ORAU Team

NIOSH Dose Reconstruction Project

Technical Basis Document for the Paducah Gaseous 

Diffusion Plant – Site Description

Document Number:

ORAUT-TKBS-0019-2 

Effective Date: 05/07/2004

Revision No.: 00

Controlled Copy No.:____ 

Page 1 of 23

Subject Expert: Baynard Turpin

Document Owner

Approval: Signature on File Date: 05/05/2004 Jay J. Maisler, TBD Team Leader

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

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

Approval: Signature on File Date: 05/07/2004 James W. Neton, OCAS Health Science Administrator

Supersedes:

None

TABLE OF CONTENTS

Section

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

Page

3

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

2.1 Introduction................................................................................................................................. 5

2.2 Site Description...........................................................................................................................5

2.3 Site Activities ..............................................................................................................................6

2.4 Site Processes............................................................................................................................9

2.4.1 Key Uranium Processing Facilities....................................................................................9

2.4.2 C-310 – Purge and Product Withdrawal Building ............................................................ 10

2.4.3 C-315 – Surge and Tails Withdrawal Building ................................................................. 11

2.4.4 Gaseous Diffusion Process Buildings (C-331, C-333, C-335, and C-337)....................... 11

2.4.5 Building C-340 – Metals Plant......................................................................................... 12

2.4.6 Building C-400 – Decontamination and Cleaning Building/Uranium Recovery ................ 14

2.4.7 C-404 – Solid Radioactive Waste Disposal Area............................................................. 16

2.4.8 Building C-409 – Stabilization Building............................................................................ 16

2.4.9 Building C-410 – Feed Plant ........................................................................................... 17

2.4.10 Building C-420 – Oxide Conversion Plant (UF4 – Green Salt Plant) ................................ 18

2.4.11 Building C-710 – Analytical Laboratories......................................................................... 18

2.4.12 Building C-720 – Maintenance Building........................................................................... 19

References ........................................................................................................................................21

Glossary ............................................................................................................................................22
Effective Date: 05/07/2004 Revision No. 00 Document No. ORAUT-TKBS-0019-2 Page 2 of 23

LIST OF TABLES

Table

2-1 Average external cumulative penetrating dose (deep dose equivalent) and 

number of workers assigned to each department during 1953-1988...........................................

Page

8

2-2 Average recorded external radiation doses (deep dose equivalent) per 

worker per year from 1953 to 1988 including maximum exposures recorded 

for any single worker for that year...............................................................................................8

LIST OF FIGURES

Figure

2-1 PGDP area map..........................................................................................................................

Page

5

2-2 PGDP site map ...........................................................................................................................6

2-3 PGDP uranium enrichment process............................................................................................7
Effective Date: 05/07/2004 Revision No. 00 Document No. ORAUT-TKBS-0019-2 Page 3 of 23

RECORD OF ISSUE/REVISIONS

ISSUE 

AUTHORIZATION 

DATE

EFFECTIVE 

DATE

REV. 

NO. DESCRIPTION

Draft 10/29/2003 00-A New technical basis document for the Paducah 

Gaseous Diffusion plant – Site Description. 

Initiated by Jay Maisler.

Draft 12/2/2003 00-B Adds Glossary, makes minor revisions to the 

report and resolves NIOSH comments. Initiated 

by Jay Maisler.

Draft 01/7/2004 00-C Resolves DOL comments. Initiated by Jay 

Maisler.

Draft 01/27/2004 00-D Resolves internal comments. Initiated by Jay 

Maisler.

Draft 04/19/2004 00-E Includes OGC paragraphs. Initiated by Jay 

Maisler.

Draft 05/05/2004 00-F Incorporates additional NIOSH comment. 

Initiated by Jay Maisler.

05/07/2004 05/07/2004 00 First approved issue. Initiated by Jay Maisler.
Effective Date: 05/07/2004 Revision No. 00 Document No. ORAUT-TKBS-0019-2 Page 4 of 23

ACRONYMS AND ABBREVIATIONS

AEC Atomic Energy Commission

ASTM American Society of Testing and Materials

CIP cascade improvement program

CUP cascade upgrade program

DOD U.S. Department of Defense

DOE U.S. Department of Energy

FP fission products

HF hydrofluoric acid

KOW Kentucky Ordnance Works

MAC Maximum Air Concentration

MgF2 magnesium fluoride

ORNL Oak Ridge National Laboratory

PGDP Paducah Gaseous Diffusion Plant

ppb parts per billion

ppm parts per million

RCRA Resource Conservation and Recovery Act

RU recycled uranium

TBD technical basis document 

Th thorium

TNT trinitrotoluene

TRU transuranic

TVA Tennessee Valley Authority

U uranium

UF4 uranium tetrafluoride (“green salt”)

UF6 uranium hexafluoride 

UO2 uranium dioxide

UO3 uranium trioxide (“yellow cake”)
Effective Date: 05/07/2004 Revision No. 00 Document No. ORAUT-TKBS-0019-2 Page 5 of 23

Figure 2-1. PGDP area map.

2.1 INTRODUCTION

This technical basis document (TBD), which is part of the Paducah Gaseous Diffusion Plant (PGDP) 

Site Profile, provides the site description (Section 2.2), routine site activities (Section 2.3), and site 

processes by specific building (Section 2.4) most relevant to worker dose reconstruction.

The purpose of the gaseous diffusion plant has been and continues to be the enrichment of uranium, 

initially for military applications and subsequently for commercial nuclear reactor fuel. PGDP enriches 

feed material in the form of uranium hexafluoride (UF6) gas from approximately 0.711% 235U up to 

about 2.5% 235U. The enriched product from PGDP was sent to other U. S. Department of Energy 

(DOE) gaseous diffusion plants at Portsmouth, Ohio, and Oak Ridge, Tennessee, for further 

enrichment. 

TBDs 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)).

2.2 SITE DESCRIPTION

PGDP was built on a portion of 16,126 acres of 

farmland acquired by the U.S. Department of 

Defense (DOD) during World War II. DOD 

acquired this land for a munitions facility, the 

Kentucky Ordnance Works (KOW), which was 

operated by Atlas Powder Company until it 

closed in 1946 (Lockheed Martin 1997a). The 

KOW included a trinitrotoluene (TNT) 

manufacturing area; an acid production area; 

coal, sulfur, toluene, and ordnance storage 

areas; a sewage treatment plant; a water 

treatment plant; and burning grounds. PGDP 

still uses the water treatment plant. In 1950, 

7,556 acres of the land east of the former KOW 

were acquired by the U.S. Atomic Energy 

Commission (AEC) as a site for a uranium enrichment facility that became PGDP (see Figure 2-1). 

The plant began operating in 1952, but construction was not complete until 1954. The facility 

reservation covered a total of 3,424 acres, with about 750 acres within the security fence. The rest of 

the land was transferred to the Tennessee Valley Authority (TVA) for the Shawnee Steam Plant and 

to the Commonwealth of Kentucky for wildlife conservation and recreational purposes (Lockheed 

Martin 1997b). The site consisted of 161 buildings, four of which contain the gaseous diffusion 

process (see Figure 2-2). 
Effective Date: 05/07/2004 Revision No. 00 Document No. ORAUT-TKBS-0019-2 Page 6 of 23

2.3 SITE ACTIVITIES

Originally, most UF6 feed material came from the depleted tails 

produced during normal diffusion operations at PGDP and 

from the Oak Ridge and Portsmouth gaseous diffusion plants. 

In addition, from 1953 through 1977, UF6 feed material was 

produced from uranium trioxide (UO3), called yellowcake, at 

PGDP in Buildings C-410 and C-420. The UO3 was supplied 

by sources such as El Dorado Mining and Refining, 

Mallinckrodt Chemical Works, and General Chemicals (now 

Allied Chemical), and comprised less than 10% of the UF6 fed 

to the cascade. From 1953 through 1964 and intermittently 

from 1968 through 1977, the Feed Plant also produced UF6

from UO3 from spent reactor fuel processed at the Hanford and 

Savannah River Sites. After 1977, all feed came in the form of 

UF6 from outside sources such as Oak Ridge, Portsmouth, and 

Allied Chemical (Bechtel Jacobs 2001).

Feed material was made from production reactor tails from 

1953 until 1964 and intermittently from 1968 to 1977. The 

percentage of PGDP cascade feed material from reactor tails averaged 17% during the periods this 

material was used, ranging from 3% in 1975 to 65% in 1973. Processing of UO3 to UF6 was 

accomplished in three steps: reduction, hydrofluorination, and fluorination. The uranium exposure 

pathway of greatest concern at PGDP was inhalation of uranium dust.

Chemical reduction involved transforming UO3 into uranium dioxide (UO2) using hydrogen gas. 

Hydrofluorination of UO2 into uranium tetrafluoride (UF4), commonly referred to as green salt, was 

accomplished by adding anhydrous hydrofluoric acid (HF). Fluorination was conducted in Building 

C-410 using heated elemental fluorine gas in tower reactors. The first two steps were performed on 

vibration tray reactors (shaker trays) from 1953 to 1956. In 1956, due to frequent equipment failures, 

spills, leaks, and the increased demand for feed, Building C-420, commonly called the green salt

plant, was completed and green salt production at Building C-410 was phased out. In Building C-420, 

reduction was performed in two-stage fluidized bed reductors; hydrofluorination was performed in 

three sets of horizontal screw reactors or in a two-stage fluidized bed hydrofluorinator. High Radiation 

Areas existed near the fluorination towers and ash receivers.

The main process buildings at PGDP (C-331, C-333, C-335, and C-337) contain the cascades, which 

are a series of compressor and converter stages and supporting equipment arranged in units and 

cells that progressively enrich UF6 in its gaseous form. Enrichment occurs as UF6 passes through 

barriers in the converters allowing isotopes of lower molecular weight to pass through. The series of 

converters results in two streams of UF6 – one of progressively higher percentage 235U that moves to 

the product withdrawal station in Building C-310, and one of progressively higher percentage 238U that 

moves toward the tails withdrawal station in Building C-315. Both the enriched product and the 

depleted tails are fed as liquid into cylinders and allowed to cool until solid. The enriched product is 

shipped to Portsmouth for further enrichment. The depleted material was either re-fed to the cascade 

or is stored on the site. Figure 2-3 shows the uranium enrichment process.

There were two cascade improvement/upgrade programs (CIP/CUP) at PGDP. The first of these ran 

from 1958 to 1962 and the second was from 1973 to 1981. These were significant because of 

possible worker exposure to transuranics while the cascade systems were open. Other major events 

were the closing of feed plant operations from 1965 to 1969 and in 1971. Feed plant operations, in 

Figure 2-2. PGDP site map.
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Buildings C-410 and C-420, and the decontamination building, C-400, were permanently shut down in 

the late 1970s.

In 1957, the presence of 237Np and 99Tc was documented at PGDP; between 1959 and 1966 

numerous studies related to the behavior, health effects, and controls for these radionuclides were 

conducted by PGDP and the AEC. The concentration of transuranic (TRU) elements, such as 237Np 

and 239Pu, and fission products (FP), such as 99Tc, in the reactor tails material was small, estimated at 

approximately 0.2 part per million (ppm) neptunium, 4 parts per billion (ppb) 239Pu, and 7 ppm 99Tc. In 

Figure 2-3. PGDP uranium enrichment process (Lockheed Martin 1997a). 

addition, 241Am, a decay product of 239Pu, builds up as 239Pu decays. These radionuclides were 

concentrated during processing at specific locations, which could increase radiation exposure to 

certain workers in these areas. (Bechtel Jacobs 2001).

Based on exposure records in the database, worker interviews, and health physics and inspection 

reports, an estimated 2,500 to 4,000 workers worked in areas with “moderate” to “high” potential for 

increased internal and external radiation exposures. This is based on a relative ranking of the 

potential of radiation exposures at PGDP. These areas included the feed plant (Buildings C-410 and -

420) with operators and mechanics receiving the highest doses, respectively, decontamination 

building (C-400) with decontamination workers receiving the highest doses, and the cascade buildings
Effective Date: 05/07/2004 Revision No. 00 Document No. ORAUT-TKBS-0019-2 Page 8 of 23

(C-331, C-333, C-335, and C-337) with operators receiving the highest doses. In addition, workers in 

Buildings C-340 and C-720 had a potential for increased external radiation exposures. Average

doses per department are listed in Table 2-1 (for 1953-1988) (PACE/University of Utah 2000).

Data indicate that approximately 200 individuals received more than 1 rem external exposure in any 

one calendar year. Health physics reports documented that many workers exceeded the weekly plant 

action level of 300 mrem/week, but suggested that workers were kept below the annual regulatory 

limits by rotation of duties and limitations on stay time in areas of higher exposure. Average and 

maximum doses are listed in Table 2-2 (PACE/University of Utah 2000).

Table 2-1. Average external cumulative penetrating dose (deep dose equivalent) and number of 

workers assigned to each department during 1953-1988. 

Dept. 

No.

Job 

description

Average 

cumulative 

dose 

(mrem)

Number 

of 

workers

5751 Feed plant operators 3,814 185

5760 Decontamination 2,788 116

5034 Feed plant mechanics 2,587 99

5015 Unknown 2,025 17

5676 Unknown 861 14

5730 Cascade operators 627 578

5785 Chemical operators 595 113

5075 Instrumentation 538 245

5020 Unknown 481 17

5008 Transportation pool 371 33

5002 Process maintenance 364 578

5108 Environmental control 338 48

5268 Unknown 316 236

5077 Electricians 298 318

5005 Material termination mgr. 295 90

5772 PEMU decontamination 253 22

5759 Unknown 220 4

Dept. 

No.

Job 

description

Average 

cumulative 

dose 

(mrem)

Number 

of 

workers

5049 Unknown 182 12

5725 Unknown 175 20

5044 Mech. inspection 170 113

5021 Plant services 147 486

5770 Converter test 145 23

5035 Feed plant mechanics 143 160

5019 Unknown 142 13

5740 Nitrogen plant 142 22

5646 Metals building 132 95

5674 Unknown 129 8

5048 Fabrication shops 127 667

5023 Unknown 115 24

5675 Unknown 114 7

5743 Steam plant 111 61

5027 Unknown 110 282

Total 4,706

Table 2-2. Average recorded external radiation doses (deep dose equivalent) per worker per year 

from 1953 to 1988 including maximum exposures recorded for any single worker for that year. 

Year

Average 

recorded dose 

(mrem)*

Maximum 

recorded 

dose (mrem)

Number 

of 

workers

1953 139.8 820 223

1954 232.5 1,580 284

1955 241.9 2,500 417

1956 358.6 4,700 471

1957 251.7 3,190 669

1958 185.3 3,630 661

1959 201.5 2,360 570

1960 201.1 2,510 526

1961 177.0 2,530 1,690

1962 149.5 2,980 1,479

1963 144.1 3,040 1,311

1964 73.4 1,860 1,219

1965 34.1 1,610 1,128

1966 31.7 1,470 1,138

1967 49.8 1,120 1,143

1968 61.8 1,400 1,241

Year

Average 

recorded dose 

(mrem)*

Maximum 

recorded 

dose (mrem)

Number 

of 

workers

1969 73.3 1,970 1,270

1970 41.7 840 1,273

1971 62.4 1,380 1,254

1972 58.9 1,760 1,288

1973 53.0 1,830 1,404

1974 26.5 1,030 1,624

1975 50.1 1,049 2,013

1976 35.1 1,224 2,426

1977 23.2 742 2,643

1978 39.9 359 2,613

1979 8.2 364 2,487

1980 18.2 344 2,308

1981 7.6 420 1,840

1982 6.5 350 1,617

1983 6.7 340 1,452

1984 9.2 420 1,434
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Year

Average 

recorded dose 

(mrem)*

Maximum 

recorded 

dose (mrem)

Number 

of 

workers

1985 6.1 350 1,365

1986 9.6 490 1,244

1987 8.0 470 1,275

Year

Average 

recorded dose 

(mrem)*

Maximum 

recorded 

dose (mrem)

Number 

of 

workers

1988 6.5 720 1,359

* The large number of zero values in the database reduces 

average values for radiation workers.

Approximately 10% of the 2,500 to 4,000 workers had a potential for higher than average radiation 

exposures. Reports that indicate extensive radioactive contamination in lunchroom areas, workers 

covered with black soot after ash-handling operations, Building C-410 floors routinely covered with 

visible green powder, and surveys indicating elevated concentrations of transuranics in almost all 

process buildings illustrate inadequacies in the PGDP contamination control program (PACE/

University of Utah 2000).

In general, job categories that had increased potential for exposures from external radiation were 

similar to those for internal radiation. The jobs and tasks that appear to involve the greatest potential 

for radiation exposure included ash handling (Building C-410), cylinder heel cleaning (Building C-400), 

derby processing (Building C-340), pulverizer operations (Building C-400), certain maintenance 

operations on the fluorination towers (Building C-410), maintenance on cascade equipment 

(cascades), cleaning of air filters (baghouses for Buildings C-400, C-410, C-420, and C-340), 

converter maintenance (Building C-720), flange grinding (Buildings C-340, C-400, C-410, and C-420), 

maintenance of the hydrogenation towers (Building C-340), and decontamination building cleaning 

operations (Building C-400). Jobs and tasks that were classified as having moderate potential 

radiation exposures include cascade operators and instrument mechanics (cascades), green salt 

sweeping (Buildings C-410 and C-420), disassembly of compressor (Building C-720), disassembly of 

block valves (Building C-720), drumming of green salt (Building C-340), and baghouse cleaning for 

cascades (Buildings C-310 and C-315). Other jobs and tasks were considered to have lower potential 

for worker radiation exposure. In addition, due to lack of aggressive contamination control programs, 

contamination from the operations identified above presented potential exposures to surrounding work 

areas. Further, legacy contamination generated from these operations could have posed a potential 

radiation exposure to individuals with access to the PGDP site in later years.

Special Incidences or Activities - High Potential for Increased Radiation Exposure

C-310 fire, 1956

C-337 fire, 1962

C-340, explosion and fire, 1962

Two workers overexposed, 1968 unknown location

C-315 fire, 1978

Cascade improvement program, 1954-1962 cascades

Cascade improvement program, 1973-1981 cascades

Neptunium production C-400

2.4 SITE PROCESSES

2.4.1

The major facilities at PGDP are:

Key Uranium Processing Facilities

• C-331, C-333, C-335, and C-337 – Gaseous Diffusion Process Buildings

• C-410/420 – UF6 Feed Plant (currently inactive)

• C-300 – Central Control Building
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• C-310 – Purge and Product Withdrawal Building

• C-315 – Surge and Tails Withdrawal Building

• C-340 – Metals Plant (currently inactive)

• C-400 – Decontamination and Cleaning Building

Other key facilities where moderate personnel radiation exposure occurred (e.g., C-404, C-409, 

C-710, and C-720) are included in this section.

NOTE: A complete list of radionuclides, quantities, chemical compounds, and inhalation class is not 

provided for every PGDP process and building discussed below. This Site Profile, in the TBD titled 

Occupational Internal Dose, provides a more complete description of this information. 

2.4.2

In Building C-310, enriched UF6 product was removed from the cascade and put in cylinders for 

transport. Building C-310 was approximately 53 x 30 feet in area and had two roll-up doors, one 

employee access door, and double doors to the storage room. The building was equipped to handle 

two 10- to 14-ton cylinders at a time.

C-310 – Purge and Product Withdrawal Building

Building C-310 received enriched UF6 gas from the cascades product via pumps that discharged 

through a condenser, piping, and cylinder pigtails to the intended receiving UF6 cylinder. Product 

cylinders were filled to no more than 95% (liquid) of capacity. 

The manual valves on the UF6 cylinder would occasionally be identified as defective and require 

replacement. This work was performed in Building C-310. According to procedures in effect in the 

1970s, a UF6 cylinder was required to cool at least 5 days before its valve was replaced. Those 

cylinders known to be above atmospheric pressure after the minimum cooling period would be vented

and further cooled, if necessary, with cold water. Until the mid-1970s, defective UF6 cylinder valves 

were routinely replaced by the mechanic standing upwind, with any escaping gases or fumes going 

the other way.

In 1980, the building was normally staffed by one to three workers with a crane operator on call if

cylinder transfer involving crane movements was required. The workers were responsible for 

completing equipment checks, logging equipment data, preparing cylinders for filling, disconnecting 

and weighing full cylinders, transferring cylinders, and maintaining cylinder records. In 1997, the 

workforce consisted of operators (three to seven per shift), maintenance mechanics (four per shift), 

instrument mechanics (two per shift), electricians (two per shift), and foremen (four per shift).

Operations Performed in Building C-310 and Maximum Radionuclide Concentrations

Expected: (Bechtel Jacobs 2001).

Connecting and disconnecting UF6 product cylinders and handling product UF6 – 1953-1999

27 ppb 237Np

0.06 ppb 239Pu

20,000 ppb 99Tc (UF6) 

Minimal external radiation exposure potential.

Changing/cleaning magnesium fluoride (MgF2) traps – 1964-1966

1900 ppb 237Np

0.01 ppb 239Pu

2,266,000 ppb 99Tc (MgF2) 

Moderate external radiation exposure potential. 
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2.4.3

Building C-315 began operation in early 1953. The building was used for the removal of depleted UF6

byproduct from a cascade and storage in cylinders. It was approximately 53 × 30 feet in area and 

contained four cart tracks and product equipment to accommodate four 10- to 14-ton cylinders. Four 

roll-up doors were in the east wall to permit the entry and exit of cylinders, while the west wall 

contained doors to the pump room and control room. Thus, six penetrations (doors) affected air flow 

in the building.

C-315 – Surge and Tails Withdrawal Building

Liquefaction was accomplished by compression of UF6 piped to the building from the enrichment 

operation (Buildings C-331, C-333, C-335, and C-337) at a pressure at which UF6 gas can be 

conveniently liquefied. After condensing, the liquid was allowed to flow conveniently into cylinders. 

The product was drained as a liquid into the multiton cylinders through a copper tube referred to as a 

pigtail. When the cylinder was filled to capacity, the cylinder and drain valves were closed and the 

pigtail was evacuated and purged. The pigtail was disconnected at the cylinder valve (PACE/

University of Utah 2000). 

In 1980, the building was normally staffed by one to three workers with a crane operator on call if

cylinder transfer involving crane movements was required. The workers were responsible for 

completing equipment checks, logging equipment data, preparing cylinders for filling, disconnecting 

and weighing full cylinders, transferring cylinders, and maintaining cylinder records. By 1997, the 

workforce consisted of operators (two or three per shift), maintenance mechanics (four per shift), 

electricians (two per shift), janitor (one per shift), and foremen (four per shift) (PACE/University of 

Utah 2000). 

Operations Performed in Building C-315 and Maximum Radionuclide Concentrations

Expected: (Bechtel Jacobs 2001).

Connecting and disconnecting UF6 tails cylinders and handling tails UF6 – 1953-1999

5 ppb 237Np, 

0.1 ppb 239Pu, 

20,000 ppb 99Tc (UF6) 

Minimal external radiation exposure potential. 

2.4.4

The main process buildings at PGDP contain the cascades, which are a series of compressor and 

converter stages and supporting equipment arranged in units and cells that progressively enrich UF6

in its gaseous form. The process buildings provide 74 acres under roof. Each process building has 

1,812 enrichment stages.

Gaseous Diffusion Process Buildings (C-331, C-333, C-335, and C-337)

PGDP construction, from 1951 through 1956, occurred in two phases. Construction of the first phase 

began January 2, 1951, and included Buildings C-331 and C-333. Authorization to proceed with the 

second phase of PGDP construction was received on July 15, 1952. Two additional enrichment 

facilities, Buildings C-335 and C-337, were added, and construction was completed in 1956. In 

September 1952, Buildings C-331 and C-333 began operation. In November 1952, the first product 

was withdrawn. In July 1953, the first reactor tails of UF6 were fed to the enrichment cascade after 

conversion in Oak Ridge. In April 1954, Building C-335 began operations; and in July 1954, Building 

C-337 began operations (Bechtel Jacobs 2001). 
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Operations Performed in the Cascade Buildings and Maximum Radionuclide Concentrations

Expected: (Bechtel Jacobs 2001).

Cascade operations – 1953-1964, 1969-1970, and 1972-1976

450 ppb 237Np

0.09 ppb 239Pu

23,000 ppb 99Tc

Moderate external radiation exposure potential. 

Cascade maintenance (cascade dust) – 1954-1961 and 1973-1981 (CIP/CUP)

2740 ppb 239Pu

3,220,000 ppb 237Np

Specific concentrations of 99Tc not available for this operation

High external radiation exposure potential. 

Isotopic concentrations in the reactor tails (approximate):

0.2 ppm 237Np

4 ppb 239Pu

7 ppm 99Tc

2.4.5

Several operations were performed in this building, two of which had a high potential for increased 

worker radiation exposure. These were the conversion of depleted UF6 to UF4 using a hydrogenation 

process and the conversion of some UF4 to uranium metal via a reaction with magnesium. The 

rationale for doing hydrogenation was to recover hydrofluoric acid for use in the oxide conversion 

process in Building C-420. Another reason might have been to convert UF6 to a form that was easier 

to store. Both processes generated considerable amounts of dust. 

Building C-340 – Metals Plant

Metals production involved several steps, each with its own unique hazards. 

• The uranium metal production process involved reducing UF6 (normally from the tails cylinder) 

to UF4 by combining it with hydrogen in a heated tower. The UF4 was mixed with magnesium 

and fed into lined firing reduction vessels, placed in furnaces, and heated until it fired into a 

metal ingot, called a derby. The next phase of the operation involved blending measured 

quantities of UF4 (depleted uranium) with measured quantities of magnesium metal, and 

pouring this mixture into the reactor liner. A refractory cap was then poured, and a lid was 

bolted to the top of the charged reactor. The charged reactor was transferred to an induction 

furnace where it was heated to the point where the uranium reduction started.

The primary hazard associated with this part of the process was exposure to airborne uranium 

dust during weighing, blending, and pouring. Respirators were required soon after initial 

production operations began. In addition, the reactors presented a significant hazard from 

burning magnesium and molten uranium metal. Phenomena described as "burnout" and "lid 

fires" occurred infrequently when the refractory liner was not correctly prepared. For example, 

burnouts occurred when burning magnesium came in contact with the steel shell, melting 

through the shell and releasing reactor contents to the furnace. Lid fires were similar, but 

occurred at the lid rather than the side of the shell. Such an occurrence led to a fatality (death 

was not due to radiological exposure) in March 1962. Burnouts resulted in significant 

contamination of the furnace refractory and would normally require relining of the entire 

furnace. 
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After the reactor was cooled, it was sent to the breakout area where the lid was removed, the 

shell was inverted, and the contents were dumped onto a grating, referred to as a grizzly. The 

slag material, at this point a hard ceramic, was broken into smaller pieces by beating it with a 

hammer. The pieces dropped through the grating into a jaw crusher and went to the slag 

plant. This operation was among the dirtiest jobs in Building C-340. Operators were 

sometimes covered with black dust; and, although respirators were required and generally 

worn, the extent of dust and contamination might have exceeded the protection factor of the 

respirator. 

• The metal ingot, or derby, was freed from the slag and roasted to oxidize the surface and 

loosen remaining slag. Loose oxides that fell from the derbies during roasting were collected, 

put in drums, and sent to a burial yard.

• After roasting, the derbies were cleaned by hand in a cleaning booth using power brushes and 

grinders to remove remaining slag. The potential hazards for airborne contamination for this 

operation were similar to those for the breakout operation.

• After cleaning, the derbies could be shipped directly or sawed into smaller shapes, depending 

on customer requirements. Derby sawing generated large amounts of uranium metal 

"sawdust," which burns readily in air. This sawdust was collected in drums of oil and kept 

covered. Despite these measures, uranium metal fires were common (daily or weekly), 

resulting in elevated levels of airborne uranium oxides.

• The Building C-340 operation was capable of remelting uranium derbies and casting specific 

shapes. Operations were conducted in a furnace with a controlled atmosphere. Graphite 

crucibles were used to receive molten uranium. The primary hazard associated with these 

operations was cleaning the crucibles between pours. Over time, oxides of uranium and betaemitting uranium decay products would impregnate a crucible. Because crucibles were 

cleaned by hand, operators received radiation doses to their hands, arms, and fingers.

• The MgF2 reaction product remaining in the reactor was captured, crushed, ball milled, and 

sized to be recycled as refractory. Although this was primarily a hands-off operation, it 

generated significant quantities of dust. Over time, the slag became contaminated with 

significant quantities of uranium oxides (several percent) that could have contributed to worker 

intakes. Reject slag (too small or too large) was collected in a hopper, and periodically 

drummed. Operators did not wear dosimetry that would have measured these extremity 

exposures (see Part 6, External Dosimetry, of this PGDP Site Profile). 

From 1978 to 1982, Building C-340 served as a shipping point for UF4. In addition, the building 

contained a remelt furnace for recasting uranium. The workforce consisted of operators (10 to 20 per 

shift), maintenance mechanics (three to five per shift), instrument mechanics (three to five per shift), 

and electricians (three to five per shift) (PACE/University of Utah 2000). After enrichment operations 

were shut down in the late 1970s, the building was used for other activities (e.g., offices and training 

programs). 

Operations Performed in Building C-340 and Maximum Radionuclide Concentrations

Expected: (Bechtel Jacobs 2001).

Production and handling of UF4 produced from tails – 1957-1962 and 1967-1977

62.6 ppb 237Np

0.063 ppb 239Pu

49 ppb 99Tc (UF4) 
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Minimal external exposure potential

Connecting and disconnecting UF6 tails cylinders and handling tails UF6 – 1953-1999

5 ppb 237Np

0.1 ppb 239Pu

20,000 ppb 99Tc (UF6) 

Minimal external exposure potential

Uranium-metal production from metal production, sawing, and metal handling – 1957-1962 and 

1967-1977

2.7 ppb 237Np, 

0.0085 ppb 239Pu,

10.6 ppb 99Tc (uranium metal) 

Minimal external exposure potential

Handling MgF2 in uranium-metal manufacturing – cleaning, roasting, liner preparation, knock out 

and drumming – 1957-1962 and 1968-1977

2,400 ppb 237Np, 

1.2 ppb 239Pu, 

1,880 ppb 99Tc (MgF2) 

High external exposure potential

2.4.6

Building C-400 was brought into service in April 1953. Its uranium recovery facilities were used to 

chemically separate and recover uranium from a variety of waste materials. Sources of feed material 

for this process included fluorination tower ash, sintered metal filters, decontamination solutions, UF6

scrubber solutions, particulates from ventilation filters and vacuum cleaners, laboratory wastes, and 

materials from spills. Building C-400 was a site for potentially increased radiation exposures, primarily 

because the following operations were performed there: converter disassembly, pulverization of waste 

UF4 and recycled UO3 containing transuranics, cylinder heel cleaning, spray booth operation, and 

237Np and 99Tc recovery. In addition, radiological hazards were associated with cleaning the building 

air filtering system (baghouses) (Bechtel Jacobs 2001). 

Building C-400 – Decontamination and Cleaning Building/Uranium Recovery

Before the mid-1970s, a complex uranium recovery process was performed in Building C-400. 

Uranium was separated from waste and scrap materials, concentrated, and converted to an oxide. 

The process included the following steps: dissolution of feed materials, filtration, solvent extraction in 

pulse columns, concentration by evaporation, and denitration to an oxide. The aqueous raffinate 

waste from solvent extraction columns, which contained 237Np, 239Pu, and 99Tc, was discharged to the 

environment (Bechtel Jacobs 2001). 

In the mid-1970s, the solvent extraction process for uranium recovery was replaced with a simpler 

precipitation and filtration process. Steps in this new process included dissolution of feed materials in 

nitric acid, addition of lime to precipitate uranium, and recovery of precipitated uranium as a filter 

cake.

Maintenance on major components in the cascade (compressors, converters, and process block 

valves) presented some of the most significant opportunities for exposure of maintenance personnel. 

Work on these components required that they be removed from the system, cleaned, rebuilt or 

repaired, and reinstalled. To remove these components, process operators isolated and bypassed 

the cascade cell containing the component, reduced the UF6 in the cell to less than 10 ppm equivalent 
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at atmospheric pressure, and purged the cell to minimize HF and UF6 exposure to workers involved in 

opening, maintaining, or modifying cell components. Once a satisfactory UF6 negative and HF purge 

were accomplished and the pressure of the isolated cell was raised to atmospheric pressure with dry 

air, the isolated cell was turned over to process maintenance for cell opening and disassembly. 

Compressors were transported from the process buildings to Building C-400 (and Building C-720). 

The compressors were disassembled into major components in pits, the parts were spray-washed to 

remove uranium deposits, and the rotor was relocated as required for deblading in Building C-400. 

The barriers were washed and disassembled, and scrap recovery was performed.

Neptunium Recovery

Soon after 237Np was identified at Paducah in 1957, the AEC placed a high emphasis on its recovery. 

A 237Np recovery process developed at ORNL began operation at PGDP in November 1958 in 

Building C-400. The process used a solvent extraction and evaporation method to recover and 

concentrate 237Np from receiver ash and cylinder heels wash solution. Receiver ash and solids that 

settled from cylinder wash water were dissolved in a nitric acid solution. Solids suspended in this 

solution were removed by filtration and discarded as solid waste. The filtrate was processed through 

solvent exchange pulse columns to separate uranium, thorium, and 237Np. These columns, originally 

in Building C-710, were moved to Building C-400 sometime after July 1959. Raffinate from these 

columns was dumped to the building drain if it contained uranium and 237Np concentrations less than 

500 ppm and 2.0 mg/L, respectively. Uranium and thorium were recovered for future use. The 237Np 

solution was concentrated to about 20 to 25 g/L by evaporation (Bechtel Jacobs 2001). 

The relatively greater radiological hazards associated with 237Np were understood at PGDP as early 

as 1959, and special practices for handling 237Np solutions and 237Np-contaminated equipment were 

recommended. Recommendations included using nonbreakable containers; maintaining tight 

systems; keeping lids on containers; preventing bubbling, frothing, or spraying of solutions; using 

rubber gloves; washing the gloves before using them in other areas; using respirators (or assault 

masks) for welding or burning; and performing alpha surveys of all equipment removed from 237Np 

processing areas (Bechtel Jacobs 2001). 

In 1958, ash and cylinder washings were processed through the Building C-400 237Np recovery 

process. The process used aqueous chemistry and ion exchange methods to recover 3.215 kg of

237Np from the cylinder wash stream and 1.074 kg of 237Np from the ash stream. The recovered 

materials were shipped to the Hanford Site. A program to recover 99Tc from cylinder wash water and 

raffinate (e.g., solvents) from 237Np recovery operations began in April 1960 (Bechtel Jacobs 2001).

Technetium Recovery

Technetium-99 is a fission product that PGDP received in recycled feed from the Hanford and 

Savannah River Sites. Technetium-99 passed through the cascade as a volatile compound of 

fluorine, depositing on internal surfaces of the cascade and contaminating the enriched uranium 

product. The AEC did not specify a limit for 99Tc in UF6 feed but indirectly controlled its concentration 

to about 10 ppm by limiting gross beta from fission products. A demand for 99Tc in the early 1960s 

prompted PGDP to begin a process to recover 25 kg of this material from various effluent streams. In 

1960, a process was begun to recover 99Tc from UF6 cylinder washwater and from raffinate generated 

during 237Np recovery. Process steps included precipitation and removal of uranium from these 

solutions by adding sodium hydroxide. This solution was processed through an ion exchange column 

and elutriated with nitric acid to produce a concentrated solution of 99Tc that was shipped to Oak 

Ridge National Laboratory (ORNL). Although the contribution to radiation dose from 99Tc was not of 

concern during most PGDP operation and maintenance activities, the isolated activity required 
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specific monitoring considerations for both internal exposure and external skin exposure (Bechtel 

Jacobs 2001). 

Cylinder Cleaning

With repeated reuse, UF6 cylinders collected deposits that did not completely volatilize in the 

autoclave. Periodically these deposits, called cylinder heels, had to be dissolved and removed, and 

the cylinder was then cleaned, refurbished as necessary, reinspected, hydrostatically tested, and 

weighed for subsequent use. Cylinder heels were composed of corrosion products, uranium salts and 

oxides, and TRU and uranium daughter product compounds. With regard to the contaminants of the 

process gas, some of the 237Np and much of the 99Tc was volatilized to the cascade, while most 239Pu 

remained behind in the cylinder heels, creating a significant radiological hazard. Cylinder cleaning 

was performed at Building C-400, where the heels were dissolved and the rinse water was collected 

in a large pan (Bechtel Jacobs 2001). 

Operations Performed in Building C-400 and Maximum Radionuclide Concentrations

Expected: (Bechtel Jacobs 2001).

Cascade maintenance - 1954-1961 and 1973-1981 (CIP/CUP)

450 ppb 237Np

0.09 ppb 239Pu

Specific concentrations of 99Tc not available for this operation

High external radiation exposure potential. 

Uranium/neptunium recovery - 1957-1990 and 1956-1976

25,000,000 ppb 237Np

100,000 ppb 239Pu

Specific concentrations of 99Tc not available for this operation (solutions) 

Moderate external radiation exposure potential. 

2.4.7

This area was the primary disposal site for radioactive waste at PGDP. It was constructed as a 

holding pond for Building C-400 liquid waste. The pond was 380 by 140 feet, with 6-foot-high dikes. 

In 1957, the C-404 Holding Pond was converted to a solid radioactive waste burial area. By 1977, 

approximately 6,400,000 pounds of materials contaminated with uranium had been drummed and 

placed in the holding area. Waste streams included incinerator ash, contaminated alumina, highly 

contaminated roofing waste, and gold recovery sludge. This area continued in use into the mid1980s. It was later determined to contain sludge that was chemically hazardous, thus requiring 

closure under the Resource Conservation and Recovery Act (RCRA) in 1987 (Bechtel Jacobs 2001). 

C-404 – Solid Radioactive Waste Disposal Area

Reported activities or concentrations for other radionuclides were not found for this area. 

2.4.8

Cascade Maintenance

Building C-409 – Stabilization Building

Once compressors had been overhauled and reassembled in Building C-400 or C-720, compressor 

openings were covered for transportation to storage or reinstallation. Converters were transported 

from the process buildings to Building C-409 for decontamination.

During the 1973-1981 cascade improvement, a shop for reassembling and testing converters was 

located in Building C-409. In addition, the building had a small spray booth for minor cleaning jobs. 

The number of personnel involved is not known. The spray booth in C-409 was used in the initial 
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stages of tearing down converters. The waste from this spray booth was piped to C-400 for removal 

of uranium oxides. It is not clear, but apparently the C-409 spray booth was operated by C-400 

personnel.

Operations Performed in Building C-409 and Maximum Radionuclide Concentrations

Expected: (Bechtel Jacobs 2001).

Cascade maintenance – 1954-1961 and 1973-1981 (CIP/CUP)

450 ppb 237Np

0.09 ppb 239Pu

Specific concentrations of 99Tc not available for this operation

High external radiation exposure potential. 

2.4.9

From July 1953 through 1977, UF6 feed material was produced from UO3 (yellowcake) at PGDP in 

Buildings C-410 and C-420. This feed material was supplied by sources such as El Dorado Mining 

and Refining, Mallinckrodt Chemical Works, and Allied Chemical and comprised less than 10% of the 

UF6 fed to the cascade. From 1953 through 1964 and intermittently from 1968 through 1977, the 

Feed Plant produced UF6 from UO3 obtained from spent reactor fuel processed at the Hanford and 

Savannah River Sites. After 1977, all feed came in the form of UF6 from outside sources such as Oak 

Ridge, Portsmouth, and Allied Chemical.

Building C-410 – Feed Plant

The next stage of the enrichment operation was to convert solid UF4 to gaseous UF6. This was done 

using heated elemental fluorine gas in the Building C-410 fluoridation towers. The operation 

consisted of introducing UF4 at the top of the tower while fluorine gas was introduced from below. The 

resulting UF6 gas/liquid was removed in large cylinders and the solid waste products were collected in 

ash receivers at the bottom. The external radiological concerns were from beta and gamma 

emissions from transuranics, fission products, and accumulated uranium daughter products, which 

were concentrated by the process in the ash at the bottom of the fluoridation towers. In addition to 

external radiation sources, inhalation of radioactive dust while cleaning plugged equipment, changing 

out the ash receivers, and cleaning of the building air filters was possible. The potential for radiation 

exposure was particularly increased for work around and with ash receivers. Workers in the 

fluorination tower area were exposed to UF4 and UO2F2 (PACE/University of Utah 2000). 

Exposure to uranium powder dusts was prevalent in operations and maintenance activities. For 

example, plugging of conveyers, hoppers, and screws with UO3 or UF4 routinely required physical 

agitation with sledgehammers or metal rods. In many cases, shear pins or chains on the associated 

drive mechanisms broke, requiring operations personnel to clean the product out of the jammed 

equipment and maintenance personnel to disassemble and repair the equipment. 

Operations Performed in Building C-410 and Maximum Radionuclide Concentrations

Expected: (Bechtel Jacobs 2001).

Cascade operations – 1953-1964, 1969-1970, and 1972-1976

450 ppb 237Np

0.09 ppb 239Pu

23,000 ppb 99Tc

Moderate external radiation exposure potential.

Feed plant operations, recycled uranium (RU) to UF6 - UO3 → UO2 UF4 UF (handling, drumming, 

bag changing, etc. (tower ash) – 1953-1964 and 1969-1977

25,602 ppb 237Np
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8,000 ppb 239Pu

4,600 ppb 99Tc

High external radiation exposure potential

Changing/cleaning magnesium fluoride (MgF2) traps – 1964-1966

1,900 ppb Np

0.01 ppb Pu

2,266,000 ppb Tc (MgF2) 

Moderate external radiation exposure potential.

2.4.10

In August 1956, the Building C-420 expansion to the feed plant was complete. Building C-420 was a 

comparatively small building attached to the west side of Building C-410. It contained fluidizing beds 

that were used in the conversion processes. This was where U3O8 was converted to uranium oxide 

and then to UF4 for use as feed stock for the fluorination towers in Building C-410. The ore was run 

through two sets of fluidized beds, the first of which converted UO3 (yellow powder) to UO2 (black 

powder), while the second converted UO2 to UF4 by reaction with hydrofluoric acid. The primary 

potential for radiation exposure to operators was the inhalation of dust generated while unplugging 

converters or while cleaning the building air filtering system (baghouse). Maintenance mechanics had 

a potential for inhalation exposure while working on the equipment.

Building C-420 – Oxide Conversion Plant (UF4 – Green Salt Plant)

The equipment used for the conversion processes consisted of a series of hoppers, conveyer belts, 

screws, chutes, etc., which were susceptible to mechanical failure. When a failure occurred, the 

system would be opened and the operators and maintenance mechanics would do what was 

necessary to get things going again. Routine operations in Building C-420 do not appear to have had 

the potential for increased radiation exposure; however, the workers in this building were rotated with 

those in Building C-410 to reduce individual radiation exposure.

The workforce consisted of operators (four per shift), maintenance mechanics (two per shift), 

electricians (two per shift), instrument mechanics (two per shift), and janitors (one per shift) (PACE/

University of Utah 2000).

Operations Performed in Building C-420 and Maximum Radionuclide Concentrations

Expected: (Bechtel Jacobs 2001).

Feed plant operations, RU to UF6 - UO3 → UO2 UF4 UF (handling, drumming, bag changing, etc. 

(tower ash) - 1953-1964, 1969-1977, and 1982-1983

25,602 ppb 237Np

8,000 ppb 239Pu

4,600 ppb 99Tc

High external radiation exposure potential

2.4.11

The PGDP analytical laboratories are in Building C-710. The laboratories, which have been in 

operation since plant startup, consist of the ASTM, Industrial Hygiene, Infra-Red Spectrometry, 

Emission Spectrometry, Sampling, Uranium Analysis, Trouble Shooting, Quality Control, 

Radiochemistry, Metallurgy, Mass Spectrometry, Counting Preparation, Alpha and Beta Counting, and 

Fission Training Laboratories.

Building C-710 – Analytical Laboratories
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The 99Tc recovery operation, initially in Building C-710, was moved to Building C-400 sometime after 

1959.

A 237Np recovery process originally in Building C-710, Room 32, was transferred to Building C-400 

sometime after July 1959 (see Section 2.4.6). After processing in Building C-400, the concentrate 

was sent to a laboratory in Building 710 for additional separation and concentration in ion exchange 

columns. The final product was siphoned into glass carboys on the loading dock at Building C-710 

and shipped to ORNL.

A Health Physics and Industrial Hygiene Department report for the first quarter of 1959 stated that 

continuous air samples collected near the 237Np recovery operation in Building C-710 averaged 

slightly above the maximum air concentration (MAC, a term used for airborne radioactivity 

concentrations in 1959) assumed for 237Np. Later analysis indicated that 29% of the alpha activity 

was attributable to 237Np. There is no indication that respiratory protection was used during these 

activities. Urine samples that were collected and sent to ORNL for analysis tested positive for 237Np

(PACE/University of Utah 2000).

Operations Performed in Building C-710 and Maximum Radionuclide Concentrations

Expected: (Bechtel Jacobs 2001).

Uranium/neptunium recovery - 1952-1990 and 1956-1976

25,000,000 ppb 237Np

100,000 ppb 239Pu

Specific concentrations of 99Tc not available for this operation (solutions)

Moderate external radiation exposure potential. 

2.4.12

Building C-720 contained the compressor, converter, and machine shops. Machinists, maintenance 

mechanics, instrument mechanics, sheet metal workers, electricians, and inspection workers 

performed the following functions: disassembly of compressors, converter maintenance, disassembly 

of block valves, machining, fabrication, welding, and grinding. In addition, stores workers and janitors 

were assigned to the building. Supervisory offices were in the middle of the building.

Building C-720 – Maintenance Building

The primary structure of interest from the standpoint of radiation safety was the compressor 

disassembly area. This was in a pit at one end of the building and was several stories high. 

Occasional releases of UF6 occurred during compressor disassembly.

Compressors were transported from the process buildings to Building C-720. The compressors were 

disassembled into major components in pits, the parts were transported to Building C-400 for spray 

washing to remove uranium deposits, and the rotor and stator were relocated as required for 

deblading in Buildings C-400 and C-410, respectively. Reusable washed parts were returned to their 

respective maintenance buildings for modification, refurbishment, degreasing, and reassembly. 

Following washing in Building C-400, the converters were modified, refurbished, and reassembled in 

Building C-720. Prior to removal from the system, block valves were slightly opened (where possible), 

inspected, cut out of the system, lifted free of process piping, decontaminated, covers installed, and 

shipped to Building C-400 for preliminary disassembly and decontamination to the limits allowed in 

Building C-720. Once decontaminated, the valve was covered and transported to Building C-720 for 

final repair and reassembly, and staged in the process building for reinstallation.
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Support operations involved the following exposure pathways:

• Cleaning of cylinder heels (potentially involving feed, product, or tails cylinders)

• Decontamination of equipment associated with feed, cascade, and other operations

• Routine and emergency maintenance operations at ancillary support facilities

• Uranium recovery from oils, cleaning solutions, and other wastes

• Effluent, sludge, and other wastes from decontamination processes

• Incineration of certain wastes

• Scrap from equipment

• Removal or drainage of sludge from waste ponds

• Analytical laboratory sampling

Operations Performed in Building C-720 and Maximum Radionuclide Concentrations

Expected: (Bechtel Jacobs 2001).

Cascade maintenance -- 1954-1961 and 1973-1981 (CIP/CUP)

450 ppb 237Np

0.09 ppb 239Pu

Specific concentrations of 99Tc not available for this operation

High external radiation exposure potential. 
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REFERENCES

Bechtel Jacobs (Bechtel Jacobs Company, LLC), 2001 DRAFT, Recycled Uranium Mass Balance 

Project Paducah Gaseous Diffusion Plant Site Report, BJC/PGDP-167, prepared for the DOE

Office of Environmental Management. 

Lockheed Martin (Lockheed Martin Energy Systems, Inc.), 1997a, Paducah Site Annual 

Environmental Report for 1996, Document No. KY/EM-206, prepared for the U.S. Department 

of Energy, Kevil, Kentucky, December.

Lockheed Martin (Lockheed Martin Energy Systems, Inc.), 1997b, Paducah Site Annual 

Environmental Report for 1995, Document No. KY/EM-176, prepared for the U.S. Department 

of Energy, Kevil, Kentucky, January.

PACE/University of Utah, 2000, Exposure Assessment Project at the Paducah Gaseous Diffusion 

Plant, submitted by Paper, Allied Industrial, Chemical and Energy Workers (PACE) 

International Union and University of Utah, Division of Radiobiology, Center for Advanced 

Medical Technologies, Center of Excellence in Nuclear Technology, Engineering and 

Research. 
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GLOSSARY

cascade

A series of compressor, heat exchanger, control valve and motor, converter stages, and 

supporting piping arranged in stages, cells, and units that progressively increase the 

concentration of 235U isotope in the uranium hexafluoride (UF6) feed. Enrichment occurs as 

UF6 passes through semiporous barriers in the converter stage. These barriers allow the 

lighter 235U molecules to pass through more easily, which results in a gas with a slightly higher 

percentage of 235U (enriched) on one side of the barrier and a slightly lower percentage 

(depleted) gas on the other side. The enriched UF6 gas flows toward the top of the cascade 

while the depleted UF6 gas travels toward the bottom of the cascade.

contamination control program

A system of controls used to reduce and confine radioactive contamination. These systems 

include contamination surveys; posting; leak prevention, detection (surveillance), and timely 

repair; and use of containment devices.

derby

A molded uranium metal ingot made from UF4 and magnesium.

dosimetry

The science of assessing absorbed dose, dose equivalent, effective dose equivalent, etc., 

from external and/or internal sources of radiation.

enrichment

Process that occurs as UF6 passes through barriers in converters allowing isotopes of lower 

molecular weight to pass through.

enriched uranium

Uranium that has a higher concentration of 235U. A 2 to 4% enrichment is used in commercial 

power reactors. Greater than 20% enrichment is considered highly enriched uranium (HEU). 

external radiation exposures

Exposure to a penetrating radiation source(s) that is external to the body.

fission products

Isotopes formed during fission of uranium or plutonium. 

High Radiation Areas

Areas where the radiation dose rate could exceed 100 millirem in 1 hour at 30 centimeters 

from the radiation source.

inhalation class

The respiratory tract inhalation classification scheme developed in ICRP Report 30 for inhaled

material according to its rate of clearance from the pulmonary region of the respiratory tract. 

Materials are classified as D (days), W (weeks), or Y (years), according to how fast they clear

the lungs; Class D in less than 10 days; Class W in 10 to 100 days; Class Y in more than 100

days. Recent recommendations in ICRP Report 66 replaced classes D, W, and Y with lung

absorption Types F (fast), M (moderate), and S (slow).
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internal radiation exposures

Radiation exposure absorbed by the body resulting from radioactive material taken into the 

body.

liquefaction

The process of causing a gas or solid to become a liquid, usually by condensing the gas or 

melting the solid.

natural uranium

Uranium is a naturally occurring radioactive element consisting of three isotopes: 238U 

(99.276%), 235U (0.719%), and a trace amount of 234U (0.0057%). 

pigtail

A flexible connecting device.

spent reactor fuel

Reactor fuel that has been used in a reactor to the point that the amount of fissionable 

uranium (235U) has been depleted.

transuranic (TRU) elements

Elements with an atomic number greater than uranium (92). 

weekly plant action level

A radiation exposure value/airborne radioactivity value that, when reached by an individual, 

would result in some compensatory action being taken for that individual, such as reassigning 

the individual to a work detail involving no or reduced radiation exposure/airborne radioactivity.