ORAU TEAM

Dose Reconstruction

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Oak Ridge Associated Universities I Dade Moeller I MJW Technical Services

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DOE Review Release 08/31/2012

Document Title:

Paducah Gaseous Diffusion Plant – Site 

Description

Document Number: ORAUT-TKBS-0019-2

Revision: 03

Effective Date: 08/22/2012

Type of Document: TBD

Supersedes: Revision 02

Subject Expert(s): Jodie L. Phillips

Approval: Signature on File Approval Date: 08/20/2012

Jodie L. Phillips, Document Owner

Concurrence: Signature on File Concurrence Date: 08/21/2012

John M. Byrne, Objective 1 Manager

Concurrence: Signature on File Concurrence Date: 08/20/2012

Edward F. Maher, Objective 3 Manager

Concurrence: Vickie S. Short Signature on File for Concurrence Date: 08/20/2012

Kate Kimpan, Project Director

Approval: Signature on File Approval Date: 08/22/2012

James W. Neton, Associate Director for Science

New Total Rewrite Revision Page Change

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

REVISION AND DISCARD / DESTROY ALL COPIES OF THE PRIOR REVISION.
Document No. ORAUT-TKBS-0019-2 Revision No. 03 Effective Date: 08/22/2012 Page 2 of 24

PUBLICATION RECORD

EFFECTIVE 

DATE

REVISION

NUMBER DESCRIPTION

05/07/2004 00 New technical basis document for the Paducah Gaseous Diffusion 

plant – Site Description. First approved issue. Initiated by Jay J. 

Maisler.

08/21/2006 01 This revision addresses comment #2 from the meeting with United 

Steelworkers Local 5-550 (formerly PACE) and SPFPA Local 111 

held on 02/10/2005. Approved issue of Revision 01. This revision 

results in no change to the assigned dose and no PER is required. 

Training required: As determined by the Task Manager. Initiated by 

Paul A. Szalinski.

05/08/2007 02 Approved Revision 02 initiated to incorporate Attributions and 

Annotations section. No changes occurred as a result of formal 

internal review. Incorporates NIOSH formal review comments. 

Constitutes a total rewrite of the document. This revision results in 

no change to the assigned dose and no PER is required. Training 

required: As determined by the Task Manager. Initiated by Daniel S. 

Mantooth.

08/22/2012 03 This revision initiated in response to Advisory Board Work group. 

Corrected references to yellowcake throughout the document. 

Section 2.1.2, added SEC discussion. Section 2.2, added brief 

overview of current site status. Revised Figure 2.2. Removed 

Tables 2-1 and 2-4, necessary information is provided in Sections

2.4.1 through 2.4.12. Removed Tables 2-2 and 2-3 because they 

were not used to perform dose reconstruction. Clarified CIP/CUP 

campaign dates throughout document. Section 2.3, paragraph 5, 

added - Activities in support of the cascade as well as other plant 

support activities continue in Building C-400. Section 2.4.1 and 

2.4.2, changed 1999 to present. Section 2.4.4, paragraph 3, 

changed enrichment to UF6 reduction and metal production. Section 

2.4.4, deleted connecting and disconnecting UF6 tails cylinders and 

handling tails as UF6, was not appropriate in this section. Section 

2.4.7, paragraph 2, corrected wording – changed reassembling to 

disassembling. Section 2.4.8, deleted Cascade operations – 1953–

1964, 1969–1970, and 1972–1976, was not appropriate in this 

section. Section 2.4.9, paragraph 1, changed converters to fluidized 

beds. Added Section 2.4.12, Building C-746B. Incorporates formal 

internal and NIOSH review comments. Constitutes a total rewrite of 

the document. Training required: As determined by the Objective 

Manager. Initiated by Jodie L. Phillips.
Document No. ORAUT-TKBS-0019-2 Revision No. 03 Effective Date: 08/22/2012 Page 3 of 24

TABLE OF CONTENTS

SECTION TITLE PAGE

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

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

2.1.1 Purpose.....................................................................................................................6

2.1.2 Scope ........................................................................................................................6

2.2 Site Description........................................................................................................................6

2.3 Site Activities............................................................................................................................8

2.4 Site Processes .......................................................................................................................11

2.4.1 C-310 – Purge and Product Withdrawal Building ..................................................... 11

2.4.2 C-315 – Surge and Tails Withdrawal Building.......................................................... 12

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

Buildings..................................................................................................................12

2.4.4 Building C-340 – Metals Plant.................................................................................. 13

2.4.5 C-400 – Decontamination and Cleaning Building/Uranium 

Recovery Building.................................................................................................... 15

2.4.6 C-404 – Solid Radioactive Waste Disposal Area ..................................................... 17

2.4.7 C-409 – Stabilization Building .................................................................................. 17

2.4.8 C-410 – Feed Plant.................................................................................................. 17

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

2.4.10 C-710 – Analytical Laboratories............................................................................... 19

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

2.4.12 C-746B – South Warehouse.................................................................................... 20

2.5 Attributions and Annotations ..................................................................................................20

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

Glossary ............................................................................................................................................22

LIST OF FIGURES

FIGURE TITLE PAGE

2-1 PGDP area map.......................................................................................................................7

2-2 PGDP site map ........................................................................................................................8

2-3 PGDP uranium enrichment process .........................................................................................9
Document No. ORAUT-TKBS-0019-2 Revision No. 03 Effective Date: 08/22/2012 Page 4 of 24

ACRONYMS AND ABBREVIATIONS

AEC U.S. Atomic Energy Commission

CIP Cascade Improvement Program

CUP Cascade Upgrade Program

DOE U.S. Department of Energy

EEOICPA Energy Employees Occupational Illness Compensation Program Act

ft foot

g gram

HF hydrofluoric acid

kg kilogram

KOW Kentucky Ordnance Works

L liter

lb pound

mg milligram

MgF2 magnesium fluoride

mrem millirem

ORNL Oak Ridge National Laboratory

PGDP Paducah Gaseous Diffusion Plant

POC probability of causation

PORTS Portsmouth Gaseous Diffusion Plant

ppb parts per billion

ppm parts per million

RU recycled uranium

SRDB Site Research Database

TBD technical basis document

TRU transuranic

UF4 uranium tetrafluoride (green salt)

UF6 uranium hexafluoride 

UO2 uranium dioxide

UO2F2 uranyl fluoride

UO3 uranium trioxide 

U3O8 triuranium octaoxide (yellowcake)

U.S.C. United States Code

USEC United States Enrichment Corporation

wk week

§ section, sections
Document No. ORAUT-TKBS-0019-2 Revision No. 03 Effective Date: 08/22/2012 Page 5 of 24

2.1 INTRODUCTION

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

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

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

dose reconstructions at particular sites or categories of sites. They will be revised in the event 

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

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

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

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

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

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

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

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

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

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

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

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

facility encompassed by EEOICPA.

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

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

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

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

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

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

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

determined in accordance with the POC [probability of causation1

] guidelines established under 

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

occupationally derived (NIOSH 2010):

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

conventional structures

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

reasons

 1 The U.S. Department of Labor (DOL) is ultimately responsible under the EEOICPA for determining the POC.
Document No. ORAUT-TKBS-0019-2 Revision No. 03 Effective Date: 08/22/2012 Page 6 of 24

2.1.1 Purpose

The purpose of the Paducah Gaseous Diffusion Plant (PGDP) 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 (Bechtel Jacobs 2000). The enriched product from 

PGDP was sent to other DOE gaseous diffusion plants at Portsmouth, Ohio (PORTS), and Oak 

Ridge, Tennessee, for further enrichment.

2.1.2 Scope

This TBD, which is part of the 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. 

PGDP is one of the original sites that was designated by Congress as part of the Special Exposure 

Cohort (SEC) under EEOICPA [42 U.S.C. § 7384l(14)]. This designation is as follows:

(A) The employee was so employed for a number of work days aggregating at least 

250 work days before February 1, 1992, at a gaseous diffusion plant located in 

Paducah, Kentucky, Portsmouth, Ohio, or Oak Ridge, Tennessee, and, during such 

employment—

(i) was monitored through the use of dosimetry badges for exposure at the plant of 

the external parts of employee’s body to radiation; or

(ii) worked in a job that had exposures comparable to a job that is or was monitored 

through the use of dosimetry badges.

Dose reconstruction guidance in this document is presented to provide a technical basis for dose 

reconstructions for nonpresumptive cancers that are not covered in the SEC class through January 

31, 1992. Dose reconstructions for individuals employed at PGDP before February 1, 1992, but who 

do not qualify for inclusion in the SEC, can be performed using this guidance as appropriate.

2.2 SITE DESCRIPTION

PGDP was built on a portion of 16,126 acres of farmland acquired by the U.S. War Department [the 

U.S. Department of Defense did not exist until 1947] during World War II. The War Department

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 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 KOW were acquired by 

the U.S. Atomic Energy Commission (AEC; a DOE predecessor agency) as a site for a uranium 

enrichment facility that became PGDP (see Figure 2-1). The construction of the plant began in 1951 

in response to the increased demand for highly enriched uranium resulting from nuclear weapons 

production. Initial operations began in 1952, but construction was not complete until 1954. Full 

operation was reached by 1955. The facility reservation covered 3,424 acres, with about 750 acres 

inside the security fence. The rest of the land was transferred to the Tennessee Valley Authority 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). 
Document No. ORAUT-TKBS-0019-2 Revision No. 03 Effective Date: 08/22/2012 Page 7 of 24

In addition to producing enriched uranium for weapons production, the plant also supplied enriched 

uranium for the Navy and for commercial fuel. The PGDP also acted as the uranium hexafluoride 

feed point for all gaseous diffusion plants until 1964. Union Carbide was the initial contractor at the 

site. Martin Marietta replaced Union Carbide in 1984. Martin Marietta merged with Lockheed to form 

Lockheed Martin in 1993. Congress established the United States Enrichment Corporation (USEC) 

as part of the Energy Policy Act of 1992. USEC was established as a government owned corporation 

to manage the uranium enrichment enterprise and prepare it for privatization through an initial public 

offering. The DOE has awarded remediation contracts to various contractors since 1998, including 

Bechtel Jacobs Company, Paducah Remediation Services, Swith and Stanley, and LATA (Los 

Alamos Technical Associates) Kentucky, the current contractor. In May of 1998 USEC terminated 

Lockheed Martin as the managing and operating contractor and assumed responsibility for the 

gaseous diffusion operations at PGDP. The net effect of these many changes is that the areas of the 

PGDP that are leased by USEC are subject to regulation by the Occupational Safety and Health 

Administration and the Nuclear Regulatory Commission and the areas that are retained by the DOE 

under the stewardship of Bechtel Jacobs LLC remain regulated by the DOE (PACE and University of 

Utah 2000). Currently, production of low enriched uranium for commercial use continues under the 

USEC, and the DOE operates the waste management and environmental restoration programs (DOE 

2011). The acreage encompassed by the DOE facility after July 28, 1998 is a subset of the acreage 

prior to that time. From 1952 - July 28, 1998 all acreage was considered a DOE facility. After that 

time, only the roads and grounds outside the perimeter fence plus approximately 200 acres of 

grounds inside the fence remain under the control of the DOE’s office of Environmental Management. 

Uranium Disposition Services has been the contractor for the design, construction and operation of 

the Depleted Uranium Hexafluoride Conversion Project since August 2002.

 Figure 2-1. PGDP area map. 
Document No. ORAUT-TKBS-0019-2 Revision No. 03 Effective Date: 08/22/2012 Page 8 of 24

Figure 2-2. PGDP site map (PACE 2000).

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), 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 produced UF6 from the recycled uranium (RU) produced from 

spent reactor fuel processed at the DOE 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 2000).

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 65% in 1973 to 3% in 1975 (Bechtel Jacobs 

2000). Processing of UO3 to UF6 occurred in three steps: reduction, hydrofluorination, and 

fluorination. The uranium exposure pathway of greatest concern at PGDP was inhalation of uranium 

dust (Bechtel Jacobs 2000).
Document No. ORAUT-TKBS-0019-2 Revision No. 03 Effective Date: 08/22/2012 Page 9 of 24

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 occurred 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 (Bechtel Jacobs 2000).

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 PORTS for further enrichment. The depleted material either was re-fed to the cascade or 

was stored on the site. Figure 2-3 shows the uranium enrichment process (Bechtel Jacobs 2000).

Figure 2-3. PGDP uranium enrichment process (Lockheed Martin 1997a).
Document No. ORAUT-TKBS-0019-2 Revision No. 03 Effective Date: 08/22/2012 Page 10 of 24

There were two cascade improvement/upgrade programs at PGDP – the Cascade Improvement 

Program (CIP) and the Cascade Upgrade Program (CUP). The first ran from 1958 to 1962. This 

campaign was an improvement program only. The second ran from 1973 to 1981. This campaign 

was both an improvement and an upgrade program. These programs were significant because of 

possible worker exposure to transuranic (TRU) elements 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 Buildings C-410 and C-420 and the decontamination building, C-400, were 

permanently shut down in the late 1970s (PACE and University of Utah 2000). Activities in support of 

the cascade as well as other plant support activities continue in Building C-400.

PGDP processed RU from initial startup in fiscal year 1953 through fiscal year 1989 (excluding fiscal 

years 1965 through 1968, 1976, and 1977 through 1985). In 1957, the presence of 237Np and 99Tc 

was documented; between 1959 and 1966, PGDP and the AEC conducted studies related to the 

behavior, health effects, and controls for these radionuclides (Bechtel Jacobs 2000). The 

concentration of TRU elements, such as 237Np and 239Pu, fission products, and 99Tc, in the reactor 

tails material was small, estimated at approximately 0.2 ppm neptunium, 4 ppb 239Pu, and 7 ppm 99Tc

(Bechtel Jacobs 2000). In addition, 241Am, a decay product of 241Pu, builds up as 241Pu decays. 

These radionuclides were concentrated during processing at specific locations, which could increase 

radiation exposure to certain workers in these areas (Bechtel Jacobs 2000). 

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 

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

building (C-400) with decontamination workers receiving the highest doses; and the cascade buildings 

(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.

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/wk, but suggested that PGDP kept workers below the annual regulatory 

limits by rotation of duties and limitations on stay time in areas of higher exposure (PACE and 

University of Utah 2000).

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 and

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 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 
Document No. ORAUT-TKBS-0019-2 Revision No. 03 Effective Date: 08/22/2012 Page 11 of 24

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 had lower potential for worker radiation 

exposure. In addition, due to the lack of aggressive contamination control programs, contamination 

from the operations discussed 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 (PACE and University of Utah 

2000).

Special Incidences or Activities – High Potential for Increased Radiation Exposure (PACE and 

University of Utah 2000)

C-310 fire, 1956

C-337 fire, 1962

C-340 explosion and fire, 1962

C-315 fire, 1978

CIP, 1954–1962

CIP/CUP, 1973–1981

C-400 neptunium production

2.4 SITE PROCESSES

The major facilities at PGDP are:

C-300 – Central Control Building

C-310 – Purge and Product Withdrawal Building

C-315 – Surge and Tails Withdrawal Building

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

C-340 – Metals Plant (currently inactive)

C-400 – Decontamination and Cleaning Building

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

This section also includes other key facilities where personnel radiation exposure occurred (e.g., C404, C-409, C-710, C-720, and C-746B). 

2.4.1 C-310 – Purge and Product Withdrawal Building

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

transport. Building C-310 was approximately 53 ft by 30 ft 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 (PACE and University of Utah 2000).

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 (Bechtel Jacobs 2000).

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 (Bechtel Jacobs 2000).
Document No. ORAUT-TKBS-0019-2 Revision No. 03 Effective Date: 08/22/2012 Page 12 of 24

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) (PACE 

and University of Utah 2000).

Primary Operations Performed in Building C-310:

Connecting and disconnecting UF6 product cylinders and handling product UF6 – 1953 –present;

Minimal external radiation exposure potential

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

Moderate external radiation exposure potential

2.4.2 C-315 – Surge and Tails Withdrawal Building

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 ft by 30 ft in area and

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

roll-up doors in the east wall permitted 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 (PACE and University of Utah 2000).

Liquefaction was accomplished by compression of UF6 gas 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 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 and 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 and University of 

Utah 2000).

Primary Operations Performed in Building C-315:

Connecting and disconnecting UF6 tails cylinders and handling tails UF6 – 1953–present;

Minimal external radiation exposure potential

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

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

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

in its gaseous form (Bechtel Jacobs 2000).
Document No. ORAUT-TKBS-0019-2 Revision No. 03 Effective Date: 08/22/2012 Page 13 of 24

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 and July 1954, Buildings C-335 and C-337, respectively, began 

operations (Bechtel Jacobs 2000).

Primary Operations Performed in the Cascade Buildings:

Cascade operations – 1952–1964, 1969–1970, and 1972–1976;

Moderate external radiation exposure potential

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

High external radiation exposure potential

2.4.4 Building C-340 – Metals Plant

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. Both processes generated considerable amounts of dust (PACE and 

University of Utah 2000).

Metals production involved several steps, each with its own hazards (Bechtel Jacobs 2000):

• 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 at which 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.

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 on 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 
Document No. ORAUT-TKBS-0019-2 Revision No. 03 Effective Date: 08/22/2012 Page 14 of 24

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

sometimes covered with black dust. 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 ORAUT 2012b).

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 and University of Utah 2000). After UF6 reduction and 

metal production operations were shut down in the late 1970s, the building was used for other 

activities (e.g., offices and training programs).

Primary Operations Performed in Building C-340:

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

Minimal external exposure potential

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

1967–1977;

Minimal external exposure potential

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

and drumming – 1957–1962 and 1968–1977;

High external exposure potential
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2.4.5 C-400 – Decontamination and Cleaning Building/Uranium Recovery Building

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

chemically separate and recover uranium from 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 occurred there: converter disassembly, pulverization of waste UF4

and recycled UO3 containing TRU elements, 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 2000).

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 2000).

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

(Bechtel Jacobs 2000).

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 that contained the component, reduced the UF6 in the cell to less than 10 ppm 

equivalent 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, that cell was turned over to process maintenance for opening and disassembly (Bechtel 

Jacobs 2000).

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 (Bechtel Jacobs 

2000).

Neptunium Recovery

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

237Np recovery process developed at Oak Ridge National Laboratory (ORNL) began operation 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 washwater 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 
Document No. ORAUT-TKBS-0019-2 Revision No. 03 Effective Date: 08/22/2012 Page 16 of 24

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 2000).

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 2000).

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 washwater and 

raffinate (e.g., solvents) from 237Np recovery operations began in April 1960 (Bechtel Jacobs 2000).

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 ORNL. 

Although the contribution to radiation dose from 99Tc was not of concern during most PGDP operation 

and maintenance activities, the isolated activity required specific monitoring considerations for both 

internal exposure and external skin exposure (Bechtel Jacobs 2000).

Cylinder Cleaning

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

autoclave. These deposits, called cylinder heels, had to be dissolved and removed periodically, 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. In relation to contaminants of the 

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

remained 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 2000).

Primary Operations Performed in Building C-400:

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

High external radiation exposure potential

Uranium/neptunium recovery – 1958 to the late 1970s;

Moderate external radiation exposure potential
Document No. ORAUT-TKBS-0019-2 Revision No. 03 Effective Date: 08/22/2012 Page 17 of 24

2.4.6 C-404 – Solid Radioactive Waste Disposal Area

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 ft by 140 ft, with 6-ft-high dikes. In 

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

approximately 6,400,000 lb 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, requiring closure 

under the Resource Conservation and Recovery Act in 1987 (Bechtel Jacobs 2000).

2.4.7 C-409 – Stabilization Building

Cascade Maintenance

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 (Bechtel Jacobs 2000).

During the 1973–1981 CIP/CUP, a shop for disassembling 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 was used in the initial stages of tearing down 

converters. The waste from this booth was piped to Building C-400 for removal of uranium oxides. It 

is not clear, but apparently the Building C-409 spray booth was operated by Building C-400 personnel

(Pace and University of Utah 2000).

Primary Operations Performed in Building C-409:

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

High external radiation exposure potential

2.4.8 C-410 – Feed Plant

From July 1953 through 1977, UF6 feed material was produced from UO3 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 the RU produced 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, 

PORTS, and Allied Chemical (Bechtel Jacobs 2000).

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 was liquefied and removed into large cylinders and the solid waste products were 

collected in ash receivers at the bottom of the towers. The external radiological concerns were from 

beta and gamma emissions from TRU elements, fission products, and accumulated uranium daughter 

products, which were concentrated by the process in the ash. In addition to external radiation 

sources, inhalation of radioactive dust while cleaning plugged equipment, changing the ash receivers, 

and cleaning 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 uranyl fluoride (UO2F2) (PACE and University of Utah 2000).
Document No. ORAUT-TKBS-0019-2 Revision No. 03 Effective Date: 08/22/2012 Page 18 of 24

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

example, the 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 (Bechtel Jacobs 

2000). 

Primary Operations Performed in Building C-410:

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

Moderate external radiation exposure potential

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

etc. (tower ash)] – 1953–1964 and 1969–1977;

High external radiation exposure potential

Changing/cleaning MgF2 traps – 1964–1966;

Moderate external radiation exposure potential

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

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 

used in the conversion processes. This was where triuranium octaoxide (U3O8) (yellowcake) was 

converted to uranium oxide and then to UF4 for use as feed stock for the fluorination towers in 

Building C-410. The ore ran through two sets of fluidized beds, the first of which converted UO3 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 fluidized beds or while cleaning the building air filtering system (baghouse). Maintenance 

mechanics had a potential for inhalation exposure while working on the equipment (Pace/University of 

Utah 2000).

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 

operators and maintenance mechanics would open the system and the operators and maintenance 

mechanics did what was necessary to resume operations. Routine operations in Building C-420 do 

not appear to have had the potential for increased radiation exposure; however, workers in this 

building were rotated with those in Building C-410 to reduce individual radiation exposure

(Pace/University of Utah 2000).

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

and University of Utah 2000).

Primary Operations Performed in Building C-420:

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

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

High external radiation exposure potential
Document No. ORAUT-TKBS-0019-2 Revision No. 03 Effective Date: 08/22/2012 Page 19 of 24

2.4.10 C-710 – Analytical Laboratories

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

operation since plant startup, consist of the American Society of Testing and Materials, 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 (Bechtel Jacobs 2000).

The 99Tc recovery operation, initially in Building C-710, moved to Building C-400 after 1959 (Pace and 

University of Utah 2000).

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

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

went 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 Building C-710 loading dock and 

shipped to ORNL (Bechtel Jacobs 2000).

A Health Physics and Hygiene report (Author unknown 1959) 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, a term used for airborne radioactivity concentrations 

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 and University of Utah 

2000).

Primary Operations Performed in Building C-710:

Uranium/neptunium recovery - 1958 – late 1970s;

Moderate external radiation exposure potential

2.4.11 C-720 – Maintenance Building

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 (Pace and 

University of Utah 2000).

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 (Pace and University of Utah 

2000).

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. Before removal from the system, block valves were slightly opened (where possible), 

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

installed, shipped to Building C-400 for preliminary disassembly and decontamination to the limits 
Document No. ORAUT-TKBS-0019-2 Revision No. 03 Effective Date: 08/22/2012 Page 20 of 24

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

(Bechtel Jacobs 2000).

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

Primary Operations Performed in Building C-720:

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

High external radiation exposure potential

2.4.12 C-746B – South Warehouse

Building C-746B was the site of smelter operations. Several smelters were operated to recycle scrap 

metals. These metals included steel, nickel, aluminum, copper, monel (a copper-nickel alloy), cobalt, 

gold, and silver (Pace and University of Utah).

Primary Operations Performed in Building C-746B: 

Metal and furnace scrap recovery – 1952–1986;

Low external radiation exposure potential

2.5 ATTRIBUTIONS AND ANNOTATIONS

All information requiring identification was addressed via references integrated into the reference 

section of this document.
Document No. ORAUT-TKBS-0019-2 Revision No. 03 Effective Date: 08/22/2012 Page 21 of 24

REFERENCES

Author unknown, 1959. Quarterly Report – January, February, March, 1959, Health Physics and 

Hygiene. [SRDB Ref ID: 18755, pp. 3-4]

Bechtel Jacobs (Bechtel Jacobs Company, LLC), 2000, Recycled Uranium Mass Balance Project 

Paducah Gaseous Diffusion Plant Site Report, BJC/PGDP-167, Final Draft, Paducah Gaseous 

Diffusion Plant, Paducah, Kentucky, June 14. [SRDB Ref ID: 16498]

DOE (U.S. Department of Energy), 2011, “DOE Facts, Paducah Gaseous Diffusion Plant,” Oak Ridge 

Office, Oak Ridge, Tennessee, September 2. [SRDB Ref ID: 101048]

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

Environmental Report for 1996, KY/EM-206, Kevil, Kentucky, December. [SRDB Ref ID: 

12316]

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

1995, Document No. KY/EM-176, Kevil, Kentucky, January. [SRDB Ref ID: 12314]

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

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

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

Support, Cincinnati, Ohio, October.

ORAUT (Oak Ridge Associated Universities Team), 2012a, Paducah Gaseous Diffusion Plant –

Occupational Internal Dose, ORAUT-TKBS-0019-5, Rev. 03, Oak Ridge, Tennessee, August 

24.

ORAUT (Oak Ridge Associated Universities Team), 2012b, Paducah Gaseous Diffusion Plant –

Occupational External Dose, ORAUT-TKBS-0019-6, Rev. 04, Oak Ridge, Tennessee, August 

24.

PACE (Paper, Allied Industrial, Chemical and Energy Workers International Union) and University of 

Utah, 2000, Exposure Assessment Project at the Paducah Gaseous Diffusion Plant, 

December. [SRDB Ref ID: 10870]
Document No. ORAUT-TKBS-0019-2 Revision No. 03 Effective Date: 08/22/2012 Page 22 of 24

GLOSSARY

absorption type

Categories for materials according to their rate of absorption from the respiratory tract to the 

blood, which replaced the earlier inhalation clearance classes. Defined by the International 

Commission on Radiological Protection, the absorption types are F: deposited materials that 

are readily absorbed into blood from the respiratory tract (fast solubilization); M: deposited 

materials that have intermediate rates of absorption into blood from the respiratory tract 

(moderate rate of solubilization); and S: deposited materials that are relatively in the 

respiratory tract (slow solubilization). Also called solubility type. See inhalation class.

cascade

At PGDP, 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 in a 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) 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

System of controls to reduce and confine radioactive contamination. These systems include 

contamination surveys, posting, containment devices, and the prevention, detection, and 

timely repair of leaks.

depleted tails

Material that has passed through one of the cascade stages that has been depleted of some 

of its 235U.

derby

At PGDP, molded uranium metal ingot of uranium tetrafluoride (UF4) and magnesium.

dosimetry

Measurement and calculation of internal and external radiation doses.

enriched uranium

Uranium in which processing has increased the proportion of 235U to 238U to above the natural 

level of 0.7%. Reactor-grade uranium is usually about 3.5% 235U; weapons-grade uranium 

contains greater than 90% 235U.

external radiation

Radiation from sources outside the body.

feed material

At PGDP, uranium hexafluoride (UF6) used in the cascade systems for the enrichment 

process.

fission product

Radionuclide produced by fission or by the subsequent radioactive decay of radionuclides.
Document No. ORAUT-TKBS-0019-2 Revision No. 03 Effective Date: 08/22/2012 Page 23 of 24

gaseous diffusion enrichment

Process by which uranium hexafluoride (UF6) is filtered through a series of semipermeable 

molecular barriers to separate the lighter 235U from the heavier, more naturally abundant 238U. 

High Radiation Area

Area with dose rates greater than 100 millirem (1 millisievert) in 1 hour at 30 centimeters from 

the source or its container. HRAs require posted signs and strict access control.

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

(ICRP 1979). 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. More recent recommendations in ICRP Report 66 replaced classes D, 

W, and Y with lung absorption Types F (fast), M (moderate), and S (slow), respectively (ICRP 

1994). See absorption type.

internal radiation exposures

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

body.

liquefaction

Process of causing a gas or solid to become a liquid, usually through condensing a gas or 

melting a solid.

natural uranium

Uranium as found in nature, approximately 99.27% 238U, 0.72% 235U, and 0.0054% 234U by 

weight. The specific activity of this mixture is 2.6 × 107 becquerel per kilogram (0.7 picocurie 

per gram).

pigtail

At PGDP, a flexible connecting device for draining depleted liquefied uranium hexafluoride 

(UF6) gas into storage cylinders.

production reactor tails

At PGDP, uranium reprocessed from spent reactor fuel used in a plutonium production reactor 

(e.g., at the Hanford and Savannah River Sites).

reactor tails

At PGDP, uranium reprocessed from spent reactor fuel, normally from production reactors, but 

some from reactors in foreign countries.

reactor vessel

At PGDP, hollow receptacle in which a chemical reaction takes place.

refractory cap

At PGDP, cover of highly heat-resistant material under the reactor vessel lid.

roasting

At PGDP, process of heating derbies to remove all the slag produced during the reduction 

process [heating uranium tetrafluoride (UF4) and magnesium to form uranium derbies].
Document No. ORAUT-TKBS-0019-2 Revision No. 03 Effective Date: 08/22/2012 Page 24 of 24

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 atomic numbers above 92 (uranium). Examples include plutonium and 

americium. All isotopes of the transuranic elements are radioactive, they are naturally either 

rare or nonexistent on Earth, and most are known only as a result of research using nuclear 

reactors and particle accelerators because of extremely short half-lives.

weekly plant action level

At PGDP, radiation exposure or 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 or 

airborne radioactivity.