ORAU Team

Dose Reconstruction Project for NIOSH

Technical Basis Document: Basis for the Development of an 

Exposure Matrix for Aliquippa Forge, Pennsylvania, Period of 

Operation: July 28, 1948 through February 28, 1950

Document Number:

ORAUT-TKBS-0021

Effective Date: 12/21/2004

Revision No.: 00

Controlled Copy No.: ______

Page 1 of 25

Subject Expert: Shelby L. Gubin

Document Owner

Approval: Signature on File Date: 11/19/2004

Cindy W. Bloom, TBD Team Leader

Approval: Signature on File Date: 11/19/2004

Judson L. Kenoyer, Task 3 Manager

Concurrence: Signature on File Date: 11/30/2004

Richard E. Toohey, Project Director

Approval: Signature on File Date: 12/21/2004

James W. Neton, Associate Director for Science

Supersedes:

None

TABLE OF CONTENTS

Section

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

Page

3

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

1.0 Introduction.................................................................................................................................5

2.0 Site Description and Operational History.....................................................................................5

3.0 Estimation of Internal Exposure ................................................................................................ 12

3.1 Uranium Air Sampling .........................................................................................................12

3.2 Occupational Internal Dose Reconstruction Assumptions and Summary............................. 15

4.0 Estimation of External Exposure ............................................................................................... 16

4.1 Submersion and Contamination Exposures.........................................................................16

4.2 Uranium Billet and Rod Exposures......................................................................................17

4.3 Occupationally Required Medical X-Ray .............................................................................18

4.4 Miscellaneous Information Related to External Dose ..........................................................18

4.5 Occupational External Dose Reconstruction Assumptions and Summary ........................... 18

5.0 Estimation of Residual Exposure .............................................................................................. 19

References ........................................................................................................................................22
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LIST OF TABLES

Table

1 Documented uranium rollings at Vulcan Crucible .....................................................................

Page

8

2 Number of assumed workdays and uranium rolling-days ....................................................... 10

3 Job titles involved in uranium rolling....................................................................................... 11

4 Daily time-weighted average air concentration information..................................................... 13

5 Internal exposures during rolling operations for Group I workers............................................ 14

6 Internal exposures during rolling operations for Group II and III workers................................ 14

7 Annual exposure due to resuspension of deposited uranium dust during 

non-AEC operations...............................................................................................................15

8 Estimated amount of uranium ingested (based on Tables 5 and 7)........................................ 15

9 Natural uranium chronic intake assumptions for unmonitored workers................................... 15

10 Calculated dose rate for uranium billet and uranium rod ........................................................ 17

11 Various direct radiation measurements from September 1948 ............................................... 17

12 External exposure summary for operational period July 23, 1948, to February 

28, 1950.................................................................................................................................18

13 Annual internal and external exposure to residual radioactivity .............................................. 20

LIST OF FIGURES

Figure

1 Aliquippa Forge Site about 1992 ..............................................................................................

Page

6

2 Aliquippa Forge Buildings 3 and 8 where AEC uranium operations occurred ........................... 7
Effective Date: 12/21/2004 Revision No. 00 Document No. ORAUT-TKBS-0021 Page 3 of 25

RECORD OF ISSUE/REVISIONS

ISSUE 

AUTHORIZATION 

DATE

EFFECTIVE 

DATE

REV. 

NO. DESCRIPTION

Draft 01/05/2004 00-A New Technical Basis Document: Basis for the 

Development of an Exposure Matrix for Aliquippa 

Forge, Aliquippa, Pennsylvania, Period of 

Operation: August 16, 1948 through February 28, 

1950. Initiated by Jeri L. Anderson.

Draft 01/20/2004 00-B Incorporates internal review comments. Initiated 

by Jeri L. Anderson.

Draft 04/27/2004 00-C Incorporates OCAS review comments. Initiated 

by Jeri L. Anderson.

Draft 06/15/2004 00-D Incorporates OCAS review comments and internal 

comments. The format is changed to be more in 

line with the IAAP TBD. The individual bioassay 

data analysis is removed. External dose 

assumptions regarding geometry and organ dose 

conversion factors, which are not site-specific, are 

eliminated in the TBDs and assigned by Task 5 to 

ensure programmatic consistency. Initiated by 

Shelby L. Gubin.

Draft 10/25/2004 00-E Incorporates ORAU Team comments. Major 

modifications include: passive to active voice, 

correction of some intake numbers in Section 3.2, 

assuming U-235 photons are in the 30 to 250 keV 

range, and the submersion/contamination external 

exposure evaluation. Initiated by Cindy W.

Bloom.

Draft 11/18/2004 00-F Incorporates OCAS comment to account explicitly 

for both nonpenetrating and penetrating radiations 

during the residual exposure period. Initiated by 

Cindy W. Bloom.

Draft 11/19/2004 00-G Corrects Typographical error of 2.5, on p. 20, line 

24, to 1.5. Initiated by Cindy W. Bloom.

12/21/2004 12/21/2004 00 First approved issue. Initiated by Cindy W. 

Bloom.
Effective Date: 12/21/2004 Revision No. 00 Document No. ORAUT-TKBS-0021 Page 4 of 25

ACRONYMS AND ABBREVIATIONS

AEC U.S. Atomic Energy Commission

BZ breathing zone

cm centimeter

dpm disintegrations per minute (also d/m)

ft foot

FUSRAP Formerly Utilized Sites Remedial Action Program

GA general area

GSD geometric standard deviation

h hour

ICRP International Commission on Radiological Protection

in. inch

IREP Interactive RadioEpidemiological Program

keV kilovolt-electron, 1,000 electron volts

L liter

m meter

MDA minimum detectable activity

mg milligram

mR milliroentgen

mrem millirem

mrep millirep

NYOO New York Operations Office

PA posterior-anterior

pCi picocuries

R roentgen

s second

U.S.C. United States Code

y year

μg microgram

μm micrometer
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1.0

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

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

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

These documents may be used to assist the National Institute for Occupational Safety and Health 

(NIOSH) in the completion of the individual work required for each dose reconstruction.

INTRODUCTION

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. Sections 7384l (5) and (12)].

This document provides an exposure matrix for workers at the facility listed as Aliquippa Forge in 

Aliquippa, Pennsylvania. At the time of U.S. Atomic Energy Commission (AEC) contract operations, 

Aliquippa Forge was known as Vulcan Crucible Steel Company. Vulcan Crucible was primarily 

involved with the rolling of natural uranium. Some time after AEC operations ended, the facility 

became known as Universal Cyclops. The facility is now owned by the Beaver County Corporation for 

Economic Development. 

2.0

The information that follows supports an assumed period of AEC operations at Aliquippa Forge from 

July 23, 1948, through February 28, 1950, involving AEC-contracted uranium work. This analysis 

assumed that the residual contamination period extended from March 1, 1950, through May 31, 1978,

and from December 1, 1987, to May 17, 1993.

SITE DESCRIPTION AND OPERATIONAL HISTORY

The Aliquippa Forge radiological source term consisted primarily of natural uranium metal, uranium 

oxides, and natural uranium’s short-lived progeny. Long-lived progeny prevent significant ingrowth 

past 234U in the 238U decay series and beyond 231Th in the 235U decay series. 

Vulcan Crucible Steel Company produced uranium rods for the AEC from billets primarily by rolling. 

Operations with uranium at Vulcan Crucible began when a trial rolling occurred on July 23, 1948 

(Jones 1948). The AEC contract for production work (AEC 1948a) was initiated on August 16, 1948, 

and was extended through February 28, 1950 (Belmore 1950; Wallo 1981). The rolling operation 

ended on March 30, 1949 (AEC 1949a) with decontamination consuming the rest of the contract’s 

term. Decontamination was completed by the Vulcan Crucible in 1950 in accordance with thencurrent AEC guidelines. 

The site consisted of about 19 buildings. The majority of the AEC work occurred in Building 3, the 

rolling mill. Survey results for Building 8 indicate its involvement with uranium activities (Adams and 

Payne 1992a; 1992b). In addition, there were indications of uranium in the locker room, tool room,

and some areas outside Building 3. Figure 1 shows the layout of the site circa 1990 and Figure 2 

shows the layout of the area that encompassed uranium operations. A sketch from 1948 (AEC 

1948b) indicates that at least part of the Building 3/8 area was referred to as C Mill. That sketch 

showed billets in an area known later as Building 8 and a boxcar containing billets and rods to the 

northwest of Building 8; it also showed the lockers, showers, and toilets to the east of Building 3. 

The uranium billets furnished by the AEC came primarily via boxcar from the Electromet facility in 

New York and the 300 Area on the Hanford Site in Washington (Stroke 1949a). The billets were 15 to 

28 in. long, 4 to 5 in. in diameter, and weighed from 120 to 270 pounds. Vulcan Crucible conducted a 
Effective Date: 12/21/2004 Revision No. 00 Document No. ORAUT-TKBS-0021 Page 6 of 25

rolling operation in Building 3 to reduce the billets to rods about 1.5 in. in diameter with an increase in 

length of a factor of 9. The billets were heated and rough-rolled twice to increase the length, then 

finish-rolled and halved before quenching and weighing (Wallo 1981). 

Figure 1. Aliquippa Forge Site circa 1992 (DOE 1996)

The process at Aliquippa Forge consisted of heating the billets to a temperature between 1,050°F and 

1,100°F. The drag-down operator used a buggy to move the heated billets from the furnace to the 

north side of the roughing roll. The billets were passed through the roughing roll two to four times to 

produce rods of rough dimensions. The rods were then passed through the finishing rolls to achieve 

the desired dimensions. They were then dragged to the shears, cut in two, and dragged back to the 

quenching area for descaling. After the rods were stamped, they were conveyed to the shipping and 

receiving area to be weighed, recorded, and loaded into boxcars (VCSC c.1949; AEC 1949b; Author 

unknown 1948). 

Little information is available on the actual uranium rollings that took place at Vulcan Crucible during 

the AEC contract period. Documents indicate that each rolling operation involved approximately 96 

billets in an 8-hour shift (Stroke 1949b; Reichard 1948a). Other documents indicate that Vulcan 

Crucible had a 9-hour workday with a rolling rate of 11 to 12 billets per hour (Breslin 1949), which 

seems consistent with the February 1949 workplace monitoring records, if one additional hour is 
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assigned to lunch and breaks. Table 1 summarizes the available uranium rolling information (Jones 

1948; Huff 1948; Schier 1948; Author unknown 1948; Reichard 1948b). Although some 

documentation indicates that 20% of the time at Vulcan Crucible was spent rolling uranium billets for 

the AEC (Wallo 1981), the AEC contract required that Vulcan Crucible be prepared to perform rolling 

work at least two consecutive weeks out of every five consecutive weeks (AEC 1948a), which would 

be 40% of the time.

Figure 2. Aliquippa Forge Buildings 3 and 8 where AEC uranium 

operations occurred (DOE 1996)

Vulcan Crucible employed about 20 to 25 workers at the rolling plant with approximately 20% of the 

total rolling time occupied by AEC work (Wallo 1981). AEC personnel made occasional visits to 

assess working conditions in the mill and attended some, but not all, rolling operations. 

As of September 2, 1948, the exhaust ventilation system consisted of two large roof ventilators for the 

building (AEC 1948b). The ventilator placement was not considered ideal for the uranium rolling 

operations. 
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Table 1. Documented uranium rollings at Vulcan Crucible.

Date of rolling Type No. of billets rolled

July 23, 1948 a Trial 8

August 23-September 2, 1948 b Production 822 Type B (80 tons)

September 27, 1948 c Experimental forging 50

October 21-November 3, 1948 d Production 

Experimental forging

Unknown

52 forged

January 3- 15, 1949e. Production 982 Type B billets

103 Type C billets 

(~120 tons)

February 14-23, 1949 f Production 112 “small billets per shift”

March 30, 1949 g Unknown 70 “large billets per shift”

a. Jones (1948)

b. Author unknown (1948)

c. Schier (1948)

d. Hauff (1948), OSTI (2002)

e. Padden (1949)

f. Breslin (1949), OSTI (2002)

g. Breslin (1949)

During the September 1948 AEC visit, a large door and windows at the end of the mill were open, 

which reportedly carried dust to the southeast of the plant (AEC 1948b). The AEC Medical Division 

recommended that Vulcan Crucible upgrade the exhaust ventilation system and install a central 

vacuum to maintain cleanliness. The recommendations were transmitted to Vulcan Crucible on 

October 14, 1948, as requirements for “the contemplated metal rolling contract” (Belmore 1948). A 

November 2, 1948, memorandum indicates that the vacuum system had been installed (Reichard 

1948a). However, as of the February 15-16, 1949, rolling, the upgrade to the ventilation system had 

not been made (AEC 1949b), although temporary ventilation over the rollers was provided for that 

rolling campaign. 

AEC visits were made before and after modifications to the ventilation system (Belmore 1948, AEC 

1948b, 1949b; Breslin 1949, Klevin 1949). The Monthly Status and Progress Report for April 1949

(AEC 1949a) states:

Since the Vulcan Crucible Steel Company cannot handle the larger-size 

billets which give better casting yields, it was decided to consolidate 

operation at Simonds. Because of this decision, rolling operations 

ceased at Vulcan after the March [1949] run. Portable Government 

property is to be removed immediately. Attached property is to be left at 

Vulcan, provision will be made for emergency standby facilities until the 

contract expires in August. 

During World War II, permissible levels for uranium dust in air were set at 500 μg/m3 for insoluble 

uranium compounds and 150 μg/m3 for soluble uranium compounds. After the war, the University of 

Rochester lowered its recommendation for soluble uranium compounds to 50 μg/m3 based on

chemical toxicity, which for natural uranium is equivalent to 70 dpm/m3

. This level was based 

primarily on animal studies. The Medical Division of the AEC New York Operations Office (NYOO) 

felt that a "maximum permissible level" was unknown and should be based on human data. 

Therefore, the 50-μg/m3 level was referred to as the "preferred level" (AEC 1949c). 

The AEC visited Vulcan Crucible on September 1, 1948, to conduct a health physics survey and “to 

determine:
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1. The type of physical examination given to the men

2. The use of protective clothing, showers, and lunchroom

3. The exposure of personnel to direct radiation

4. The spread of contamination in the plant

5. The concentration of radioactive dust in the air.” (AEC 1948b)

Medical examination requirements for Vulcan Crucible uranium workers were specified in October 

1948 (Belmore 1948). The AEC staff noted on March 29, 1949, that “a complete blood count, a urine 

and a chest x-ray were done on all employees exposed in the rolling area before work on the AEC 

Project was started” (Tabershaw 1949) and “a repeat blood count and urine [medical, not radioactivity 

analyses] were done on all workers after a period of 6 months.” The X-ray examinations were 

performed by Dr. W. T. Rice, who kept the films in his Rochester, Pennsylvania, office.

As of the September 1, 1948 visit, there were few health physics controls in place. AEC (1948b) 

noted, “workers furnish and launder their own clothes, shoes and gloves and usually change before 

going home. Separate lockers are not provided for clean and dirty clothes. Smoking is permitted in 

the mill and some men were observed smoking with their dirty gloves on. The men also eat their 

lunch in the vicinity of the mill.” In addition, the report noted that workers were loitering or standing 

near uranium billets or rods, and one even sat on billets while taking a break. 

In October 1948, the AEC specified “protective clothing and hygienic procedures” (Belmore 1948). 

Protective clothing was listed as dedicated work clothing, gloves, and shoes. The AEC recommended 

that Vulcan Crucible provide workers with clothing that could be kept separate from personal clothes. 

Workers were to be instructed to not eat or smoke with gloved hands, and to wash hands thoroughly 

before smoking, eating, or leaving the shift. The AEC also recommended showering at the end of the 

shift. No later reports were found to indicate that the company supplied clothing, but there was 

indication that some workers showered and changed at the end of the shift. The available records 

indicate contamination controls were not strictly implemented. 

The air samples taken during the September 1948 visit indicated one sample (during the third pass in 

back of the mill) showing an air concentration as high as 1,800 times the preferred level. Although 

peak values need to be considered, the fact that work tasks and worker locations were constantly 

changing resulted in time-weighted exposures that were typically much lower than the peak values. 

The other air sampling results for September 1 and 2, 1948, showed concentrations in the range of 

2.6 to 510 times the preferred level, with 15 of the 22 results less than or equal to 50 times the 

preferred level. 

A NYOO report of an AEC visit to Vulcan Crucible on February 15 and 16, 1949, describes timeweighted radioactive dust exposures between 2.7 and 5300 times the preferred level depending on 

the type of job (AEC 1949b). A review of the February 1949 report and calculations shows that 

February report was in error and overstated the maximum exposure by about a factor of 10. When 

discussing the higher concentration, the report states: 

It was noted during the sampling that relatively large flakes of scale were 

being thrown from the rods at this operation. The above concentrations 

may therefore represent some number of large, non-respirable particles, 

and not be a true indication of exposure. 

A later summary report (AEC 1949c) stated that four people directly involved in the rolling were 

exposed to as much as 530 times the preferred level. The summary report mentioned that the 
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recommended ventilation system upgrade had not been installed, but was on order and that a 

temporary although inadequate system was in place. 

The AEC visit of March 30, 1949, occurred after the installation of a new exhaust fan with twice the 

capacity of the previous fan. During this visit, the largest calculated time-weighted uranium exposure

was 21 times the preferred level. A reevaluation of one worker’s exposure based on a statement that 

one sample appeared to be unreasonably low indicates that the largest exposure might have been 57 

times the preferred level. This is still indicative of a reduction of uranium air concentrations due to the 

improved ventilation. The AEC noted during this visit that rather than rolling 112 “small” billets during 

a 9-hour shift, 70 “larger” billets were rolled in an 8-hour shift, which might also have contributed to the 

lower uranium dust concentrations (Breslin 1949). The AEC record indicates that there were no 

rolling operations after March 1949 and that only cleanup operations were taking place (AEC 1949a).

The purpose of the April 24, 1949, AEC visit was to survey the contamination resulting from previous 

rolling. The visit resulted in recommendations for the cleanup of the mill. The visit of July 21, 1949, 

evaluated the effectiveness of the recommended decontamination. During these two visits (Belmore 

1949b, Klevin 1949), AEC took direct measurements of surface contamination using a Zeuto, but 

collected no smear samples. (A Zeuto is a portable ionization chamber. The early models were used 

to measure alpha contamination; some models also measured beta and gamma radiation.)

Based on the time-motion information collected on February 15-16, 1949 (AEC 1949b) and March 30, 

1949 (Breslin 1949), this evaluation assumed 10-hour workdays prior to March 30, 1949, and 8-hour 

workdays thereafter. Table 2 lists the assumed number of workdays and uranium rolling days in each 

period.

Table 2. Number of assumed workdays and 

uranium rolling-days.

Start End

Non-uranium 

rolling workdays Rolling days

Calendar days

7/23/1948 12/31/1948 116 50 162

1/1/1949 3/29/1949 63 30 88

3/30/1949 12/31/1949 198 90 277

1/1/1950 2/28/1950 42 20 58

The analyses in this Technical Basis Document divide the workers into three categories, as listed in 

Table 3. While different tasks in the mill resulted in differences in exposures (AEC 1949b), it is not 

known if each worker always performed the same task within a group, or if workers temporarily 

worked in locations where higher or lower exposures occurred. Workgroup exposure assignments 

are based on data that are suggestive of workers’ exposures and further modified by uncertainty 

parameters to ensure that the reconstructed dose distributions capture the larger exposures. 

Depending on the organ of interest and the ancillary data on a specific claim, additional 

considerations might be appropriate.

• Group I workers were assumed to have been involved primarily with heating, rolling, finishing, 

and quenching of uranium. They spent the largest part of their time in the general mill area 

near the furnace and rollers.

• Group II workers were assumed to have worked primarily with the finished uranium rods and 

to have spent the largest part of their time in the rod storage area and boxcar. 
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• Group III workers were not engaged directly in the processing of uranium. They probably had 

lower internal and external exposures than the other groups, with the possible exception of the 

guards, who might have spent time near the boxcar and the uranium storage areas. 

Table 3. Job titles involved in uranium rolling.

Group I Group II Group III

Hook Man (Front Hooker, Back Hooker) Shipping and Receiving Guard

Drag-down Operator (Buggy Man) Checker Technical Supervisor

Rougher Rod Weigher Office Workers

Finisher

Shear Man

Shear Man Helper

Heater (Furnace Man)

Heater Helper (Furnace Man Helper)

Mill Roller

Catcher

Hot Sawyer 

Hot Sawyer Helper

Rod Stamper

Quencher

Table 3 categorizes Vulcan Crucible uranium rolling job titles in the designated groups. Exposures to 

workers who do not fit in the groups are designated to an “unknown” category, which is included in the 

internal and external exposure sections of this Technical Basis Document. 

In July of 1949, a survey was performed to determine Vulcan Crucible cleanup requirements (Belmore 

1949a). Additional AEC assessments/surveys were made throughout the cleanup process; and

although it was noted that the cleanup personnel had no monitoring equipment, AEC assessments

concluded that a sufficient job of cleanup had been done (Author unknown 1950; Belmore 1950). 

The Formerly Utilized Sites Remedial Action Program (FUSRAP) began in 1976. 

A radiological survey in 1978 identified contamination (primarily uranium238) in and around onsite buildings. [Uranium-238 is the predominant 

isotope by mass in natural uranium and is more easily identified than the 

other isotopes, so some may refer to it as uranium-238 rather than natural 

uranium, which consists of approximately equal activities of U-234 and U238 and a smaller amount of U-235. Reported U-238 quantities may 

include all of the uranium activity or just part, depending on actual 

analysis techniques and reporting procedures.] The site was designated 

for further remediation under FUSRAP, and the small operation was shut 

down and the building evacuated (Perry 1993). 

Radioactive contamination was found during the survey of May 2-8, 1978, on the dirt floor, concrete 

floor, steel floor plates, and the overhead beams above the furnaces used in the uranium processing

(Wynveen et al. 1982). The actual date of building evacuation is unknown and is assumed to have 

occurred on May 31, 1978, after the FUSRAP survey.

In August 1983, the Aliquippa Forge site was designated for remedial action under FUSRAP (DOE 

1996). In December 1987, storage activities began in Building 3. Interim remedial actions were taken 

from October to December 1988 to enable additional restricted use of Building 3 for expansion of a 

small forging operation (Baublitz 1988, Harbert 1989, DOE 1996). Controlled areas were established 

to prevent access to contamination (Seay 1988, DOE 1996). The exact date of reoccupation is 
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unknown, but is assumed to have occurred as early as December 1, 1987. As of May 17, 1993, the 

buildings were no longer in use, although the date that use ended is not clear. This analysis assumed 

that a second period of residual exposure occurred from December 1, 1987, to May 17, 1993. Final 

remedial activities occurred from about June 1993 to September 1994 (Abelquist 1995, DOE 1996). 

A final survey and decontamination were performed in 1995 (DOE 1996). A Department of Energy 

notice of certification was published in October 1996 in the Federal Register (61 FR 211, pp. 55981-

55982).

Because recycled uranium was not available to contracted AEC facilities until after March 1952 (DOE 

2001), the Vulcan Crucible exposure analysis did not consider it. 

3.0

The primary source of internal radiation exposure at Vulcan Crucible was uranium dust produced from 

the manipulation and oxidation of uranium metal during the rolling and related processes. There is no 

indication that uranium rolled at Vulcan Crucible was enriched, so the analysis assumed natural 

uranium enrichment. 

ESTIMATION OF INTERNAL EXPOSURE

Human and animal studies have indicated that oxides of uranium can be very insoluble (ICRP 1995), 

indicating absorption type S. Other in vitro dissolution studies of compounds found at uranium 

facilities have shown that oxides of uranium exhibit moderate solubility (Eidson 1994; Heffernan et al. 

2001) suggesting absorption type M. In vitro dissolution tests on oxides produced from uranium metal 

during depleted uranium armor penetrator tests have indicated multicomponent dissolution rates, with 

25% of uranium dissolving with a half-time of less than or equal to 0.14 days and 75% dissolving with 

a half-time of 180 days. Because there was no specific information on the solubility of aerosols 

produced at Vulcan Crucible, this analysis assumed that both types M and S were available. The 

selection of absorption type should depend on the organ of interest. Dose reconstructions should 

assume International Commission on Radiological Protection (ICRP) Publication 66 default 

parameters for particle deposition (ICRP 1994).

The uranium fusion photofluorimetry urinalyses performed by the University of Rochester and the 

AEC NYOO were similar to those performed at the Fernald Plant. The default detection threshold for 

uranium urinalysis is 14 μg/L, based on the Fernald reevaluation of its uranium fluorimetry urinalysis 

minimum detectable activity (MDA) in 1993 (Blalock 1993).

Individual uranium urinalysis data are available for some Vulcan Crucible workers (AEC 1949d). For 

unmonitored workers or unmonitored periods, this Technical Basis Document analyzes air monitoring 

data for use in reconstructing internal doses. 

3.1 URANIUM AIR SAMPLING

Air sampling was performed at Vulcan Crucible during some of the uranium rolling (AEC 1949c; 

Breslin 1949). The air samples consisted of collection on filters of radioactive particulate from

breathing zones (BZs), general areas (GAs), processes, and effluents. The alpha activity measured 

on the filter was used to determine airborne alpha activity concentrations. The AEC matched these

air concentration determinations with information about worker categories, locations, tasks, and 

workers’ time at each location or task. For some tasks and locations, multiple samples were 

collected; the mean count rate was calculated and used to calculate an average air concentration. 

The AEC used the information on work tasks with the measured air concentration to determine an 

average air concentration weighted by time, and summed these weighted average air concentrations 
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to determine a daily weighted average air concentration for specified job categories. These air 

concentration results are further analyzed here to determine group geometric means. The daily 

weighted average air concentrations were weighted further by the AEC-reported number of workers 

exposed at a given concentration. The geometric means of the workgroups’ daily weighted average 

air concentrations were calculated. The geometric standard deviations (GSDs) of both the job 

category concentrations and the workgroup concentrations (this latter GSD includes consideration of 

the number of people included in each category in the AEC study) were determined to provide an 

indication of the distribution of the data. This analysis assumed that, because data are limited and 

rigorous analyses to determine distribution type are not likely to be meaningful, a lognormal 

distribution could represent the time-weighted exposures and the subsequently derived annual organ 

doses. 

The report on the February 1949 rolling shows time-weighted air concentrations measured at the plant 

before the upgrade of the ventilation system. The report on the March 1949 rolling (Breslin 1949) 

shows the time-weighted air concentrations measured at the plant after the upgrade of the ventilation 

system. During spot-checks of the February 1949 air concentrations, this analysis revealed a 

mathematical error for the Hook Man’s daily weighted air concentration, which resulted in a change 

from the 36,800 dpm/m3 to 38,200 dpm/m3

. An error was also found for the Roughers February 

result, but the change was small and correction of the error would have slightly reduced the exposure, 

so it was ignored. The March report noted that “one unsubstantiated sample of unusually low 

concentration” probably resulted in an “erroneous” daily weighted average air concentration for the 

Hook Man. To remedy this, the March 30 average ratio of the second-to-first rolling pass air 

concentrations was determined. This ratio, 19.2, was used to modify the low air concentration result. 

The two sets of data from February and March 1949 were then used to estimate the air 

concentrations to determine internal exposures by workgroup. Table 4 lists the geometric means and 

GSDs for the job category and the workgroups’ daily time-weighted average air concentrations.

Table 4. Daily time-weighted average air concentration information.

Air samples collection date

Group I Group II and III

2/15/1949 3/30/1949 2/15/1949

Job categories Number of categories 9 9 6

Geometric means (dpm/m3) 2,210 545 484

GSD 5 3.2 2

Workgroup Number of workers 15 9 10

Geometric means (dpm/m3) 2,610 479 608

GSD 5.3 3 1.8

This analysis of intakes based on air concentrations made the claimant-favorable assumption that 

uranium rolling took place between July 23, 1948, and February 28, 1950. It also assumed that 10 

days of every month were spent rolling uranium because the AEC required Vulcan Crucible to be 

prepared to spend 2 consecutive weeks of every 5 consecutive weeks performing AEC work (AEC 

1948a). Although rolling did not actually occur after March 1949, data to estimate exposure directly 

from cleanup operations were not available. The analysis assumed that the internal exposure rates

during cleanup would not have exceeded the internal exposure rates during continued rolling 

operations with the improved ventilation. 

The breathing rate is based on the default for light work shown in ICRP Publication 66 (ICRP 1994, 

Table 6, p. 23). The intakes, in pCi, were calculated by dividing the geometric mean of a workgroup’s 

time-weighted air concentration by 2.22 dpm/pCi and multiplying this result by the breathing rate and 

the assumed number of hours exposed at the given concentration. Vulcan Crucible internal organ 

doses are assumed to be lognormally distributed and the GSDs for the calculated internal organ 
Effective Date: 12/21/2004 Revision No. 00 Document No. ORAUT-TKBS-0021 Page 14 of 25

doses are assumed to be 5.3. The primary bases for selecting a GSD of 5.3 for all Vulcan Crucible 

internal organ dose calculation is to simplify and expedite dose reconstructions, and to encompass 

the largest distribution from the air sampling data. Several assumptions included in the intake/dose 

reconstruction are likely to be overestimating assumptions, which increase the estimate of the median 

intakes. This overestimation of the median, combined with the assumed GSD of 5.3, is believed to be 

sufficiently large to describe the organ dose distributions.

Tables 5 and 6 list estimated annual inhalation intakes during rolling assigned to workers in each 

category. Because air concentrations were not available for Groups II and III in March, the March-toFebruary ratio of the Group I job category geometric mean air concentration, 4, was used to estimate 

a Group II and III March 30, 1949, air concentration.

Table 5. Inhalation exposures during rolling operations for Group I workers.

Work period

Number of 

months

Number of 

potential AEC 

workdays

Air concentration 

(pCi/m3)

Breathing 

rate (m3/h)

Hours 

worked 

per day Intake (pCi)

7/23/1948-3/29/1949 8 80 1,180 1.2 10 1.13E+6

3/30/1949-2/28/1950 11 110 216 1.2 8 2.28E+5

Total 1.36E+6

Table 6. Inhalation exposures during rolling operations for Group II and III workers.

Work period

Number of 

months

Number of 

potential AEC 

workdays

Air concentration 

(pCi/m3)

Breathing 

rate (m3/h)

Hours 

worked 

per day Intake (pCi)

7/23/1948-3/29/1949 8 80 274 1.2 10 2.63E+5

3/30/1949-2/28/1950 11 110 68.5 1.2 8 7.23E+4

Total 3.35E+5

There was a potential for internal exposure to resuspended material from the AEC work during 

non-AEC operations. To estimate exposure from resuspended materials, this analysis assumed that 

surfaces in the building became contaminated by deposition of uranium dust during rolling operations. 

The level of contamination was determined by multiplying the largest air concentrations, listed in 

Table 5, by the indoor deposition velocity and the assumed deposition time. The indoor deposition 

velocity is dependent on the physical properties of the room (such as air viscosity and density, 

turbulence, thermal gradients, and surface geometry). It is also dependent on the physical properties 

of the aerosol particles (such as diameter, shape, and density). In this case, these characteristics are 

not known, so the terminal settling velocity was calculated for an aerosol with the ICRP Publication 66 

default particle size distribution of 5-μm activity median aerodynamic diameter (ICRP 1994). The 

calculated terminal settling velocity was 0.00075 m/s, which is within the range of deposition velocities 

(2.7 × 10-6 to 2.7 × 10-3 m/s) measured in various studies (NRC 2002a).

The calculated surface contamination level created from airborne dusts during the uranium rolling 

from July 23, 1948, to February 28, 1950, was 5.35 × 106 pCi/m2 (119,000 dpm/100 cm2). The 

claimant-favorable assumption was made that all of the surface contamination was present for the 

entire period of AEC operations. Thus, using a resuspension factor of 1 × 10-6

/m (NRC 2002b), the 

air concentration due to resuspension would have been 5.35 pCi/m3

. Table 7 lists the assumed

annual inhalation intake received from resuspension of deposited material. The intakes listed in Table 

7 are added to the intakes listed in Table 5 or 6 before calculation of annual organ dose: this is done 

in Table 9.
Effective Date: 12/21/2004 Revision No. 00 Document No. ORAUT-TKBS-0021 Page 15 of 25

Table 7. Annual inhalation exposure during non-AEC operations due to resuspension of deposited 

uranium dust.

Work period

Hours 

per day 

Workdays per 

work period

Breathing 

rate (m3/h)

Resuspended air 

concentration (pCi/m3) Intake (pCi)

7/23/1948-3/29/1949 10 179 1.2 5.35 1.15E+4

3/30/1949-2/28/1950 8 240 1.2 5.35 1.23E+4

Total 2.38E+4 

In the case where inhalation intakes are calculated from air concentrations, ingestion intakes are also 

to be considered. NIOSH (2004) states that the daily ingestion rate in picocuries can be estimated by 

multiplying the daily air concentration in picocuries per cubic meter by a factor of 0.2 for an 8-hour

workday. For a 10-hour workday, the multiplier would be 0.223. The daily ingestion rates during AEC

uranium work are estimated from the air concentrations in Table 5, which are larger than the 

concentrations in Table 6. The daily ingestion intakes from resuspended uranium are estimated from 

Table 7. The ingestion intake rates are multiplied by the number of workdays exposed at the 

calculated levels and summed. The ingestion intakes listed in Table 8 apply to workers in Groups I, II, 

and III. 

Table 8. Estimated amount of uranium ingested (pCi) (based on Tables 5 and 7).

Work period

Days of 

uranium 

rolling

Uranium ingestion 

rate (during uranium 

rolling) pCi/d

Nonuranium 

rolling 

workdays

Uranium ingestion 

rate (during normal 

operation) pCi/d Intake pCi

7/23/1948-3/29/1949 80 2.62E+02 99 1.19 2.12E+4

3/30/1949-2/28/1950 110 4.32E+01 130 1.07 4.89E+3

Total 2.62E+4

3.2 OCCUPATIONAL INTERNAL DOSE RECONSTRUCTION ASSUMPTIONS AND 

SUMMARY

The assumed uranium photofluorimetry urinalysis MDA is 14 μg/L. 

The assumed operational exposure period ran from July 23, 1948, to February 23, 1950. The 

uranium-rolling period was over by March 31, 1949, and the analysis assumed that the exposure after 

that time was due to cleanup activities, which might have resulted in changing uranium intake rates by 

some workers. If the limited bioassay data are used to calculate intakes, the assignment of the 

exposure period needs to be considered carefully, because assuming that bioassay results are 

collected during periods of elevated exposure, when the result is actually collected days or months 

after the period of elevated exposure, could result in underestimation of intake. For individuals

unlikely to have been involved in cleanup, but who have positive bioassay results, it would be 

reasonable to set the intake period to the period of uranium rolling operations (July 23, 1948, through 

March 30, 1949) and to add an additional intake based on the Table 9 intakes for the period after the 

last bioassay. 

For unmonitored workers or unmonitored periods, Table 9 lists intake rate assumptions for natural 

uranium. The intake mode is chronic. The dose distribution is assumed to be lognormal with a GSD 

of 5.3.

Table 9. Internal exposure summary for operational period July 23, 1948, to February 28, 

1950.

Start End

Intake 

route

Absorption 

type Intake (pCi/day)
Effective Date: 12/21/2004 Revision No. 00 Document No. ORAUT-TKBS-0021 Page 16 of 25

Group I and unknown 7/23/1948 3/29/1949 Inhalation M, S 4,610

7/23/1948 3/29/1949 Ingestion (a) 85.1

3/30/1949 2/28/1950 Inhalation M,S 731

3/30/1949 2/28/1950 Ingestion (a) 14.6

Group II and III 7/23/1948 3/29/1949 Inhalation M, S 1,120

7/23/1948 3/29/1949 Ingestion (a) 85.0

3/30/1949 2/28/1950 Inhalation M,S 267

3/30/1949 2/28/1950 Ingestion (a) 14.6

a. Choose same f1-value as used for inhalation per NIOSH (2004).

4.0

Individual external dosimetry results for Vulcan Crucible are not available. If Vulcan Crucible 

individual external dosimetry results are found, dose reconstructors should consider those results in 

the evaluation of external dose.

ESTIMATION OF EXTERNAL EXPOSURE

For dose reconstruction, this analysis assumed that there was a potential for external exposure to 

natural uranium metal from five sources: 

• Submersion in air contaminated with uranium dust,

• Exposure from contaminated surfaces,

• Exposure to electrons from the surface of the uranium billets and rods,

• Exposure to photons from the uranium billets and rods, and

• Exposure to an annual diagnostic X-ray.

The majority of photons from natural uranium metals are in the 30 to 250 keV range. Solid uranium 

objects provide considerable shielding of the lower energy photons and harden the spectrum, causing 

the majority of photons emitted from a solid uranium object, such as a billet or a rod, to have energies 

greater than 250 keV. While it is recognized that solid uranium sources will have a hardened photon 

spectrum, exposure to a thin layer of uranium on a surface will result in a larger fraction of exposure to 

lower energy photons. The claimant favorable assumption for this analysis is workers were exposed 

to photon energies in the 30 to 250 keV range. Nonpenetrating dose from natural uranium consists 

primarily of electrons with energies >15 keV. For consistent presentation, exposure or dose is 

reported as either penetrating or nonpenetrating and is assumed to be associated with photons of 

energies 30 keV or greater, and with photons of energies less than 30 keV or with electrons, 

respectively. 

4.1 SUBMERSION AND CONTAMINATION EXPOSURES

In a survey at Simonds Saw and Steel, the AEC suspended 20 film badges about 5 feet from the floor 

in the rolling mill for 192 consecutive hours “to determine the long term direct [external] radiation to 

individuals” (Author unknown 1949). When the badges were retrieved, they were covered with 

radioactive dust from the plant, which would probably result in an overestimate of the true area 

radiation levels. The maximum results were reported as 5.6 mR/h beta and 0.34 mR/h gamma. The 

results of these measurements are assumed to be representative of the general levels of external 

exposure from submersion in air and contaminated surfaces at Vulcan Crucible. This analysis 

assumed that the data distribution was lognormal. The calculated geometric means were 1.3 mR/h

with a GSD of 2.3 for the nonpenetrating radiation, and 0.26 mR/h with a GSD of 1.2 for the 

penetrating radiation. This assumption does not appear to be inconsistent with the reported Zeuto 

(portable ionization chamber) beta and gamma readings at Simonds Saw and Steel of 2 mR/h or less 

for most areas (Author unknown 1949), some of which appear to be contact readings. The beta 
Effective Date: 12/21/2004 Revision No. 00 Document No. ORAUT-TKBS-0021 Page 17 of 25

reading is assumed to be related to the nonpenetrating dose and the gamma reading is assumed to 

be related to the penetrating dose. These assumed exposures at Vulcan Crucible during operational 

years are listed in Table 12. This analysis assumed that all workers were exposed to penetrating and 

nonpenetrating radiation from submersion in air and contamination for each workday for 10 hours/day 

prior to March 30, 1949, and 8 hours thereafter.

4.2 URANIUM BILLET AND ROD EXPOSURES

Another assumption was that workers received a deep dose due to photon exposure from the uranium 

billets and rods. According to reports, the AEC work involved rolling uranium billets 4 to 5 in. in 

diameter and 15 to 28 in. long. The billets were rolled into rods approximately 1.5 in. in diameter and 

20 ft long. Monte Carlo N-Particle (MCNP) calculations determined the photon (including 

bremsstrahlung) dose rate at the surface, 1 ft, and 1 m from a 5-in.-diameter by 28-in.-long cylindrical 

billet and a 1.405-in.-diameter by 20-ft-long rod. Table 10 lists calculated photon dose rates for the 

uranium billet and rod.

Table 10. Calculated photon dose rate for uranium billet 

and uranium rod.

Distance from source

Billet dose rate

(mrem/h)

Rod dose rate

(mrem/h)

Surface 7.74 5.09

1 ft 0.703 0.285

1 m 0.108 0.0883

This analysis assumed that Group I workers were exposed primarily to the billet dose rate and that 

Group II workers were exposed primarily to the rod dose rate. It also assumed that the dose rate at 

1 ft was the median dose rate, and the dose rate at the surface was the upper 95th percentile. 

The annual penetrating dose rates listed in Table 12 were calculated by multiplying the median 

photon dose rates by the number of rolling days per year and the number of work-hours, 10 hours/day 

prior to March 30, 1949, and 8 hours thereafter.

Shallow dose from the uranium billets and rods were estimated using the measurements in Table 11. 

These measurements were taken during an AEC survey in September 1948 (Belmore 1948).

Table 11. Direct radiation measurements from September 1948.

a

Location of measurement

Dose rate

(mrep/h)b

GROUP I

Contact with floor next to the quench tank where oxide scale has collected 8

Contact with floor in front of rolls where oxide scale has collected 5-10

Same location but 18” high 2-5

GROUP II

4 ft. above a pile of rods in the boxcar 20

5 ft. from the end of a pile of rods next to the door of the boxcar 5

2 ft. from the end of the same pile 13

a. Belmore 1948

b. A rep is an obsolete unit of dose equivalence (roentgen-equivalent-physical) approximately equal to a 

rem.

This analysis estimated the shallow dose for Group I by assuming that the median dose rate was 5

mrem/h and that the upper 95th percentile dose rate was 10 mrem/h, giving a GSD of 1.5. For Group 
Effective Date: 12/21/2004 Revision No. 00 Document No. ORAUT-TKBS-0021 Page 18 of 25

II, the assumed median dose rate was 5 mrem/h, and the assumed upper 95th percentile dose rate 

was 20 mrem/h, giving a GSD of 2.3. These exposure rates were multiplied by the assumed number 

of uranium rolling hours in the period. Table 12 lists these doses for Group I and II workers. The 

analysis assumed that Group III workers were unlikely to be in close contact with the rods and billets 

for extended periods; it also assumed that air and surface external exposures account for their 

external exposure. 

4.3 OCCUPATIONALLY REQUIRED MEDICAL X-RAY

X-ray machine characteristics, beam measurements, and example films for Vulcan Crucible were 

unavailable. The analysis assumed that workers received a preemployment occupationally related 

diagnostic PA chest X-ray. The specification for a preemployment X-ray was a “standard 14” x 17” 

film.” X-rays of the chest were to be repeated annually (Belmore 1948). Organ doses can be 

obtained from the current revision of ORAUT-OTIB-0006, Technical Information Bulletin: Dose 

Reconstruction from Occupationally Related Diagnostic X-Ray Procedures (ORAU 2003) for these 

pre-1970 examinations.

4.4 MISCELLANEOUS INFORMATION RELATED TO EXTERNAL DOSE

This section includes external dose information that might be of interest for specific dose 

reconstructions, but that this analysis did not consider generically because of its limited applicability or 

because of limited information.

In September 1948, AEC (1948b) noted that some workers were smoking while wearing dirty gloves 

and one worker was observed sitting on billets during a break. These activities are not directly 

considered in the external dose evaluations. They were observed early in the Vulcan Crucible

uranium rolling operations and might have been limited occurrences. 

During the 1949 decontamination activities, an AEC inspector noted:

There were 15 pairs of shoes in two steel drums in the storeroom of the plant. The soles of the shoes 

gave an average reading of 14,000 alpha d/m and 1.3 mr/hr. beta-gamma. The leather tops of the 

shoes showed an average reading of 5,000 alpha d/m and less than [0].4 mr/hr. beta-gamma. Inside 

the shoes there were negligible alpha and beta-gamma readings. (Klevin 1949)

4.5 OCCUPATIONAL EXTERNAL DOSE RECONSTRUCTION ASSUMPTIONS AND 

SUMMARY

Table 12 summarizes occupational external doses during uranium operations at Vulcan Crucible.

Table 12. External exposure summary for operational period July 23, 1948, to February 28, 1950.

Workers

Exposure 

mode

Exposure 

type

Exposure 

or dose 

rate Basis

Exposure 

time 

assumption Year

Annual 

exposure

IREP 

distribution

All

Submersion/

area 

contamination

Penetrating 0.26

mR/h Film badge 2000

work-h/y

1948

1949

1950

0.302 R

0.576 R

0.087 R

Lognormal

GSD 1.2

Nonpenetrating1.3

mR/h Film badge 2000

work-h/y

1948

1949

1950

1.508 R

2.878 R

0.437 R

Lognormal

GSD 2.3

Medical X-ray Pre-1970 PA

Initial plus 

one exam 

per year

1948

1949

1950

See ORAUT-OTIB-0006, 

(ORAUT 2003)
Effective Date: 12/21/2004 Revision No. 00 Document No. ORAUT-TKBS-0021 Page 19 of 25

Group I U billets

Penetrating 0.703 

mrem/h

MCNP 

calculation

3

h/rolling-day

1948

1949

1950

0.105 rem

0.253 rem

0.042 rem

Lognormal

GSD 4.2

Nonpenetrating5

mrep/h

Instrument 

measurement

3

h/rolling-day

1948

1949

1950

0.750 rem 

1.800 rem 

0.300 rem

Lognormal

GSD 2.7

Group II

and

unknown 

U rods

Penetrating 0.285 

mrem/h

MCNP 

calculation

7

h/rolling-day

1948

1949

1950

0.100 rem 

0.239 rem 

0.040 rem

Lognormal 

GSD 5.7

Nonpenetrating5

mrem/h

Instrument 

measurement

7

h/rolling-day

1948

1949

1950

1.750 rem

4.200 rem 

0.700 rem

Lognormal 

GSD 2.3

5.0

After the shutdown of AEC rolling operations, an AEC September 1949 letter described a July 1949 

survey, which identified areas that needed cleaning, but did not report quantifiable results (Belmore 

1949a). A survey of the site on December 30, 1949, identified areas that still needed cleanup, but 

again provided no numbers (Author unknown 1950). This analysis assumed that the residual 

exposure period began on March 1, 1950. A May 1978 survey by Argonne National Laboratory

reported 14 areas where the contamination exceeded the criterion of 5,000 dpm/100 cm2

, of which 

1,000 dpm/100 cm2 can be classified as removable. In general, the areas were between 500 and 

2,000 cm2 with one area as large as 5,000 cm2

. The greatest contamination was found on a small 

patch of dirt floor of about 100 cm2

; the activity levels in this area were 2.2 × 105 dpm/100 cm2

beta-gamma and 1.1 × 104 dpm/100 cm2 alpha, and were due to natural uranium. The contact 

Geiger-Muller end window detector reading at this location was 2.0 mR/h, and the reading at 1 m was 

at the background level (0.03 to 0.05 mR/h). Contact beta-gamma exposure rates throughout the 

facility varied between 0.1 and 2 mR/h with a median of 0.4 mR/h. Beta-gamma exposure rates at 1 

m were reported as background, which ranged from 0.03 to 0.05 mR/h (Wynveen et al. 1982). 

ESTIMATION OF RESIDUAL EXPOSURE

Vulcan Crucible received essentially pure uranium metal (no radium) for processing. This is 

confirmed by the 1978 Argonne National Laboratory survey (Wynveen et al. 1982), which showed 

radon levels varying between 0.11 and 0.27 pCi/L (0.0.0011 to 0.0027 Working Level assuming 100% 

equilibrium), which are within the normal range of atmospheric radon content of 0.1 to 0.5 pCi/L 

(Eisenbud 1987). The building was reportedly evacuated in 1978 (Perry 1993); this is assumed to

have occurred on May 31, 1978, after the Argonne National Laboratory survey.

The Aliquippa Forge site was included in the DOE FUSRAP in August 1983. In December 1987, 

Bechtel National, Inc. surveyed Aliquippa Forge for the purpose of allowing the use of portions of 

Building 3 for storage. Additional remediation in the fall of 1988 to allow expansion of the forging 

program resulted in removal of contaminated materials and equipment and barricading of the 

remaining Building 3 contamination. This analysis assumed that the second exposure period to 

residual contamination started on December 1, 1987. DOE noted that access to the contaminated 

areas was not allowed (Seay 1988). 

In 1992 and 1993, areas in and adjacent to Buildings 3 and 8 were further characterized (Abelquist 

1994, Adams 1992, DOE 1996). It was reported that as of May 17, 1993, the building was no longer 

in use, so this analysis assumed that date to be the end of the second residual contamination 

exposure period. In 1992, the maximum reported exposure rate at 1 meter was equal to 0.014 mR/h 

(Adams et al 1992a). A few spots of contamination were found to approach 1,000,000 dpm/100 cm2

, 

but most areas of contamination were less than 5,000 dpm/100 cm2

. The greatest removable 

contamination was 410 dpm/100 cm2

, but the majority of areas checked for removable contamination 

showed beta results less than 16 dpm/100 cm2 and alpha results less than 12 dpm/100 cm2

. The 
Effective Date: 12/21/2004 Revision No. 00 Document No. ORAUT-TKBS-0021 Page 20 of 25

largest grid-block-average was a floor area having 230,000 dpm/100 cm2 (Adams and Payne 1992b, 

p. 37-40). Most of the contamination results were less than this amount by an order of magnitude or 

more. In addition, not all areas of the buildings were contaminated. 

To calculate internal exposure from residual activity this analysis assumed that the building was 

uniformly contaminated to 5% of the largest grid-block-average value, which would be 

11,500 dpm/100 cm2

. Using a resuspension factor of 1 × 10-6/m (NRC 2002b) and an air intake rate 

of 2,400 m3 per work year, the calculated annual inhalation intake was 1,240 pCi. Using the method 

described in Section 3.0, the calculated annual ingestion intake was 25.9 pCi. Table 13 summarizes 

the intake rates. 

To reconstruct external exposure to residual radioactivity after the end of AEC operations, this 

analysis assumed that the upper 95th percentile worker penetrating exposure was 0.05 mR/h, which 

was the upper end of the exposure rate readings at 1 m in 1978 (this reading could be interpreted as 

the minimum recording level for this type of measurement). The median penetrating exposure rate 

was assumed to be 0.025 mR/h, one-half of the 95th percentile. A GSD of 1.5 was calculated for the 

assumed lognormal distribution. This exposure bounds the 1-m exposure rate measurements, 0.007 

to 0.014 mR/h, reported by Adams (1992a) in 1992, which were measured after limited remedial 

activities in 1988. 

The non-penetrating exposure rate was determined by assuming that 5.0, the ratio of non-penetrating 

to penetrating exposure rates for submersion and contamination external exposures during the 

operational exposure period, provided a reasonable estimate of the ratio of non-penetrating to 

penetrating exposure rate during the residual exposure period. The resulting non-penetrating 

exposure rate is 0.125 mR/h. A GSD of 2.6 for the non-penetrating dose rate was calculated by 

raising the number, e, to the square root of the sum of the squares of the lognormals of the GSDs of 

the three distributions (2.3, 1.2 and 1.5) used to calculate the non-penetrating dose rate:

2

3

2

2

2

1 (ln( )) (ln( )) (ln( )) g g g

GSD e

σ + σ + σ

=

The estimated annual penetrating and non-penetrating external exposures to residual radioactivity 

from AEC operations at the site, listed in Table 13, were calculated by assuming that workers were 

exposed for 2,000 hours per year. 

Assumptions regarding residual exposures should be consistent with assumptions from the 

operational period.

Table 13. Annual internal and external exposure to residual radioactivity.

Exposure Start End Exposure

Absorption 

type

Intake

(pCi/d) IREP distribution

Internal

3/1/1950 5/31/1978 Inhalation M, S 3.40 Lognormal GSD 3

3/1/1950 5/31/1978 Ingestion (a) 0.071 Lognormal GSD 3

12/1/1987 5/17/1993 Inhalation M,S 3.40 Lognormal GSD 3

12/1/1987 5/17/1993 Ingestion (a) 0.071 Lognormal GSD 3

Start End Exposure Basis R/y

External 3/1/1950 5/31/1978 Penetrating Survey Instrument 0.050 Lognormal GSD 1.5

12/1/1987 5/17/1993 Penetrating Survey Instrument 0.050 Lognormal GSD 1.5

3/1/1950 5/31/1978 NonPenetratingSurvey Instrument 0.250 Lognormal GSD 2.6

12/1/1987 5/17/1993 Non- Survey Instrument 0.250 Lognormal GSD 2.6
Effective Date: 12/21/2004 Revision No. 00 Document No. ORAUT-TKBS-0021 Page 21 of 25

Penetrating

a. Choose same f1-value as used for inhalation per NIOSH (2004).
Effective Date: 12/21/2004 Revision No. 00 Document No. ORAUT-TKBS-0021 Page 22 of 25

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