by water. Waste inventory records indicate that Waste Pit 6 

contains 1,432 MT of uranium. The radionuclide distribution for buried wastes in Waste Pit 6 are 

listed in Table 2-40. 

Table 2-40. Waste Pit 6 isotopic content. 

Radionuclides

Activity Concentration 

(pCi/gm of waste)

Avg Min Max

Cesium-137 17 4 31

Neptunium-237 2.03 0.9 3.6

Plutonium-238 0.78 0.4 1.4

Plutonium-239/240 8.7 4 15

Ruthenium-106 -- -- --

Strontium-90 4.1 2.6 5.1

Technetium-99 127 84 164

Thorium-228 0.5 0.2 1.2

Thorium-230 31 14 44

Thorium-232 0.5 0.2 1.2

Uranium-234 3,417.5 2,000 5,330

Uranium-235 1,042.5 350 1,750

Uranium-238 16,975 12,500 18,700
Effective Date: 05/20/2004 Revision No. 00 Document No. ORAUT-TKBS-0017-2 Page 66 of 83

The airborne radon flux density for Waste Pit 6 is assumed to be less than 0.1 pCi/m2/sec or 3.13E-05 

WL, for dose reconstruction purposes. 

Table 2-41 provides the estimated airborne uranium particulate concentration due to the fugitive 

uranium emissions from wind erosion at Waste Pit 6.

Table 2-41. Airborne uranium concentration at Waste Pit 6 by years.

Year

A

Exposed pit area 

m2

B

Uranium 

emission

kg/yr

C

Uranium fluxa

(B/A)×28.6 

Bq/m2-sec

D

Uranium 

Concentrationb,c

Bq/m3

1979 301.00 8.26 7.85E-01 2.45E-01

1980 451.49 12.4 7.85E-01 2.45E-01

1981 752.49 20.7 7.87E-01 2.46E-01

1982 752.49 20.7 7.87E-01 2.46E-01

1983 752.49 20.7 7.87E-01 2.46E-01

1984 752.49 20.7 7.87E-01 2.46E-01

1985 752.49 20.7 7.87E-01 2.46E-01

1986 752.49 20.7 7.87E-01 2.46E-01

1987 752.49 20.7 7.87E-01 2.46E-01

1988 752.49 20.7 7.87E-01 2.46E-01 a Based on 2% enriched uranium.

b Assumed an air velocity of 3.2 m/sec at Waste Pit 6. c

 Radionuclide (e.g., Pu, Np, Tc and U isotopes) radioactivity fractions for recycled uranium is given in Table 

2-54

Waste Pit 6 is within the beta/gamma direct radiation dose rate contour of 1.3 mRad/hr. The highest 

dose rate, 35 mRad/hr, was identified near the southwest perimeter of Waste Pit 6.

Clearwell

The Clearwell was constructed in 1959 during Waste Pit 3 construction activities. It is approximately 

200 ft long by 180 ft wide, with a maximum depth of 27 ft. It was a final settling basin for surfacewater runoff from the waste pits and supernatant from Waste Pits 3 and 5. The Clearwell was 

dredged in the late 1960s or early 1970s, but has never been emptied or dredged again. 

Measurements indicate the presence of approximately 11 ft of sludge in the bottom of the Clearwell.

The radionuclide distribution in the Clearwell are listed in Table 2-42. 

The airborne radon flux density for the Clearwell is assumed to be less than 0.1 pCi/m2

/sec or 3.13E05 WL for dose reconstruction purposes. 

The Clearwell is within the beta/gamma direct radiation dose rate contour of 1 mRad/hr. 

Burn Pit

The Burn Pit is roughly between Waste Pits 2 and 4; it was initially excavated in 1957 to provide clay 

for lining Pits 1 and 2. Subsequently, combustible materials, including pyrophoric and reactive 

chemicals, oils, and other low-level contaminated combustible materials were burned in this pit. In 

addition, laboratory chemicals were disposed of at this site. Backfilling of the Burn Pit occurred prior 

to 1984. In 1984, all of the contents of the burn pit were transferred to Waste Pit 4.

The radionuclide isotopic content and isotopic distribution for buried wastes in Burn Pit are listed in 

Table 2-43. 
Effective Date: 05/20/2004 Revision No. 00 Document No. ORAUT-TKBS-0017-2 Page 67 of 83

Table 2-42. Clearwell isotopic content. 

Radionuclices

Activity Concentration (pCi/gm of waste)

Avg Min Max

Cesium-137 189.5 18 450

Neptunium-237 1.04 0.3 2.2

Plutonium-238 -- -- --

Plutonium-239/240 -- -- --

Ruthenium-106 -- -- --

Strontium-90 8.48 1.3 26

Technetium-99 98.85 0.4 278

Thorium-228 36.75 20 56

Thorium-230 1,678.3 105 5,600

Thorium-232 19.2 2.8 39

Uranium-234 289.75 242 376

Uranium-235 35 24 49

Uranium-238 621.5 548 670

Table 2-43. Burn Pit isotopic content. 

Radionuclides

Activity Concentration 

(pCi/gm of waste)

Avg Min Max

Cesium-137 -- -- --

Neptunium-237 0.27 0.1 0.6

Plutonium-238 0.13 0.1 0.5

Plutonium-239/240 0.11 0.1 0.4

Ruthenium-106 -- -- --

Strontium-90 -- -- --

Technetium-99 29.21 1.2 64

Thorium-228 7.63 1.5 19

Thorium-230 60.28 5.7 218

Thorium-232 7.18 1.1 21

Uranium-234 135.15 9.9 415

Uranium-235 8.37 0.5 27

Uranium-238 175.33 22 454

The airborne radon flux density for the Burn Pit is assumed to be less than 0.1 pCi/m2/sec, or in term 

of concentration, 3.13E-05 WL for dose reconstruction purposes.

The Burn Pit is within the beta/gamma direct radiation dose rate contour of 1.3 mRad/hr.

Underground Storage Tanks

There are 11 underground storage tanks on the FEMP. At present, Tank 8 is the only tank in use that 

contains a hazardous substance (soluble oil machine coolant). Tank 6 contains waste oil, but it is no 

longer in use.

Sanitary Landfill

The FEMP sanitary landfill is on a 3-acre tract in the northwest corner of the production area. The 

facility has 17 individual cells, five of which are full and out of service. The 12 remaining cells are 

awaiting issuance of an Ohio Environmental Protection Agency (OEPA) permit to install. Each cell is 
Effective Date: 05/20/2004 Revision No. 00 Document No. ORAUT-TKBS-0017-2 Page 68 of 83

estimated to provide approximately 2,000 cubic yards (cy) of gross disposal volume. Materials 

accepted at the facility include: nonburnable, nonradioactive sanitary wastes generated on the site (20 

cy/wk); nonradioactive construction-related rubble (variable quantity); and water treatment lime sludge 

(7 cy/wk). 

2.4 FEMP REMEDIATION AND ENVIRONMENTAL RESTORATION ACTIVITIES SINCE 

1989

The site production mission has been terminated and the site has been undergoing remediation and 

cleanup since 1989. The production level during the first half of 1989 was greatly reduced as 

compared to earlier years, and operations consisted mostly of processing waste residues. When 

production was suspended on July 10, most government production programs which used FEMP 

products were advised to find private-sector suppliers. The FEMP remains in standby for some 

segments of production until the government is confident commercial suppliers can meet future 

demand for uranium metal. Operations during the second half of 1989 were limited to containing and 

treating site effluents and packaging and shipping radioactive wastes for offsite disposal. Table 2-44 

provides an abbreviated history timeline and further presents a summary of the remediation activities 

at FEMP by the years since 1990.

Table 2-44. Abbreviated history timeline and remediation and environmental restoration 

activities since 1990. 

Year Activities

Abbreviated Timeline

1951 Construction of the Feed Materials Production Center began.

1952 Uranium production began (through 1989).

1986 EPA and DOE signed the Federal Facilities Compliance Agreement which

initiated the remedial investigation/feasibility study process.

1989 Uranium production was suspended. The Fernald site was placed on the 

National Priorities List, which is the list of CERLA sites most in need of cleanup.

1991 Uranium production formally ended. The site mission changed from uranium 

production to environmental remediation and site restoration.

As part of the Amended Consent Agreement, the site was divided into operable 

units for characterization and remedy determination.

1993 Environmental remediation activities were initiated at the FEMP.

1994 Environmental remediation activities under each of the operable unit’s records of 

decision were initiated.

1996 All five operable units had signed records of decision, signifying the end of the 10-

year remedial investigation/feasibility study process.

1997 Environmental remediation activities continued at the FEMP, including 

construction of Cell 1 of the on-site disposal facility with the first waste placement 

beginning in December. Remedy for Silos 1 and 2 was separated from the 

remedy for Silo 3.

1998 Decontamination of nuclear buildings and facilities (safe shutdown) neared 

completion, full-scale aquifer restoration was implemented, excavated soil 

volumes exceeded expectations, and cell construction at the on-site disposal 

facility continued.

1999 Excavation of the waste pits was initiated and 89,627 tons of waste was 

transported to Envirocare of Utah, Inc. Safe shutdown was completed ahead of 

schedule and 20 site structures were dismantled. Over 2 billion gallons of water 

were processed and 280,000 cubic yards of contaminated soil were excavated. 

The remedy for Silo 3 was selected.
Effective Date: 05/20/2004 Revision No. 00 Document No. ORAUT-TKBS-0017-2 Page 69 of 83

Year Activities

2000 The Record of Decision Amendment for Operable Unit 4 Silos 1 and 2 Remedial 

Actions was signed by EPA in July. On-site disposal facility Cell 1 was filled to 

capacity. Southern waste units’ excavation was completed to design grade.

2001 On-site disposal facility Cell 1 was capped. Remediation of the southern waste 

units was completed.

2002 The Silos 1 and 2 Radon Control System began operations and successfully 

reduced radon levels within the silos. The off-site transfer of nuclear product 

material was completed. The on-site disposal facility conducted waste placement 

into Cells 2, 3, 4, and 5.

Remediation and Environmental Restoration Activities since 1990

1990-1993 Handling and Storing Radioactive and Hazardous Materials

• Pitchblende ore residues containing radium stored in the K-65 Silos,

Radioactive Materials

• Thorium and thorium compounds stored in several location within the 

production area,

• Radioactive materials in the waste pits,

• Uranium metal,

• Uranium compounds, 

• Magnesium fluoride (MgF2) contaminated with uranium, and 

• Scrap metal contaminated with uranium compounds.

• Nitric acid,

Hazardous Materials

• Laboratory chemical,

• Hydrochloric acid,

• Sulfuric acid,

• Methanol, and 

• Process waste. 

FEMP was refurbishing and adding buildings to store hazardous waste, 

repackaging some materials into new drums, and removing materials no longer 

needed since production has ended. For example, two new warehouses 

originally built to store uranium products had been converted to meet the 

requirements for hazardous waste storage. Also, thorium previously stored in a 

deteriorating above-ground silo, in bins, and in drums on an outdoor pad were 

repackaged in new drums and stored in a warehouse.

1995 UNH Neutralization Project (approximately 200,000 gallons of Uranyl Nitrate 

Hexahydrate (UNH) were neutralized and prepared for disposal).

Vitrification Pilot Plant (construction continued)

Thorium Nitrate Stabilization (almost 6,000 gals of thorium nitrate were treated 

and solidified.)

Plant 4 Decontamination & Decomissioning (D&D)

1996 Thorium Overpacking Project (To overpack 5,600 deteriorated drums of thorium 

for safe transportation and permanent, off-site disposal).

Vitrification Pilot Plant (construction complete, full scale remediation would use 

vitrification to convert residues from silos 1 and 2 into a glass form. 

Approximately 36 tons of glass were produced during Phase I vitrification).
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Year Activities

Plant 4 Decontamination & Decommissioning (D&D) (The building was emptied, 

cleaned, and stripped down to its structural-steel frame work.

Plant 1 Decontamination & Decommissioning (D&D) (continued decontamination 

and dismantling activities).

Above-Grade Dismantlement of (Temporary) High and Low Nitrate Tanks

Advance Wastewater Treatment (AWWT) Slurry Dewatering Facility Online

Legacy Mixed Waste Shipped Off Site (shipment of 28,000 lbs of legacy mixed 

waste to Envirocare of Utah).

FEMP Continued Shipments of Uranium Metal Inventory (approximately half of 

the FEMP’s 32 million net pound inventory of uranium metal products had been 

removed).

Liquid Waste Project (completed shipping legacy liquid mixed waste to the Toxic 

Substance Control Act (TSCA) incinerator, located in Oak Ridge.)

1997-2002 Waste Pits Remedial Action Project (This project is responsible for the completion 

of remediation activities for the excavation, drying, loading, and rail transport of 

content of Waste Pits 1-6, the burn pit and the clearwell to an off-site disposal 

facility).

Soil Characterization and Excavation Project (This project is responsible for the 

completion of remediation activities to address contaminated soil at the FEMP 

and miscellaneous waste units.

On-Site Disposal Facility Project (This project is responsible for the construction 

of an eight-cell engineered disposal facility, and operation and maintenance of a 

leachate collection system).

Facilities Closure and Demolition Project (This project is responsible for the 

completion of decontamination and dismantling of the above-grade portion of the 

former uranium processing facilities and all remedial action facilities).

Silos Project (This project is responsible for the completion of remediation 

activities for the contents of K-65 Silos 1 and 2 and Silo 3, including the removal 

stabilization, and transport of the inventoried residues for off-site disposal).

Aquifer Restoration and Wastewater Project (This project is responsible for the 

completion of remediation activities necessary to restore the water quality in the 

affected portions of the Great Miami Aquifer).

2.5 FEMP RADIOACTIVE MATERIALS AND CHARACTERISTICS

The primary radiological materials at the FEMP are uranium and uranium compounds, recycle 

uranium, thorium and thorium compounds, and radon and thoron.
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2.5.1

Since 1951, uranium metal production has been the primary activity at the FEMP. Because of this, 

uranium has been widely distributed throughout the site and comprises the most likely source of 

external beta exposure and internal exposure at the site. Uranium received at the FEMP usually has 

been through one or more chemical separations at other sites. The separations remove most of the 

daughter products. Significant in-growth of daughters from purified uranium is limited to beta-gamma 

emitters 234Th and 234mPa, both of which are daughters of 238U. Those isotopes reach equilibrium 

within a few months after refining and are insignificant for radiological dose considerations when 

compared to the uranium isotopes present.

Uranium and Uranium Compounds

High quality uranium compounds were introduced into the FEMP processes at several points. Impure 

starting materials were dissolved in nitric acid and the uranium was extracted into an organic liquid. 

The uranium was then back-extracted into dilute nitric acid to yield a solution of uranyl nitrate.

Evaporation and heating were used to convert the nitrate solution to uranium trioxide (UO3 ) powder. 

This compound was reduced to uranium dioxide (UO2) with hydrogen and then converted to uranium 

tetrafluoride (UF4) by reaction with anhydrous hydrogen fluoride. Uranium metal was produced by 

reacting UF4 and magnesium metal in a refractory-lined reduction vessel. The primary uranium metal 

was remelted with scrap uranium metal to yield a purified uranium ingot that was shipped off the site 

for extrusion.

Based on the discussion of past FEMP operations, there are a variety of compounds of uranium that 

have been used and are present at the FEMP. For example, UO3 was the major refinery product at 

Plants 2/3, while the Plant 4 process started with UO3 and ended with a UF4 product. The production 

conducted in Plant 5 generated dusts of UF4 and uranium oxides (in MgF2). The principal uranium 

dust produced in the Plant 5 remelt area, Plant 9, and Plant 8 were uranium oxides that are generally 

considered to be U3O8. Various compounds were handled in Plant 1, Plant 8, and the Pilot Plant. 

Table 2-45 summarizes the uranium compounds in each plant.

During production, the FEMP processed uranium with enrichment levels that ranged from depleted to 

as high as 20%. In general, 2% enriched uranium was the highest enrichment that was processed in 

significant quantities. A history (1961-1984) of the average uranium enrichment in dust collector stack

discharges indicated 235U enrichment levels ranging from 0.2% to 1.68% with an average of 0.7%.

Table 2-45. Uranium compounds in 

former production areas. 

Plant Likely compounds

1 Mixed

2/3 Uranyl Nitrate, UO3, U-oxides

8 Mixed

Pilot Mixed

4 UO3, UF4

5 UF4, U-oxides, U-metal

6 U-metal, U-oxides

9 U-metal, U-oxides

Much of the onsite uranium has been stored in drums. These drums contain what are called “uranium 

residues.” These residues have been assayed by the analytical laboratory. The analyses have 

provided detailed information on the percent uranium in the drums (as well as the percentage of 
Effective Date: 05/20/2004 Revision No. 00 Document No. ORAUT-TKBS-0017-2 Page 72 of 83

enrichment of the material in the drums) that closely matches that reported for dust collector stack 

discharges. Several of the drums contain enrichment levels of approximately 3%, but they represent 

very small quantities of uranium. 

The specific activity of uranium increases with enrichment principally because of an increase in 234U 

activity. In the gaseous diffusion process, 234U is enriched to a greater extent than 235U. This effect 

on 234U enrichment must be considered when performing dose calculations for a particular degree of 

235U enrichment. Table 2-46 provides relative activity characteristics of depleted, natural, and 2% 

enriched uranium.

Table 2-46. Relative activity of uranium mixtures. 

nCi/mg of material

Mixture U-234 U-235 U-238 Total

Depleted 0.023 0.005 0.336 0.364

Natural 0.338 0.015 0.334 0.687

2% Enriched 0.801 0.043 0.329 1.174

In 1980, 89,000 lb of material identified as “Feed Plant Ash or Other Paducah Scrap” was shipped 

from the Paducah Gaseous Diffusion Plant to the FEMP. This material contained 22.5 metric tons of 

uranium and transuranic contamination that was 200 times higher than what was typical for recycled 

uranium. The material introduced plutonium, neptunium, and technetium to the process stream. The 

isotopic composition and related dosimetry of this material is discussed in the section on recycled 

uranium in this TBD.

A five-day field study was conducted by the Technical Safety Support Branch of EG&G Idaho, Inc. 

(Alvarez et al. 1984). The basic findings of this field study are presented in the following discussions.

The principal gamma emissions from depleted uranium included the 63- and 93-keV transitions that 

follow the beta decay of 234Th, and the 765- and 1,001-keV transitions associated with the isomeric 

state of 234Pa. Uranium X-rays are major contributors to the photon flux from depleted uranium. 

These X-rays arise from the removal of K and L electrons by both alpha and beta particles. Uranium 

X-ray emissions occur at 13.4, 13.6, 16.4, and 17.2 keV corresponding to the ULα2, and ULα1, ULβ2

and ULβ1 transitions. The uranium K series of X-ray emission occurs at 94.6, 98.4, 101.3 and 114.5 

keV, corresponding to the UKα2, UKα1, UKβ2, and UKβ1 transitions. Gamma emission at 185 keV 

accompanied by weaker lines at 143, 163, and 205 keV indicate the presence of 235U.

The beta spectrum of processed uranium metal and compounds tends to be dominated by the hard 

beta spectrum of 234Pa with an Emax=2.28 MeV (98%). The two betas from the decay of 234Th with 

Emax=0.191 MeV (79%) can be important contributors to the shallow skin dose at close range because 

intermediate beta energies deposit more energy per beta particle in the critical tissue depth defining 

skin dose.

The source materials presently on the site are no different than those present when the 1984 study 

was performed. Therefore, the same exposure sources are present but in a shielded state due to 

packaging in preparation for shipping or storage. The major change in the energy spectrum found 

would be a decrease in average energies due to shielding from packaging materials.

2.5.2

In 1985, a Department of Energy Task Force evaluated the processing of “recycle uranium” at several 

DOE facilities. Recycle uranium was defined as uranium that had been recovered from irradiated 

Recycle Uranium
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production reactor fuel. Recycle uranium is known to contain traces of transuranic (TRU) and fission 

product impurities. TRU impurities were usually limited to 237Np, 238Pu, and 239Pu, and the radioactivity 

of the TRU impurities was generally less than 0.1% of the total radioactivity of the recycle uranium. 

Recycle uranium also contains trace quantities of 241Am and 99Tc. 

From an internal dosimetry perspective, the most significant recycle uranium on the site is that which 

has been dubbed “POOS” for plutonium-out-of-specification. POOS uranium is that in which the total 

alpha activity from TRU elements exceeds 0.1% of the alpha activity from the uranium. This equates 

approximately to 11 parts of plutonium per billion parts of uranium on a mass basis.

Since the specific activity of 239Pu is 2.27 dps/nanogram, natural uranium containing one part per 

billion 239Pu has a corresponding activity concentration of 61.4 pCi 239Pu per gram of uranium. 

Therefore, uranium that exceeds the TRU specification of 11 ppb will have a 239Pu activity 

concentration of 675 pCi 239Pu per gram of uranium. Most POOS uranium at the FEMP contains less 

than 80 ppb 239Pu, which corresponds to an activity concentration of 4,910 pCi 239Pu per gram of 

uranium. A few containers at the FEMP have uranium residues containing TRU constituents 

significantly higher than 80 ppb.

2.5.3

Since 1972, the FEMP served as the thorium materials repository for DOE. Approximately two-thirds 

of the material in the repository was processed at FEMP. The remainder originated at other DOE 

facilities. Thorium was stored on the site in approximately 15,000 containers of various sizes. Many 

containers have been removed from the site during the remediation process. The remaining thorium 

(approximately 500,000 lbs) is stored in approximately 200 metal boxes. The thorium is primarily a 

mixture of thorium metal, thorium oxides, and process residues. The thorium metals, oxides, and 

residues are segregated and stored separately from uranium products. These storage areas have the 

highest gamma dose rates at the FEMP. Projects involving thorium material handling routinely have 

yielded the highest worker whole-body doses since 1990. Several operations involved the processing 

or handling of thorium. The major potential source of exposure to thorium material currently at the 

FEMP is the repacking and shipping of the remaining containers of thorium. 

Thorium and Thorium Compounds

In addition to the stored thorium, a number of locations on the site contain significant quantities of Th230. These sites includes the K-65 silos and the waste pits, and, to a lesser degree, various parts of 

Plant 2/3 and Plant 8. The 230Th at these locations resulted from the processing of uranium ores. 

Once the uranium had been extracted from the ore, the remaining material, known as raffinate, 

contained elevated levels of uranium decay products.

The thorium metal stored on the site is primarily 232Th. Because most of the material has been stored 

for some time, it can be assumed that it has a significant amount of daughter in-growth. The 228Th 

daughter of 232Th is an alpha-emitter nuclide with specific effective energy values comparable to those 

of the parent. For determining daughter isotope concentrations, the material is considered to be 35 

years old. 

The other thorium isotope of concern is 230Th. As previously discussed, 230Th is present as a residue

from the processing of ore. Much of the 230Th on the site is the result of processing pitchblende ore 

from the Belgian Congo.
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2.5.4

Past operations at the FEMP have resulted in enhanced radon and thoron levels. The site stores 

materials that are parent materials to both radon and thoron. At present, it stores approximately 8,800 

metric tons of uranium residues produced during plant operations since the1950s. During the period 

of production, uranium was chemically separated from pitchblende ore, and the waste residue, which 

contains high concentrations of 226Ra, the parent to radon, was stored in the K-65 silos. As 226Ra 

decays, it produces 222Rn, an inert gas, which can diffuse into the atmosphere through the silo cover. 

A sealant barrier of bentonite clay was applied to the contents of the silo in November 1991 to reduce 

radon emissions. As a result of past operations, certain areas of the site have been contaminated 

with 226Ra-bearing materials. This has resulted in elevated radon gas concentrations at several 

locations in the former process area.

Radon and Thoron

Rn-222, commonly referred to as thoron, is a radioactive gas resulting from the decay of 224Ra, which 

is in the thorium decay chain. The thorium metal and residues stored on the site are primarily 232Th. 

Because most of the material and residues have been stored for some time, there is a significant 

amount of 224Ra present, which results in the generation of thoron gas.

Radon and thoron data have been collected in specific radium- and thorium-bearing areas on the 

FEMP site to determine areas that have elevated concentrations. Available data were provided above 

in the descriptions of production plants.

Rn-222 has a half-life of 3.82 days. The decay of radon gas results in the production of four shortlived solid decay products: 218Po, 214Pb, 214Bi, and 214Po. The decay products build up rapidly with 

time and can approach secular equilibrium with the radon gas. Table 2-47 lists the progeny, decay 

modes, energies, and half-lives. The extremely short half-life of 214Po dictates that it is in equilibrium 

with 214Bi. The effective half-life of the radon progeny is approximately 40 minutes. Therefore, it takes 

approximately 3 hours for the radon progeny to come into equilibrium with the radon gas in a closed 

system. 

Table 2-47. Physical properties of radon and its progeny. 

Element Principle radiation Decay energies Half-life

Rn-222 α 5.5 MeV 3.82 days

Po-218 α 6.0 3.05 min.

Pb-214 Β,γ 1.0 (max) 26.8 min

Bi-214 Β,γ 3.3 (max) 19.7 min.

Po-214 α 7.7 164 μsec

In an open system, the progeny exist in atomic form for a period dependent on the availability of 

aerosol surfaces for attachment. The percent of radon progeny associated with airborne particulates 

is referred to as the attached fraction. The higher the concentration of ambient aerosols, the larger 

the attached fraction. The radioactive aerosol usually contains both attached and unattached 

progeny. Attached and unattached progeny are continually plating out onto floors, walls, etc., or being 

removed by filtration. As a result, radon progeny will not be in secular equilibrium. Typical ratios for 

activities of 222Rn/218Po/214Pb/214Bi (214Po) are 1.0/0.5/0.3/0.2 for indoors and 1.0/0.9/0.7/0.7 for 

outdoors.

Thoron (220Rn) has a half-life of 55.6 seconds. The decay of thoron gas produces four progeny: 216Po, 

212Pb, 212Bi, and 212Po. Table 2-48 lists the progeny, decay modes, energies, and half-lives. For 

thoron and its progeny, there is a greater degree of disequilibrium when compared to radon and its 

progeny. The longer half-life of 212Pb results in a greater opportunity for it to be lost from an air mass 
Effective Date: 05/20/2004 Revision No. 00 Document No. ORAUT-TKBS-0017-2 Page 75 of 83

by deposition on nearby surfaces. With the exception of the short-lived 216Po, thoron progeny, like 

radon progeny, are primarily in the attached state. This results in a significant increase in equilibrium 

when there is an increase in aerosol concentrations. The equilibrium factor has been reported to 

range between 0.002 and 0.06. At the FEMP, equilibrium values have been documented to range 

between 0.02 and 0.10 at several locations.

Table 2-48. Physical properties of thoron and its progeny. 

Element Principle radiation Decay energies Half-life

Rn-220 α 6.3 MeV 55.6 sec.

Po-216 α 6.8 0.15 sec.

Pb-212 β,γ 0.6 (max) 10.64 hours

Bi-212 β,γ 2.2 (max) 60.6 min.

Po-212 α 8.8 0.3 μsec

2.5.5

Table 2-49. contains specific activities and activity fractions for uranium, recycled uranium and 

thorium and conversion factor for radon concentration unit of Working Levels (WL).

Specific Activities and Conversion Factors 

Table 2-49. Conversion factors and isotopic and radionuclide activity fractions for recycled uranium

Specific Activities for Recycled Uranium:

Uranium 

Enrichmt

U+Cont.

(Bq/mg)

U Pu - 80ppb Np – 300 ppb Tc- 5000 ppb

Bq/mg f act. Bq/mg f act. Bq/mg f act. Bq/mg f act.

Depleted 4.53E+01 1.38E+01 0.305 1.83E-01 0.004 7.77E-03 1.72E-04 3.13E+00 0.69

Natural 5.76E+01 2.61E+01 0.453 1.83E-01 0.003 7.77E-03 1.35E-04 3.13E+00 0.54

2% Enrich 9.14E+01 6.03E+01 0.66 1.83E-01 0.002 7.77E-03 8.50E-05 3.13E+00 0.34

Specific Activities for Enriched Uranium:

Uranium Enrichment Isotopes Mass % f activities

Depleted U-234

U-235

U-238

05

0.25

99.75

0.083

0.014

0.903

Natural U-234

U-235

U-238

0.0057

0.7204

99.273

0.504

0.023

0.473

2% Enrich U-234

U-235

U-238

0.02

2.0

97.98

0.770

0.026

0.204

Specific Activities for Thorium:

Thorium-232 would be the default isotope. The degree of equilibrium is impossible to estimate, due to the variation in 

times since chemical separation of the feed stock. The specific activity value for 232Th and daughters is 1.11E-07 Ci/gm 

or 4.11Bq/mg.

Working Level (WL) Conversion Factor for Rn daughters:

1 WL = 100 pCi/L for 100% Rn daughter equilibrium

1 WLM (100% equilibrium) = 1.70E+04 pCi

No. of, Rn Concentration

WLM

Equalibriu Factionm Exposure Time

WLM × × =

100

0.1

, L

pCi Concentration eq

f

×

×

=

17000

2000

WLM = 0.118 (feq )(Rn concentration in pCi/L), equilibrium factor feq = 0.4(for indoor), 0.7(for outdoor) and 2,000 hrs 

exposure time.
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REFERENCES

Bryce L. Rich, et. al., June, 1988, Health Physics Manual of Good Practices for Uranium Facilities. 

EGG-2530, UC-41, Idaho National Engineering Laboratory.

ASI (Advanced Sciences, Inc.), date unknown, Remedial Investigation of the Feed Material 

Production Center Fernald, Ohio. Part 1: Evaluation of Current Situation, Feed Material 

Production Center, Fernald, Ohio.

Alvarez, J. L., S. H. Daniel, L. O. Johnson, D. E. Martz, B. L. Rich, 1984, Radiation Field 

Characterization and Radiological Equipment Evaluation at the Feed Material Production 

Center, HS-84-033, EG&G Idaho, Inc. for National Lead of Ohio, Inc., Fernald, Ohio.

Boback, M. W. and L. E. Green, April 2, 1981, External Radiation Exposure Analysis- Plant 9 

LeBlonde Rapid Bore, National Lead of Ohio, Fernald, Ohio.

Dolan, L.C., and C.A. Dolan, December 1988, History of FEMP Radionuclide Discharges, Addendum 

to FEMP-2082, Special UC-11, Feed Materials Production Center, Fernald, Ohio.

Robinson, S., 1997, FEMP External Dosimetry Technical Basis Manual, SD-2004, Rev.3, Fluor 

Fernald, Inc., Fernald, Ohio.

FEMP (Feed Materials Production Center), date unknown, Safety Analysis Report, Thorium 

Overpacking, Rev.0, Feed Materials Production Center, Fernald, Ohio.

FEMP (Feed Materials Production Center), 1988, A Closer Look at Uranium Metal Production, U.S. 

Department of Energy, Fernald, Ohio.

FEMP (Feed Materials Production Center), 1989, Final Safety Analysis Report for Overpacking 212 

Drums of Thorium, Westinghouse Materials Company of Ohio, Cincinnati, Fernald, Ohio.

FEMP (Feed Materials Production Center), 1994, Remedial Investigation Report for Operable Unit 1 

of the Fernald Environmental Management Project (FEMP), FEMP-OU01-6 FINAL, Fernald, 

Ohio.

FEMP (Feed Materials Production Center), 1998, Integrated Hazard Analysis for Operable Unit 4 

(OU4) Silos, 40000-RP-0027, Revision 2, Fernald, Ohio.

FEMP (Feed Materials Production Center), 2003, 2002 Site Environmental Report, 5150-RP-0022, 

Fluor Fernald, Fernald Closure Project, Fernald, Ohio.

Hill, C. A., Historical Process Descriptions, 1989, memorandum to T. R. Clark, PT: 89-108, 

Westinghouse Materials Company of Ohio, Fernald, Ohio.

Tomes, T., 1997, Technical Basis for Internal Dosimetry at the FEMP, SD-2008, Rev.1, Fluor Fernald, 

Inc., Fernald, Ohio.

A.S. Weaver, A. S., 987, Radon and Thoron Sampling Results, memorandum to H.D. Christiansen, 

WMCO:EH(HP): 87: 0112, Westinghouse Materials Company of Ohio, Fernald, Ohio.
Effective Date: 05/20/2004 Revision No. 00 Document No. ORAUT-TKBS-0017-2 Page 77 of 83

Westinghouse Materials Company of Ohio, Date unknown, Feed Materials Production Center, Annual 

Environmental Report for Calendar Year 1989, Westinghouse Materials Company of Ohio, 

Fernald, Ohio.

Westinghouse Materials Company of Ohio, Date unknown, Feed Materials Production Center, Annual 

Environmental Report for Calendar Year 1990, Westinghouse Materials Company of Ohio, 

Fernald, Ohio.

Westinghouse Materials Company of Ohio, December,1992, Feed Materials Production Center, 

Annual Environmental Report for Calendar Year 1991, Westinghouse Materials Company of 

Ohio, Fernald, Ohio.

Fernald Environment Restoration Management Corporation, June, 1993, 1992 Fernald Site 

Environmental Report, Fernald Environment Restoration Management Corporation, Fernald, 

Ohio.

Fernald Environment Restoration Management Corporation, June, 1994, 1993 Fernald Site 

Environmental Report, Fernald Environment Restoration Management Corporation, Fernald, 

Ohio.

Fernald Environment Restoration Management Corporation, July, 1995, 1994 Fernald Site 

Environmental Report, Fernald Environment Restoration Management Corporation, Fernald, 

Ohio.

Fernald Environment Restoration Management Corporation, June, 1996, 1995 Fernald Site 

Environmental Report, Fernald Environment Restoration Management Corporation, Fernald, 

Ohio.

Fluor Daniel Fernald (Fluor), June, 1997, 1996, Fernald Site Environmental Report, Fluor Daniel 

Fernald, Fernald, Ohio

Fluor Daniel Fernald (Fluor), June, 1998, 1997 Integrated Site Environmental Report, 51350-RP0001, Fluor Daniel Fernald, Fernald, Ohio.

Fluor Daniel Fernald (Fluor), June, 1999, 1998 Integrated Site Environmental Report, 51350-RP0005, Fluor Daniel Fernald, Fernald, OH.

Fluor Daniel Fernald (Fluor), June, 2000, 1999 Integrated Site Environmental Report, 51350-RP0010, Fluor Daniel Fernald, Fernald, OH.

Fluor Daniel Fernald (Fluor), June, 2001, 2000 Integrated Site Environmental Report, 51350-RP0015, Fluor Daniel Fernald, Fernald, OH.
Effective Date: 05/20/2004 Revision No. 00 Document No. ORAUT-TKBS-0017-2 Page 78 of 83

GLOSSARY

Atomic Energy Commission

Original agency established for nuclear weapons and power production; a predecessor to the 

U.S. Department of Energy.

beta radiation

Radiation consisting of charged particles of very small mass (i.e., the electron) emitted 

spontaneously from the nuclei of certain radioactive elements. Most (if not all) of the direct 

fission products emit beta radiation. Physically, the beta particle is identical to an electron 

moving at high velocity.

black oxide

Uranium oxide (U3O8) characterized by its olive green to black coloration.

brown oxide

Uranium dioxide (UO2) characterized by its brown coloration.

curie

A special unit of activity. One curie exactly equals 3.7 × 1010 nuclear transitions per second.

exposure

As used in the technical sense, exposure refers to a measure expressed in roentgens (R) of 

the ionization produced by photon radiation (i.e., gamma and X rays) in air. 

gamma rays (G or γ)

Electromagnetic radiation (photons) originating in atomic nuclei and accompanying many 

nuclear reactions (e.g., fission, radioactive decay, and neutron capture). 

green salt

Uranium tetrafluoride (UF4) characterized by its green coloration.

ionizing radiation

Electromagnetic or particulate radiation capable of producing charged particles through 

interactions with matter.

Jolter

The jolter is used to compact the magnesium fluoride slag lining in the metal reduction reactor 

pot by an up and down jolting action.

orange oxide

Uranium trioxide (UO3) characterized by its orange coloration.

photon

A unit or "particle" of electromagnetic radiation consisting of x- and/or gamma rays. 

Rad

The unit of absorbed dose. 

radiation

Alpha, beta, neutron, and photon radiation. 
Effective Date: 05/20/2004 Revision No. 00 Document No. ORAUT-TKBS-0017-2 Page 79 of 83

radioactivity

The spontaneous emission of radiation, generally alpha or beta particles, gamma rays, and 

neutrons from unstable nuclei.

rem

The rem is a unit of dose equivalent, which is equal to the product of the number of rad 

absorbed and the "quality factor."

Roentgen (R or r)

A unit of exposure to gamma or X-ray radiation. It is defined precisely as the quantity of 

gamma or X-ray radiation that will produce a total charge of 2.58 × 10-4 coulomb in 1 kg of dry 

air STP. An exposure of 1 R is approximately equivalent to an absorbed dose of 1 rad in soft 

tissue for higher (~>100 keV) energy photons.

shielding

Any material or obstruction that absorbs (or attenuates) radiation and thus tends to protect 

personnel or materials from radiation.

smoking area

An area designated for cigarette smoking.
Effective Date: 05/20/2004 Revision No. 00 Document No. ORAUT-TKBS-0017-2 Page 80 of 83

Appendix 2A. Default Model of FEMP Production Rate History for Dose Reconstruction Purposes. 

 

 

10,000 MTU (2 shifts @

100% system capacity)

1,230 MTU (1 shift @

24.6% capacity)

3,690 MTU (1 shift @

74% capacity

1960 1964 1968 1975 1981 1989

5,000 MTU (1 shift @

100% system capacity)

1970 1973

2,000

4,000

6,000

8,000

10,000

 Metric Ton Uranium

Years

1951 1953 1956

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

Plants 7, 9

Pilot Plant

9,000 MTU (2 shifts @

90% system capacity)

7,000 MTU (2 shifts @

70% system capacity)

3,000 MTU (1 shift @

60% system capacity)
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 Appendix 2B. FEMP uranium emissions summary by plant (kg). 

Year Plant 1 Plant 2/3 Plant 4 Plant 5 Plant 6 Plant 7 Plant 8 Plant 9

Pilot 

Plant Lab

Oil & 

graphite 

burner

Solid & 

liquid waste 

incineration

N.A.R

.

Cooling 

towers Year total

1951 0.0 0.0 0.0 0.0 0.0 0.0 0.0 0.0 123.0 0.0 0.0 0.0 0.0 2.9 125.9

1952 0.0 0.0 0.0 0.0 6.0 0.0 0.0 0.0 493.0 0.0 0.0 0.0 3.0 2.9 504.9

1953 3.8 6.0 1,473.0 90.0 14.1 0.0 0.0 0.0 493.0 1.9 0.0 0.0 3.0 2.9 2,087.7

1954 46.2 281.4 5,891.0 4,119.1 34.3 4,261.0 418.1 0.0 271.0 1.9 0.0 15 3.0 2.9 15,345.0

1955 46.2 1,115.1 12,452.0 10,410.4 65.3 7,268.0 1,825.4 0.0 443.0 1.9 0.0 118 3.0 2.9 33,751.2

1956 43.4 1,981.4 5,148.1 3,501.5 42.2 1,743.0 2,758.6 0.0 32.0 1.9 0.0 118 3.0 2.9 15,376.0

1957 49.4 3,735.4 819.7 3,664.8 51.8 0.0 2,366.7 0.4 18.0 1.9 0.0 118 3.0 2.9 10,814.0

1958 407.4 3,526.5 668.4 715.5 177.3 0.0 2,525.7 681.0 27.0 1.9 0.0 118 3.0 2.9 8,854.6

1959 46.0 3,936.4 1,433.8 478.9 143.1 0.0 2,360.9 420.5 34.0 1.9 0.0 118 3.0 2.9 8,979.4

1960 20.0 4,240.9 219.0 203.4 288.5 0.0 2,903.1 222.8 718.0 1.9 0.0 118 3.0 2.9 8,941.5

1961 52.8 3,714.1 268.5 76.7 136.0 0.0 2,581.0 74.0 174.0 1.9 0.0 118 3.0 2.9 7,202.9

1962 14.0 2,141.1 708.8 356.5 76.6 0.0 2,923.0 142.5 174.0 1.9 20 118 3.0 2.9 6,682.3

1963 82.6 0.0 1,475.4 783.6 199.7 0.0 3,165.9 169.2 51.8 1.9 27 118 0.0 2.9 6,078.0

1964 18.0 0.0 549.4 330.8 50.9 0.0 3,917.2 266.8 13.0 1.9 27 118 0.0 2.9 5,295.9

1965 4.1 193.1 338.9 226.9 59.0 0.0 6,200.7 83.02 10.0 1.9 28.2 118 0.3 2.9 7,267.0

1966 12.2 514.9 230.9 77.1 23.7 0.0 1,253.6 52.72 18.1 1.9 34.0 118 0.3 2.9 2,340.3

1967 20.4 648.0 284.16 148.3 14.3 0.0 2,207.6 80.64 11.8 1.9 34.0 118 0.3 2.9 3,572.3

1968 0.5 1,121.6 271.1 88.4 39.0 0.0 3,983.8 121.9 3.6 1.9 34.0 118 0.3 2.9 5,787.0

1969 27.2 699.5 52.0 119.6 9.83 0.0 3,547.4 13.63 3.6 1.9 34.0 94 0.3 2.9 4,605.9

1970 4.5 357.3 31.7 53.3 5.84 0.0 1,235.2 14.15 0.0 1.9 34.0 71 0.3 2.9 1,812.1

1971 9.0 306.5 0.89 0.1 2.7 0.0 632.1 0.65 0.0 1.9 34.0 71 0.3 2.9 1,062.0

1972 28.4 1,361.8 9.8 33.1 1.61 0.0 5.04 24.6 0.0 1.9 34.0 71 0.3 2.9 1,574.5

1973 1.0 1,398.3 58.2 79.1 3.03 0.0 53.02 15.87 0.0 1.9 34.0 71 0.3 2.9 1,718.6

1974 1.4 2,449.6 25.2 40.1 1.69 0.0 11.0 38.8 0.0 1.9 34.0 71 0.3 2.9 2,677.9

1975 5.6 2,850.0 120.9 19.1 1.25 0.0 3.51 0.68 0.4 1.9 34.0 71 0.3 2.9 3,111.5

1976 2.7 3,345.5 26.9 13.8 4.17 0.0 7.22 3.37 0.0 2.4 34.0 71 0.4 2.9 3,514.4

1977 0.6 757.61 130.0 53.4 1.89 0.0 4.73 0.61 10.4 1.9 34.0 71 0.3 2.9 1,069.3

1978 1.8 0.0 13.30 29.2 1.92 0.0 0.04 72.6 2.2 1.9 34.0 71 0.0 2.9 230.8

1979 0.8 0.0 47.23 12.4 1.5 0.0 0.06 2.84 0.0 1.9 22.0 71 0.0 2.9 162.6

1980 13.4 2.7 135.4 89.6 1.64 0.0 16.14 0.54 3.3 1.9 7.0 0.7 0.0 2.9 275.2

1981 1.3 30.1 424.7 135.7 2.10 0.0 10.01 0.75 0.0 1.9 7.0 1.2 0.3 2.9 618.0

1982 2.1 52.4 23.68 122.0 3.91 0.0 118.33 6.23 0.0 1.9 7.0 1.8 0.3 2.9 342.6

1983 6.4 130.2 45.81 41.6 3.87 0.0 82.8 1.43 0.0 1.9 2.4 5.4 0.3 2.9 325.0

1984 12.1 574.5 43.93 84.1 5.05 0.0 46.19 171.4 2.8 1.9 6.4 10.4 0.3 2.9 962.0

1985 130.1 3.33 0.22 3.42 0.0 0.06 1.12 1.9 0.3 2.9 218.72

1986 0.0 3.75 0.24 3.77 0.0 0.06 1.46 1.9 0.3 2.9 43.72

1987 200.1 3.29 0.18 2.09 0.0 0.37 0.36 1.9 0.3 2.9 246.92

1988 90.1 1.09 0.12 0.80 0.0 0.09 0.18 1.9 0.3 2.9 97.5

Total 985.3 41,892.21 33,433.26 26,198.86 1483.88 13,272 47,164.67 2686.75 3131 68.9 596 2490.5 39.4 110.2 173,675.16
Effective Date: 05/20/2004 Revision No. 00 Document No. ORAUT-TKBS-0017-2 Page 82 of 83

Appendix 2C. Fugitive uranium and thorium emissions from wind erosion of Waste Pits 1, 2 and 3. 

Year

Pit 1 Pit 2 Pit 3

Exposed 

pit area

(ft2)

Uranium 

emission

(kg/yr)

Thorium 

emission

(kg/yr)

Exposed 

pit area

(ft2)

Uranium 

emission

(kg/yr)

Thorium 

emission

(kg/yr)

Exposed 

pit area

(ft2)

Uranium 

emission

(kg/yr)

Thorium 

emission

(kg/yr)

1953 4,293 0.15 -- -- -- -- -- -- --

1954 6,868 0.25 -- -- -- -- -- -- --

1955 38,633 1.4 -- -- -- -- -- -- --

1956 55,803 2.0 -- -- -- -- -- -- --

1957 60,095 2.16 -- 17,781 46.4 0.014 -- -- --

1958 60,095 2.16 -- 35,562 92.7 0.032 -- -- --

1959 60,095 2.16 -- 35,562 92.7 0.032 0 0.0 0.0

1960 60,095 2.16 -- 40,008 104.3 0.032 0 0.0 0.0

1961 51,510 1.85 -- 44,453 115.9 0.036 0 0.0 0.0

1962 42,925 1.54 -- 44,453 115.9 0.036 44,431 0.63 0.0

1963 17,170 0.62 -- 44,453 115.9 0.036 44,431 0.63 0.0

1964 8,585 0.31 -- 44,453 115.9 0.036 236,966 3.36 0.01

1965 8,585 0.31 -- 22,227 57.95 0.018 266,587 3.78 0.014

1966 8,585 0.31 -- 13,336 34.8 0.01 266,587 3.78 0.014

1967 8,585 0.31 -- 0 0 0.0 0 0.0 0.0

1968 12,878 0.46 -- 0 0 0.0 0 0.0 0.0

1969 68,680 2.5 -- -- -- -- 103,673 1.47 0.005

1970 60,095 2.15 -- -- -- -- 177,725 2.52 0.01

1971 51,510 1.85 -- -- -- -- 177,725 2.52 0.01

1972 42,925 1.54 -- -- -- -- 88,862 1.26 0.005

1973 34,340 1.23 -- -- -- -- 88,862 1.26 0.005

1974 25,755 0.92 -- -- -- -- 29,621 0.42 0.0

1975 17,170 0.62 -- -- -- -- 236,966 3.35 0.01

1976 8,585 0.31 -- -- -- -- 236,966 3.35 0.01

1977 -- -- -- -- -- -- 236,966 3.35 0.01

1978 -- -- -- -- -- -- 207,346 2.94 0.01

1979 -- -- -- -- -- -- 177,725 2.52 0.01

1980 -- -- -- -- -- -- 133,294 1.89 0.005

1981 -- -- -- -- -- -- 103,673 1.47 0.005

1982 -- -- -- -- -- -- 44,431 0.63 0.0

1983 -- -- -- -- -- -- 0 0.0 0.0

1984 -- -- -- -- -- -- -- -- --

1985 -- -- -- -- -- -- -- -- --

1986 -- -- -- -- -- -- -- -- --

1987 -- -- -- -- -- -- -- -- --

1988 -- -- -- -- -- -- -- -- --

Total 29.2 -- 892.45 0.282 41.13 0.133
Effective Date: 05/20/2004 Revision No. 00 Document No. ORAUT-TKBS-0017-2 Page 83 of 83

Appendix 2C. (Continued) Fugitive uranium and thorium emissions from wind erosion of Waste Pits 

4, 5 and 6. 

Year

Pit 4 Pit 5 Pit 6

Exposed 

pit area

(ft2)

Uranium 

emission

(kg/yr)

Thorium 

emission

(kg/yr)

Exposed 

pit area

(ft2)

Uranium 

emission

(kg/yr)

Thorium 

emission

(kg/yr)

Exposed 

pit area

(ft2)

Uranium 

emission

(kg/yr)

Thorium 

emission

(kg/yr)

1953 -- -- -- -- -- -- -- -- --

1954 -- -- -- -- -- -- -- -- --

1955 -- -- -- -- -- -- -- -- --

1956 -- -- -- -- -- -- -- -- --

1957 -- -- -- -- -- -- -- -- --

1958 -- -- -- -- -- -- -- -- --

1959 -- -- -- -- -- -- -- -- --

1960 9,583 1.9 0.38 -- -- -- -- -- --

1961 19,166 3.79 0.76 -- -- -- -- -- --

1962 38,333 7.58 1.53 -- -- -- -- -- --

1963 43,124 8.52 1.72 -- -- -- -- -- --

1964 47,916 9.47 1.91 -- -- -- -- -- --

1965 47,916 9.47 1.91 -- -- -- -- -- --

1966 52,708 10.42 2.1 -- -- -- -- -- --

1967 52,708 10.42 2.1 -- -- -- -- -- --

1968 57,499 11.36 2.3 0 0.0 0.0 -- -- --

1969 57,499 11.36 2.3 0 0.0 0.0 -- -- --

1970 57,499 11.36 2.3 0 0.0 0.0 -- -- --

1971 57,499 11.36 2.3 0 0.0 0.0 -- -- --

1972 62,291 12.3 2.49 0 0.0 0.0 -- -- --

1973 62,291 12.3 2.49 0 0.0 0.0 -- -- --

1974 62,291 12.3 2.49 17,860 0.28 0.1 -- -- --

1975 67,082 13.26 2.68 44,649 0.71 0.24 -- -- --

1976 71,874 14.2 2.87 142,877 2.27 0.76 -- -- --

1977 81,457 16.1 3.25 53,579 0.85 0.29 -- -- --

1978 91,040 18.0 3.64 53,579 0.85 0.29 -- -- --

1979 95,832 18.94 3.83 53,579 0.85 0.29 3,240 8.26 --

1980 95,832 18.94 3.83 53,579 0.85 0.29 4,860 12.4 --

1981 95,832 18.94 3.83 53,579 0.85 0.29 8,100 20.7 --

1982 95,832 18.94 3.83 53,579 0.85 0.29 8,100 20.7 --

1983 95,832 18.94 3.83 53,579 0.85 0.29 8,100 20.7 --

1984 95,832 18.94 3.83 53,579 0.85 0.29 8,100 20.7 --

1985 95,832 18.94 3.83 53,579 0.85 0.29 8,100 20.7 --

1986 95,832 18.94 3.83 53,579 0.85 0.29 8,100 20.7 --

1987 95,832 18.94 3.83 89,296 1.42 0.48 8,100 20.7 --

1988 95,832 18.94 3.83 119,124 1.9 0.64 8,100 20.7 --

Total 394.9 79.8 15.1 5.1 186 --