ORAU TEAM

Dose Reconstruction

Project for NIOSH

Oak Ridge Associated Universities I Dade Moeller & Associates I MJW Corporation

Page 1 of 46

Document Title:

An Exposure Matrix for W.R. Grace 

and Company in Erwin, Tennessee

Document Number: ORAUT-TKBS-0043

Revision: 01

Effective Date: 07/16/2008

Type of Document TBD

Supersedes: Revision 00

Subject Expert(s): Paul J. Demopoulos, Milton Gorden, and Donald Palmrose

Site Expert(s): N/A

Approval: Signature on File Approval Date: 06/18/2008

Karen S. Kent, Document Owner

Concurrence: Signature on File Concurrence Date: 06/18/2008

John M. Byrne, Task 3 Manager

Concurrence: Signature on File Concurrence Date: 06/18/2008

Edward F. Maher, Task 5 Manager

Concurrence: Signature on File Concurrence Date: 07/01/2008

Kate Kimpan, Project Director

Approval: Signature on File Approval Date: 07/16/2008

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-0043 Revision No. 01 Effective Date: 07/16/2008 Page 2 of 46

PUBLICATION RECORD

EFFECTIVE 

DATE

REVISION 

NUMBER DESCRIPTION

02/14/2006 00 New technical basis document for the W.R. Grace and Company in 

Erwin, Tennessee. First approved issue. Initiated by Paul J. 

Demopoulos.

07/16/2008 01 Approved revision to incorporate changes made from Task 5 

comments. Incorporates formal internal and NIOSH review 

comments. Incorporates SEC-00082 and a methodology for 

reconstruction dose during the residual period. Constitutes a total 

rewrite of the document. Training required: As determined by the 

Task Manager. Initiated by Karen S. Kent.
Document No. ORAUT-TKBS-0043 Revision No. 01 Effective Date: 07/16/2008 Page 3 of 46

TABLE OF CONTENTS

SECTION TITLE

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

PAGE

6

1.0 Introduction ..............................................................................................................................8

1.1 Purpose........................................................................................................................9

1.2 Scope...........................................................................................................................9

2.0 Site Description, History, and Process ................................................................................... 10

2.1 Process Description....................................................................................................11

2.1.1 Production of Uranium Metal and Uranium Metal Alloys Enriched 

up to 12% Uranium-235 .................................................................................. 12

2.1.1.1 Conversion of Uranium Hexafluoride to Uranium Metal .................... 12

2.1.1.2 Reduction of Uranium Tetrafluoride to Uranium Metal ...................... 13

2.1.1.3 Pickling............................................................................................. 13

2.1.1.4 Other Metal Products........................................................................ 13

2.1.1.5 Packaging and Shipping................................................................... 13

2.1.2 Scrap Recovery Operations ............................................................................ 13

2.1.3 Highly Enriched Uranium from Scrap Recovery .............................................. 14

2.1.3.1 Receiving and Storage ..................................................................... 14

2.1.3.2 Sampling .......................................................................................... 14

2.1.3.3 Head-End Operations....................................................................... 14

2.1.3.4 Final Process Steps.......................................................................... 14

2.1.4 Production of Uranium Oxide Mixed with Thorium Oxide and 

Zirconium Oxide.............................................................................................. 15

2.1.4.1 Receiving and Storage ..................................................................... 15

2.1.4.2 Solution Concentration ..................................................................... 15

2.1.4.3 Precipitation...................................................................................... 15

2.1.4.4 Drying and Grinding.......................................................................... 15

2.1.4.5 Calcination to Uranium Dioxide......................................................... 15

2.1.4.6 Blending ........................................................................................... 15

2.1.4.7 Diluent Addition and Blending........................................................... 15

2.1.4.8 Binder Addition ................................................................................. 15

2.1.4.9 Lubricant Addition and Pressing ....................................................... 15

2.1.4.10 Binder Removal and Sintering .......................................................... 16

2.1.4.11 Physical Measurement and Grinding ................................................ 16

2.1.4.12 Tube Loading ................................................................................... 16

2.1.5 Production of Mixed Oxide Fuel ...................................................................... 16

2.1.6 Uranium Hexafluoride Cylinder Washing......................................................... 16

3.0 Estimation of Internal Exposure.............................................................................................. 16

3.1 Uranium Exposures................................................................................................................17

3.1.1 Uranium Air Sampling ..................................................................................... 18

3.1.2 Enriched and Recycled Uranium..................................................................... 22

3.1.3 Uranium Bioassay........................................................................................... 23

3.2 Plutonium Exposures..................................................................................................23

3.2.1 Plutonium Bioassay ........................................................................................ 24

3.3 Uranium-233 Considerations ......................................................................................25

3.4 Internal Dose Reconstruction Assumptions ................................................................25
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4.0 Estimation of External Exposure ............................................................................................ 26

4.1 Limitations ..................................................................................................................26

4.2 Recorded External Dose.............................................................................................26

4.2.1 Nonpenetrating Dose ...................................................................................... 27

4.2.2 Penetrating Dose ............................................................................................ 28

4.3 Neutron Dose .............................................................................................................28

4.3.1 Extremity Doses.............................................................................................. 28

4.4 Missed Dose...............................................................................................................28

4.5 Dose Reconstruction Recommendations....................................................................29

4.5.1 Beta Dose....................................................................................................... 29

4.5.2 Photon Dose ................................................................................................... 30

4.5.3 Glovebox Geometry Assumptions................................................................... 30

4.5.4 Uranium-233 ................................................................................................... 30

4.6 Occupational Medical Dose ........................................................................................30

4.7 Summary of External Dose Reconstruction Recommendations .................................. 31

5.0 Estimation of Exposure to Residual Activity ........................................................................... 31

5.1 Internal Exposure to Residual Activity.........................................................................31

5.1.1 Introduction..................................................................................................... 31

5.1.2 Uranium .......................................................................................................... 32

5.1.3 Plutonium........................................................................................................ 33

5.1.4 Thorium........................................................................................................... 37

5.2 Estimation of External Exposure from Residual Radioactivity ..................................... 38

6.0 Environmental Dose...............................................................................................................41

7.0 Attributions and Annotations .................................................................................................. 41

References ........................................................................................................................................43
Document No. ORAUT-TKBS-0043 Revision No. 01 Effective Date: 07/16/2008 Page 5 of 46

LIST OF TABLES

TABLE TITLE

2-1 W.R. Grace and NFS AEC weapons-related work period and locations................................. 

PAGE

12

3-1 Fuel types, chemical forms, isotopes, and enrichments of W.R. Grace process 

material..................................................................................................................................17

3-2 Uranium source term information ........................................................................................... 17

3-3 Uranium-233 Shippingport Reactor fuel activity fractions ....................................................... 18

3-4 Occupational exposures for 93% uranium-aluminum alloy recovery....................................... 18

3-5 Average breathing-zone samples for 93% uranium-aluminum alloy recovery......................... 19

3-6 Average general air-sampling concentrations for 93% uranium-aluminum alloy 

recovery ................................................................................................................................. 19

3-7 Occupational exposures for 3.6% uranium recovery from slag............................................... 20

3-8 Average breathing-zone samples for 3.6% uranium recovery from slag................................. 20

3-9 Average general air-sampling concentrations for 3.6% uranium recovery from slag............... 21

3-10 Estimated uranium inhalation intake rates based on measured time-weighted air 

concentrations during uranium recovery operations ............................................................... 22

3-11 Estimated uranium ingestion rates based on measured time-weighted air 

concentrations during uranium recovery operations ............................................................... 22

3-12 Intakes of contaminants in RU as fractions of uranium intake ................................................ 22

3-13 Activity composition of Hanford Site reference fuel-grade plutonium mixture.......................... 24

3-14 Uranium internal exposure summary for operational period January 1, 1958, to 

December 31, 1970................................................................................................................25

4-1 Dosimetry for external whole-body, wrist, and extremity exposures ....................................... 27

4-2 MDL and maximum potential annual missed photon or beta dose.......................................... 29

4-3 IREP input summary ..............................................................................................................31

5-1 Uranium air concentration and annual intake in the residual period........................................ 32

5-2 Plutonium urinalysis results from operational period as of April 17, 2007 ............................... 33

5-3 Summary of excretion rates from plutonium urinalysis results from operational period........... 34

5-4 Air-sampling results in percentage of MPCs during the residual period, 1976 to 1989 ........... 36

5-5 Plutonium air concentration and annual intake in the residual period ..................................... 37

5-6 Thorium air concentration and annual intake in the residual period ........................................ 38

5-7 W.R. Grace/NFS external exposure summary........................................................................ 40

LIST OF FIGURES

FIGURE TITLE

2-1 Location of Erwin, Tennessee ................................................................................................

PAGE

10

2-2 Plant site at W.R. Grace/NFS................................................................................................. 11
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ACRONYMS AND ABBREVIATIONS

ADU ammonium diuranate

AEC U.S. Atomic Energy Commission

AP anterior-posterior

AWE atomic weapons employer

CFR Code of Federal Regulations

Ci curie

cm centimeter

d day

DCF dose conversion factor

DL decision level

DOE U.S. Department of Energy

dpm disintegrations per minute

DU depleted uranium

EEOICPA Energy Employees Occupational Illness Compensation Program Act of 2000

EU enriched uranium

F fast (absorption type)

g gram 

gal gallon

GM geometric mean

GSD geometric standard deviation

HASL Health and Safety Laboratory

HEU highly enriched uranium

hr hour

ICRP International Commission on Radiological Protection

IMBA Integrated Modules for Bioassay Analysis

in. inch

IREP Interactive RadioEpidemiological Program

keV kiloelectron-volt, 1,000 electron-volts

kg kilogram

L liter

lb pound

LEU low-enriched uranium

LOD limit of detection

m meter

M moderate absorption type

MDA minimum detectable amount or activity

MDL minimum detectable level

MeV megaelectron-volt, 1 million electron-volts

mg milligram

mL milliliter

MOX mixed oxide
Document No. ORAUT-TKBS-0043 Revision No. 01 Effective Date: 07/16/2008 Page 7 of 46

MPC maximum permissible concentration

mR milliroentgen

mrem millirem

nCi nanocurie

NFS Nuclear Fuel Services

NIOSH National Institute for Occupational Safety and Health

NRC U.S. Nuclear Regulatory Commission

ORAU Oak Ridge Associated Universities

PA posterior-anterior

pCi picocurie

PFG photofluorography

POC probability of causation

R roentgen

RU recycled uranium

s second

S slow (absorption type)

SEC special exposure cohort

SEFOR South-West Experimental Fast Oxide Reactor

SRDB Ref ID Site Research Database Reference Identification (number) 

TLD thermoluminescent dosimeter

U.S.C. United States Code

yr year

μCi microcurie

μg microgram

§ section or sections
Document No. ORAUT-TKBS-0043 Revision No. 01 Effective Date: 07/16/2008 Page 8 of 46

1.0

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.

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

Employees Occupational Illness Compensation Program Act of 2000 [EEOICPA; 42 U.S.C. § 7384I(5) 

and (12)]. EEOICPA, as amended, provides for employees who worked at an AWE facility during the 

contract period and/or during the residual period.

Under EEOICPA, employment at an AWE facility is categorized as either (1) during the DOE contract 

period (i.e., when the AWE was processing or producing material that emitted radiation and was used 

in the production of an atomic weapon), or (2) during the residual contamination period (i.e., periods 

that NIOSH has determined there is the potential for significant residual contamination after the period 

in which weapons-related production occurred). For contract period employment, all occupationally 

derived radiation exposures at covered facilities must be included in dose reconstructions. This 

includes radiation exposure related to the Naval Nuclear Propulsion Program and any radiation 

exposure received from the production of commercial radioactive products that were concurrently 

manufactured by the AWE facility during the covered period. NIOSH does not consider the following 

exposures to be occupationally derived (NIOSH 2007a):

• 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

For employment during the residual contamination period, only the radiation exposures defined in 42 

U.S.C. § 7384n(c)(4) [i.e., radiation doses received from DOE-related work] must be included in dose 

reconstructions. Doses from medical X-rays are not reconstructed during the residual contamination 

period (NIOSH 2007a). It should be noted that under subparagraph A of 42 U.S.C. § 7384n(c)(4), 

radiation associated with the Naval Nuclear Propulsion Program is specifically excluded from the 

employee’s radiation dose. This exclusion only applies to those AWE employees who worked during 

the residual contamination period. Also, under subparagraph B of 42 U.S.C. § 7384n(c)(4), radiation 

from a source not covered by subparagraph A that is not distinguishable through reliable 

documentation from radiation that is covered by subparagraph A is considered part of the employee’s 

radiation dose. This site profile covers only exposures resulting from nuclear weapons-related work. 

Exposures resulting from non-weapons-related work, if applicable, will be covered elsewhere.

A decision has been made that internal dose from thorium cannot be reconstructed with sufficient 

accuracy for W.R. Grace employees during the period from 1958 through 1970. Therefore only 

internal dose from uranium and plutonium is considered in Section 3.0 of this site profile. The Special 

Exposure Cohort (SEC) evaluation report (NIOSH 2007b) defines a single class of employees for 

which NIOSH cannot estimate radiation doses with sufficient accuracy. This class includes all AWE 

employees who were monitored or should have been monitored for potential exposure to thorium 

while working at the W.R. Grace site at Erwin, Tennessee, for a number of workdays aggregating at 
Document No. ORAUT-TKBS-0043 Revision No. 01 Effective Date: 07/16/2008 Page 9 of 46

least 250 workdays from January 1, 1958, through December 31, 1970, or in combination with 

workdays within the parameters established for one or more other classes of employees in the SEC. 

This site profile provides specific information on documentation of historical practices at the W.R. 

Grace and Company plant. If the dose reconstructor cannot determine whether the work was U.S. 

Atomic Energy Commission (AEC) weapons-related work or non-AEC-related work, the dose 

reconstructor should assume that all work was AEC weapons related to ensure favorability to 

claimants.

1.1 PURPOSE

This site profile provides technical basis information to be used to evaluate the total occupational 

radiation dose that can reasonably be associated with a worker’s radiation exposure at the W.R. 

Grace and Company plant. This dose results from exposure to external and internal radiation sources 

in W.R. Grace facilities and to occupationally required diagnostic X-ray examinations. This site profile

includes methods for estimating doses that could have occurred while an employee was not 

monitored or inadequately monitored, or that were missed due to analytical detection limits or 

incomplete or missing monitoring records (i.e., missed dose). 

Dose reconstructors use the NIOSH Interactive RadioEpidemiological Program (IREP) and the 

Integrated Modules for Bioassay Analysis (IMBA) computer programs to evaluate radiation doses.

Information on measurement uncertainties is an integral component of the NIOSH approach to dose 

reconstruction. This site profile describes how to evaluate uncertainty in relation to W.R. Grace 

exposure and dosimetry records.

1.2 SCOPE

Sections 2.1 and 2.2 detail major W.R. Grace facilities and operations. W.R. Grace began processing 

materials in the late 1950s at its Erwin, Tennessee, site. The principal operation at the site has been 

to convert highly enriched uranium (HEU) and low-enriched uranium (LEU) from UF6 or to recover

uranium from scrap to a product that meets customer requirements. Thorium, depleted uranium (DU), 

233U, recycled uranium (RU), and plutonium have also been processed at various times to oxides or 

metals with subsequent processing into the form necessary for the manufacture of nuclear fuel. In 

addition, the recovery of HEU from uranium-aluminum alloy occurred in the storage, chemical, and 

ceramic buildings with sampling being conducted in the metallurgical chemistry and spectrographic 

laboratories (AEC 1959, p.6).

Section 3.0, for the estimation of internal exposure, describes the internal dosimetry program at W.R. 

Grace. Workers handled a variety of radionuclides as part of their routine tasks. The key elements in 

the source term are uranium, plutonium, and thorium.

During the AEC contract years from the late 1950s though 1970, the monitoring methodology for W.R. 

Grace employees for intakes of radionuclides was bioassay, with the primary focus being uranium

urinalysis. The UO2/PuO2 mixed-oxide (MOX) facility is believed to have been completed in 1965, and 

plutonium processing began in 1966, when plutonium was analyzed by urine bioassay. Few feces 

samples in the 1960s were collected for uranium and plutonium. Air monitoring in workplaces and in 

the breathing zones of employees has been a common surveillance method. Although W.R. Grace

apparently did not use the data acquired from the air monitoring program to prepare the dose of 

record for employees; these data have been used in this site profile in certain instances for individual 

exposure evaluations. In vivo monitoring began in 1970, with capabilities that focused on uranium 

monitoring, which expanded to plutonium and thorium monitoring in the 1990s. There was not a 

substantial amount of in vivo monitoring in the AEC contract years.
Document No. ORAUT-TKBS-0043 Revision No. 01 Effective Date: 07/16/2008 Page 10 of 46

Section 4.0, for the estimation of external exposure, describes the external dosimetry program at W.R. 

Grace, and includes electron, photon, neutron, and occupational medical X-ray dose. W.R. Grace

utilized Nuclear Chicago and then Landauer for its external dosimetry program. The program 

included whole-body, wrist, and extremity film badges for photon and beta monitoring. Neutron 

monitoring was not conducted at W.R. Grace during the AEC contract years. Thermoluminescent 

dosimeters (TLDs) were provided by Landauer in 1989. 

Section 5.0, for the estimation of exposure to residual activity, discusses external and internal dose 

estimates from surface contamination. The exposures could have occurred from resuspension of 

removable contamination from AEC contract year operations and ongoing operations. Section 6.0 is 

the environmental section.

Attributions and annotations, indicated by bracketed callouts and used to identify the source, 

justification, or clarification of the associated information, are presented in Section 7.0.

2.0

The Davison Chemical Company, a division of W.R. Grace, began processing radioactive materials in 

the late 1950s at the site of the current Nuclear Fuel Services (NFS) facility near Erwin, Tennessee 

(NFS 2005). 

SITE DESCRIPTION, HISTORY, AND PROCESS

The Report to Joint Committee on Atomic Energy Congress of the United Stated by the United States 

General Accounting Office (GAO 1967, p.5-6) from September 14, 1967, states:

AEC authorized the Davison Chemical Division of W.R. Grace & Company (Grace) to receive 

and process special nuclear material under license number SNM-124. Grace received its first 

material as an AEC licensee by lease agreement in March 1958 and its first nuclear material as 

an AEC contractor in May 1959. 

In 1964, NFS was formed by the merger of W.R. Grace and the American Machine and Foundry 

Company. In 1965, the UO2/PuO2 MOX facility was completed, and in 1968, the light-water breeder 

reactor thorium, thoria, and 233U oxide program was initiated (NFS 2005). Ownership of NFS 

transitioned from Getty Oil to Texaco as part of a stock buyout of Getty Oil in 1969. In 1987, NFS 

Services, a private Atlanta-based limited partnership purchased NFS from Texaco. Figure 2-1 shows 

the location of Erwin, Tennessee. Figure 2-2 shows the principal site infrastructure for the NFS facility 

as of 1986 (Congress 1986, p.108).

Figure 2-1. Location of Erwin, Tennessee. 

Regulatory authority over operations was originally under the AEC (1954 to 1974) and transitioned to 

the U.S. Nuclear Regulatory Commission (NRC) (1975 to present) under Special Nuclear Material 

License SNM-124, as amended. 
Document No. ORAUT-TKBS-0043 Revision No. 01 Effective Date: 07/16/2008 Page 11 of 46

Figure 2-2. Plant site at W.R. Grace/NFS. 

The principal operation at the site has been to convert HEU and LEU from UF6 to a product that meets 

customer requirements. Thorium, DU, 233U, and plutonium have also been processed at various times 

to oxides or metals with subsequent processing into the form necessary for the manufacture of 

nuclear fuel. Specific tasks at the facility included (1)processing ThO2 that was mixed with 233U to 

make the light-water breeder reactor fuel for the Shippingport Reactor and (2) fabricating plutonium 

and DU MOX fuel for the South-West Experimental Fast Oxide Reactor (SEFOR) in Building 234 with 

laboratory capabilities in Building 110; both operations ceased in 1970 (Congress 1986, p.31). Other 

operations were associated with the conversion operations such as scrap recovery operations 

(uranium and other nuclear fuel material) and cleaning and certification of empty cylinders for

transport of LEU UF6. Table 2-1 lists the approximate processing history of the site.

2.1 PROCESS DESCRIPTION

Brief summaries for the principal operations are given below, and additional details can be found in 

the series of Feasibility Reports in the cited references. Inherent in all of the operations is nuclear 

criticality safety that governs not only the operations and storage but also the movement of material 

within the facility. Nuclear criticality safety is maintained at the facility through the control of one or 

more of the conventional parameters of geometry, mass, concentration, and control of neutron 

interaction between subcritical units. The standard administrative policy at this and other similar 

facilities is to control two such parameters whenever possible. For birdcage units, the two control 

parameters are geometry (birdcage dimensions) and a 235U mass limit that can be placed within the 

birdcage. Examples of birdcages that were used include a 233U storage birdcage that used a shielded 

55-gal drum that contained inserts with spacers; UF6 cylinders storage birdcage that used a 2-in. × 

2-in. frame made of metal that held the 5-in. UF6 cylinder in place at its center with a total dimension of 

52 in. high and 30 in. wide; and a U3O8 storage birdcage that used a 2- × 2-in. metal frame made of 

metal that held a 10.75-in. outside diameter × 12-in.-high cylinder in the center with a total dimension 

of 36 in. high and 30.625 in. wide (W.R. Grace 1959, p.23, Runion 1959, p.5, Glauberman 1962). A 

051905_01_TB Source: Erwin Nuclear Fuels Services. Hearing Before the Subcommittee on Energy Conservation and

Power of the Committee on Energy and Commerce House of Representatives. Serial No. 99-178.

Burial Sites

Waste Retention

Ponds

Nearest Residence

Clinchfield Railroad Main Track

Area Inside Security Fence: 21.2 Acres

Total Plant Site Area: 57.8 Acres

2

3

1

Property Boundary

Security Fence

Barbed Wire Fence
Document No. ORAUT-TKBS-0043 Revision No. 01 Effective Date: 07/16/2008 Page 12 of 46

review of the available literature shows that no criticality accidents have occurred during W.R. Grace 

and later NFS operations.

Table 2-1. W.R. Grace and NFS AEC weapons-related work period and locations. 

Operations

Period of AEC weaponsrelated worka Building location

Thorium (metal and oxide) 1/1/1958 through 12/31/1970 110C, 110D, 111, 130, 234B, 234C, 

310

Uranium metal (HEU and DU) 1/1/1958 through 12/31/1970 110, 110E, 111, 130, 135, 234, 301

LEU UO2 1/1/960 through 12/31/1970 301, 110E, 111, 130, 135, 301

U-233 fuel (sometimes mixed with 

ThO2)

01/01/1961 through 12/31/1970 234B, 234C, 110C, 110D

Plutonium fuelb and MOX 1/1/1966 through 12/31/1970 234A, 234B, 234C, 110C, 110D

HEU scrap recovery 1/1/1958 through 12/31/1970a 130, 220, 230, 233

HEU fuel 1/1/1966 through 12/31/1970 100, 105, 120, 131, 132/133, 220, 

233, 300, 302, 303, 304, 310, 330, 

301 

LEU scrap recovery 1/1/1960 through 12/31/1970 

(AEC 1961), or (Congress 

1986)

111, 220, 230, 233

LEU cylinder washing 1/1/1958 through 12/31/1970 111,130

Sources: AEC (1961), Congress (1986, pp. 107-114 and 127-143).

a. The end of 1970 is when AEC-weapons-related operations ended.

b. A review of the claimant files (to date) indicates that there are plutonium bioassay records that begin in 1967. Because

the plutonium facility was completed in 1965, it is reasonable to assume that plutonium operations occurred from 1966 

to 1970 (NFS 2005, p.2; Congress 1986, p.107).

2.1.1

The UF6 was supplied in approved standard cylinders and received in approved packaging such as 

birdcages. The cylinders were check-weighed and placed into storage in special concrete cells or 

birdcages in one of the warehouse buildings.

Production of Uranium Metal and Uranium Metal Alloys Enriched up to 

12% Uranium-235

2.1.1.1 Conversion of Uranium Hexafluoride to Uranium Metal

The overall process involves vaporization, reduction to UF4, reduction to uranium metal, pickling, 

processing into other metal products, packaging, and shipment. The following information is from 

Feasibility Report for the Production of Uranium Metal and Uranium Metal Alloys Enriched Up To 12% 

235U (Housholder 1963a, p.6-8).

UF6 was received in solid cylinders and had to be vaporized to transfer it to the UF6-to-UF4 reactor 

(the 6-to-4 unit). Vaporization was accomplished by heating the UF6 cylinders in an electric oven. Up 

to six cylinders could be placed into the oven for processing.

The UF6 was piped into the 6-to-4 unit where it was reduced with hydrogen. The solid UF4 powder 

dropped into a product hopper where it was metered into safe-diameter product cans. The offgas 

flowed through two cyclone separators where any entrained UF4 dust was collected in additional 

product cans. The offgas was then filtered to remove the last traces of UF4 dust and was scrubbed 

with potassium hydroxide to remove hydrogen fluoride vapors. The gas, free of uranium and acid, 

was vented to the atmosphere through a flame arrester where excess hydrogen was burned off.
Document No. ORAUT-TKBS-0043 Revision No. 01 Effective Date: 07/16/2008 Page 13 of 46

2.1.1.2 Reduction of Uranium Tetrafluoride to Uranium Metal

The UF4 was weighed into a reduction batch and blended with a reducing agent such as magnesium 

metal. The charge was then heated under vacuum in an induction furnace to form a uranium metal 

derby. After cooling, the derby was broken out and separated from the slag. The slag was packaged 

for scrap recovery where any remaining uranium in the slag was recovered.

2.1.1.3 Pickling

The uranium derby was pickled in acid to remove adherent slag and scale. The pickle solution was 

sent to scrap recovery for recovery of uranium. The pickled derby was then sampled for impurities.

2.1.1.4 Other Metal Products

As necessary, the derbies were broken into smaller pieces on a large hydraulic press before 

shipment. On other occasions, the derbies were remelted and cast into various shapes such as 

slugs, rods, and plates. The uranium could also be alloyed with other metals during remelting and 

casting. Melting occurred in a large vacuum induction furnace. Cast pieces could be pickled as 

previously described.

2.1.1.5 Packaging and Shipping

All enriched uranium (EU) metal products were packaged for shipment in approved birdcages. The 

products could be stored on the site for a time before shipment to the customer. 

2.1.2

This operation was designed to recover the scrap uranium from the manufacturing processes. The 

process steps were assumed to be carried out in a similar manner for scrap with either HEU or LEU 

(Katine 1960, p.18). The exception between handling the different enrichment levels would be in the 

dissolution and filtration steps. There was potential for different end processes that were adjusted 

based either on the customer’s desired final material form (for scrap recovery from outside sources) or 

on how the facility was going to reinsert the recovered uranium in the fuel fabrication process. 

Scrap Recovery Operations

In one particular job, dependent on whether the material was greater or less than 2% enrichment, two 

different batch sizes and dissolver tanks were used. For scrap material equal to less than 2% 

enrichment, a maximum batch of 87.95 kg uranium (220 lb) of UO2 is dissolved in a 500-gal capacity 

dissolver tank. For all other enrichments, a 20-gal dissolver tank was employed and the batch would 

contain a maximum of 400 g of 235U. This process would result in two types of solutions. There would 

be solutions of 2% or less enrichment at a concentration of less than 2 grams of 235U per liter. There 

were solutions ranging from 2.8 to 9.9% enrichment at a concentration of 5 grams of 235U per liter. 

From these two types of solutions, a third solution will be formed that will not exceed 3% enrichment, 

and will contain a maximum of 88 pounds of uranium (39.9 kilograms) which is the safe mass for 3% 

enrichment. 

Of note for this process, the memorandum that was attached to a Feasibility Report for HEU scrap 

recovery specifically states, “the chopping and sawing of solid metal plates, billets, and rods, and the 

operations of grinding, screening, and blending of the resultant oxides from the calcinations step, 

could be possible dust sources. Air samples should be obtained at these operations and evaluated at 

the start of operations to ascertain the effectiveness of the existing ventilation” (Glauberman 1962, 

p.117). A summary of the HEU scrap recovery follows (Housholder and Runion 1962, p.122-123).
Document No. ORAUT-TKBS-0043 Revision No. 01 Effective Date: 07/16/2008 Page 14 of 46

2.1.3

2.1.3.1 Receiving and Storage

Highly Enriched Uranium from Scrap Recovery

All scrap was received in approved shipping containers and birdcages and sent to storage 

warehouses to await processing.

2.1.3.2 Sampling

Before processing, each container was inspected to decide what head-end steps could be necessary 

before the scrap could be dissolved. In addition, at this time samples were taken for uranium assay to 

determine batch sizes.

2.1.3.3 Head-End Operations

Due to the many types of scrap the facility received, it was often necessary to pretreat the scrap 

before dissolution and subsequent extraction. Metal chips and turnings, for instance, were routinely 

received stored under oil, which had to be removed before dissolution. This was done by draining the 

oil off the metal in wire baskets. The metal was then washed with solvents and dried.

Solid metal plates and rods were chopped or sawed into smaller pieces to facilitate handling and 

weighing into batches. Combustible wastes such as filters, sponges, grinder sludge, sweepings, etc., 

were calcined to reduce their bulk and to remove hydrogenous and carbonaceous materials, such as 

water and oil. After calcining, the resultant oxides were ground and screened; material not passing 

the screen was recycled to the grinder. The oxides were then blended to ensure homogeneity for 

sampling for uranium assay, from which accountability and batch sizes could be determined.

Oxide pellets were ground and screened and could be calcined and blended to facilitate dissolution.

2.1.3.4 Final Process Steps

A review of the Feasibility Reports showed there were different final process steps that appear to 

have depended on the final material form needed either to be reinserted into the facility process or to 

be shipped back to the customer. The extracted uranium liquid (also known as O.K. Liquor) could be 

converted into a solid through either a boildown and crystallization process (Housholder and Runion 

1962, p.124) or by precipitation and filtration followed by calcination and blending (AEC 1959, p.7-9). 

A summary of both final process steps follows:

• Boildown and Crystallization

•

: The pure uranium solution could be boiled down and evaporated 

to produce uranyl salts, which were packaged for shipment. At times, the concentrated 

solution from the boildown step was packaged in bottles for shipment as a solution.

Precipitation/Filtration and Calcination/Blending: The extracted uranium liquid was 

precipitated by batch process with ammonium hydroxide in an approved container with an 

agitator. The precipitated slurry was transferred to the filtration hood for filtration with a large 

Buchner funnel. The filter cake was loaded into metal trays and transferred to a hooded muffle 

furnace where it was calcined to U3O8. Once cooled, the material was transferred to a 

blending station for introduction into a ball mill for grinding and blending. After ball milling, the 

U3O8 was sampled, packaged, weighed, and returned to the storage area ready for shipment. 
Document No. ORAUT-TKBS-0043 Revision No. 01 Effective Date: 07/16/2008 Page 15 of 46

2.1.4

This line of production (233UO2/ThO2 and 233UO2/ZrO2) was in operation from approximately 1961 to 

1969 for the light-water breeder reactor fuel of the Shippingport Reactor. This process was more 

involved than other process lines with 11 separate steps to produce the 233U fuel (Housholder 1963b, 

p.35-37). A brief summary of the process follows.

Production of Uranium Oxide Mixed with Thorium Oxide and Zirconium Oxide 

2.1.4.1 Receiving and Storage

Uranium-233 was received as a uranyl nitrate solution in an approved shipping cask. A typical receipt 

shipment was approximately 7 kg of 233U in storage columns awaiting processing.

2.1.4.2 Solution Concentration

The received uranyl nitrate solution had a concentration of approximately 150 g/L 233U. Before 

precipitation, a higher concentration was required and was obtained through evaporation.

2.1.4.3 Precipitation

The uranyl nitrate solution was measured into safe batches and the uranium was precipitated by the 

addition of a precipitating agent such as NH4OH.

2.1.4.4 Drying and Grinding

The resultant precipitate was dried at a low temperature and ground to a fine powder.

2.1.4.5 Calcination to Uranium Dioxide

The dried and ground uranium precipitate was calcined to UO2 in a continuous muffle furnace under a 

hydrogen atmosphere. The resultant oxide was stored in birdcages to await blending.

2.1.4.6 Blending

To ensure homogeneity of the oxide, the precipitation batches were accumulated and blended before 

addition of the diluents (ThO2 or ZrO2). After a homogenous UO2 blend was obtained, it was sorted in 

birdcages to await blending with the diluents.

2.1.4.7 Diluent Addition and Blending

Safe batches of UO2 and diluents were weighed out and blended together in a twin-shell blender. To 

ensure thorough mixing, each batch was also ball-milled.

2.1.4.8 Binder Addition

A binder was mixed with the oxide blend, and the wet mixture was granulated and dried. After drying, 

the granules were broken up by screening.

2.1.4.9 Lubricant Addition and Pressing

To improve pressability, a die lubricant was added to the granules and blended in. The oxides were 

then compacted into small pellets on a 40-ton press.
Document No. ORAUT-TKBS-0043 Revision No. 01 Effective Date: 07/16/2008 Page 16 of 46

2.1.4.10 Binder Removal and Sintering

The resultant pellets were loaded into trays and heated in an oven to drive off the binder. They were 

then sintered in a continuous muffle furnace under a hydrogen atmosphere.

2.1.4.11 Physical Measurement and Grinding

The sintered pellets were inspected and measured to see that they met customer requirements. 

Before packaging, the sintered pellets could have required grinding to the correct diameter. If grinding 

was necessary, the pellets were cleaned by washing them in water, drying them, and remeasuring 

them.

2.1.4.12 Tube Loading

The finished pellets were stacked, weighed, and loaded into Zircaloy tubes. The tubes were welded 

closed and loaded into shielded 55-gal-drum birdcages for temporary storage and eventual shipment 

to the customer. 

2.1.5

Documentation of the process steps for the production of MOX fuel (combination of PuO2 with LEU

UO2) for SEFOR at the facility was not available. However, other DOE sites have performed similar 

operations. Production of MOX fuel would have followed similar process steps for the production of 

233UO2/ThO2 and 233UO2/ZrO2 fuel from the blending step onto tube loading. The process steps were 

completely within shielded and filtered gloveboxes for worker health and safety. It is assumed that the 

plutonium sent to the facility was in approved shipping packages as PuO2 powder in sealed cans. 

The final product would have had approximately 5% 239Pu in fuel pin assemblies loaded into shielded 

55-gal-drum birdcages for temporary storage and shipment to the customer. 

Production of Mixed Oxide Fuel

2.1.6

UF6 cylinder washing was performed in the Building 200 complex to recover uranium in a ventilated 

glovebox using water or steam. The removed wash solution was transferred to the HEU scrap 

recovery process to recover the uranium.

Uranium Hexafluoride Cylinder Washing

3.0

The primary sources of internal radiation exposure at W.R. Grace were uranium and thorium dust

produced from the manipulation and chemical processing of those materials during recovery and fuel 

fabrication processes. Beginning in approximately 1966, there was also the potential for internal dose 

from plutonium-containing dust. 

ESTIMATION OF INTERNAL EXPOSURE

Uranium enrichment levels included DU, natural uranium (NU), LEU (3.5%), and HEU (93%) as well 

as 232U and 233U. Uranium-233 contains 232U as an impurity due to nuclear reactions from the neutron 

irradiation of 232Th. The 232U impurity results in elevated gamma and beta dose rates due to the 

ingrowth of 232U progeny (228Th and its progeny). Uranium from recycling operations would have 

included small activities of non-uranium isotopes such as 99Tc, 237Np, 232

Th, 106Ru, and 238Pu. There 

was one indication of RU processing in the case of a recycled 233U pellet (Housholder 1963b).

Table 3-1 lists the various enrichments and chemical forms of the processed radionuclides. 
Document No. ORAUT-TKBS-0043 Revision No. 01 Effective Date: 07/16/2008 Page 17 of 46

3.1

The chemical form and the uranium enrichment varied over time at W.R. Grace. The manufacture of 

uranium occurred in most of the buildings at W.R. Grace, with the exception of Buildings 110 and 234 

where plutonium was primarily processed. 

URANIUM EXPOSURES

In most cases, it will not be known how much of a specific uranium enrichment versus another, to 

which a worker could have been exposed. Due to the presence of many forms of uranium, the most 

favorable to claimant of the absorption types (F, M, and S) should be assumed (ORAUT 2007a). 

Table 3-1. Fuel types, chemical forms, isotopes, and enrichments of W.R. Grace process material. 

Radionuclide or fuel

Chemical form and 

solubility type(s) Isotope Enrichment

Uranium UF6, UO2F2, & UO2(NO3)2 (F)

UO3 & UF4 (M)

U3O8 & UO2, (S)

U-233

U-234

U-235

U-236

U-238

DU, NU, 

LEU (3.5%),

HEU (93%)

Plutonium PuO2 (S) Pu-238 0.8%, Pu-239 

2.6%, Pu-240 1.33%, 

Pu-241 93.4%, Am-241 

1.9%, Pu-242 0.0002% 

(% activity) 

Fuel grade

aged 10 yr (12% 

Pu-240)

Technetium or other 

transuranic elements

Same as the U or Pu matrix Tc-99, Np-237 NA 

MOX; Pu U-235 fuel PuO2 /UO2 (S) 20% PuO2 and 80% UO2

by weight

(Sharma et al. 2002, p.40)

About 3.5% U-235

MOX; Th U-233 fuela ThO2 /

233UO2, ZrO2 /233UO2

and ZrO2 /235UO2 (S)

1% UO2 and 99% ThO2 or 

ZrO2 by weight

(Housholder and Runion 

1962, p.189)

HEU (likely >20%)

a. See Table 3-3 for 233U fuel composition.

ICRP (1995) lists UF6, UO2F2, and UO2(NO3)2 (uranyl nitrate) as type F; UF4 and UO3 as type M; and 

U3O8 and UO2 as type S. Table 3-2 identifies uranium source term information. .

Table 3-2. Uranium source term information. 

Uranium 

source term Reference

Specific activity

(pCi/μg)

Activity fractions

U-234 U-235 U-236 U-238

NU IMBA 0.683 0.489 0.023 - 0.489

93.% IMBA 68.1 0.968 0.030 0.002 0.0003

3.5% IMBA 2.20 0.818 0.034 - 0.147

2% HPSa 1.616 0.648 0.041 0.0009 0.311

Typical DU IMBA 0.402 0.155 0.011 0.0005 0.834

Specific constituent activity in 

mixture (μCi/g, nCi/mg, or pCi/μg)

NU IMBA 0.683 0.334 0.016 - 0.334

93.% IMBA 68.1 65.9 2.04 0.136 0.020

3.5% IMBA 2.20 1.80 0.075 - 0.323

2% HPSa 1.616 0.778 0.049 0.001 0.373

Typical DU IMBA 0.402 0.062 0.004 0.0002 0.335

a. American National Standards Institute Standard N13.22 (HPS 1995). 

The composition of the Shippingport Reactor fuel, total core is shown in Table 3-3. (Olson, McCardell, 

and Illum 2002, p.37). For dose reconstruction, the composition of the 233U fuel mixture in Table 3-3 
Document No. ORAUT-TKBS-0043 Revision No. 01 Effective Date: 07/16/2008 Page 18 of 46

should be used for assessing uranium intakes beginning on January 1,1961 (Congress 1986, p. 107)

because it is more favorable to claimants. 

Table 3-3. Uranium-233 Shippingport Reactor fuel activity fractions, total core). 

G Bq/g Total activity % activity % weight

U-232 3.7795 7.93E+11 3.00E+12 1.56172% 0%

U-233 500600.60 3.57E+08 1.79E+14 93.18272% 2%

U-234 6547.90 2.31E+08 1.51E+12 0.78607% 0%

U-235 440.90 8.01E+04 3.53E+07 0.00002% 0%

U-236 112.80 2.40E+06 2.70E+08 0.00014% 0%

U-238 1900.60 4.47E+09 8.49E+12 4.41967% 0%

Th-232 23481000.00 4.06E+03 9.54E+10 0.04966% 98%

Total 23990606.58 1.92E+14 100

3.1.1

This discussion focuses on documented air-sampling data from two separate Health and Safety 

Laboratory (HASL) reports by the AEC. The first air sampling was performed at W.R. Grace during

the recovery of 93% HEU from uranium-aluminum alloy scrap in the storage, ceramics, and chemical 

buildings (AEC 1959, p.5-6). The second air sampling was performed during the recovery of 3.6% 

LEU from slag scrap (AEC 1961). Tables 3-4 to 3-9 list the results of the sampling analyses. 

Uranium Air Sampling

Table 3-4. Occupational exposures for 93% uranium-aluminum alloy recovery.

a 

Operator

Number of 

persons

Average daily weighted 

exposures (dpm/m3)

Column operator 6 31

Ceramic building operator 3 45

Charge makeup 3 14

Accountability 3 15

Chemist and technicians 8 3.1

Spectrographic operator and technician 3 2.5

Machine shop 5 12

Laundry 1 7

GMb 10.84

GSDb 2.76

95% confidence level valueb 57.63

90% confidence level valueb 39.84

50% confidence level valueb 10.84

a. Data from AEC (1959, p.11). 

b. Statistics calculated using a lognormal distribution and the Crystal Ball program. 

The air samples consisted of collection on filters of radioactive particulates from breathing zones and 

general areas during processing. The measured alpha activity on the filter was used to determine the 

airborne alpha activity concentrations. When multiple samples at a location were collected, AEC used 

the mean air concentration in subsequent calculations. AEC matched air concentration 

determinations with information about worker categories, locations, tasks, and times at each location 

or task. 

Although it is unlikely that workers would have been exposed to the same air concentrations from the 

many other processes that were conducted at NFS, this is the only currently available air sample 

information. In addition, changes in the processes and the safety controls could have resulted in both 

increased and decreased exposure for any given period. In this document, an estimate of the intake 

was made by calculating the 95th percentile value of the maximum average breathing-zone 

concentration for the 1961 AEC air sample results (from Table 3-8 later in this section) by assuming a 
Document No. ORAUT-TKBS-0043 Revision No. 01 Effective Date: 07/16/2008 Page 19 of 46

lognormal distribution of the data in the AEC report (AEC 1961). Using Crystal Ball, a lognormal 

distribution calculation yielded a distribution with a geometric mean (GM) of 167 dpm/m3

,a geometric 

Table 3-5. Average breathing-zone samples for 93% uranium-aluminum alloy recovery.

a

Operation

Number of 

samples

Average concentration 

(dpm/m3)

Shearing U-Al alloy for charge makeup and weighing 3 43

Charging digester with batch of U-Al alloy 2 170

Running material from digester through filter press and column 3 50

Cleaning residue from filter press 2 19

Securing O.K. Liquor from columns 1 65

Precipitation of O.K. Liquor 2 16

Filtering ppt. on Buchner funnel 2 15

Removing filter cake, placing in tray 1 1

Transferring tray from furnace to cooling area 2 65

Transferring tray from furnace to dry box, weighing and unloading 

tray in dry box, cleaning residue tray

3 280

Digestion of organic ashes in hood 1 1

Filtration of digested organic ashes 1 65

GMb 26.65

GSDb 5.84

95% confidence level valueb 485.13

90% confidence level valueb 255.57

50% confidence level valueb 26.65

a. Data from (AEC 1959, p.12). 

b. Statistics calculated using a lognormal distribution and the Crystal Ball program. 

Table 3-6. Average general air-sampling concentrations for 93% 

uranium-aluminum alloy recovery.

a

Area

Number of 

samples

Average concentration 

dpm/m3

Shearing and weighing 2 9

Solvent extraction area 6 19

Chemical building–oven area 3 23

Ceramics building 9 2

Machine shop 4 16

Wet chemistry lab 3 3

Spectrographic lab 3 2

Laundry 3 6

Lunch room 3 7

Clean locker room 3 2

Contaminated locker room 3 4

GMb 5.81

GSDb 2.52

95% confidence level valueb 26.55

90% confidence level valueb 18.98

50% confidence level valueb 5.81

a. Data from AEC (1959, p.13). 

b. Statistics calculated using a lognormal distribution and the Crystal Ball program. 

standard deviation (GSD) of 2.12, and a 95th-percentile value of 578 dpm/m3. The calculated 

95th-percentile air concentration was used to calculate upper estimates of internal exposures. The 

most conservative air concentrations were generated from the radiological task-oriented information in 

Table 3-8. If actual operator information from Tables 3-4 and 3-7 were to be used, the calculated 

intakes would be about a factor of 4 to 10 less. Task-oriented weighted average samples involve the 
Document No. ORAUT-TKBS-0043 Revision No. 01 Effective Date: 07/16/2008 Page 20 of 46

highest encountered air concentrations because of the closest proximity to the actual work and no 

accounting of the time away from the actual or lower exposure rate activities. The use of the taskoriented time-weighted average air concentrations is favorable to claimants.

Table 3-7. Occupational exposures for 3.6% uranium recovery from slag.

a

Operator

Number of 

persons

Average daily weighted 

exposures (dpm/m3)

Williams Roll Mill 3 170

Digestion 3 45

Solvent extraction 3 10

ADU precipitation 3 45

ADU oxide (ceramics) 3 17

Micropulverizer-drifter 3 71

Accountability-shipping and receiving 2 22

Laundry 1 9.4

Janitor 1 37

Health physics technician 2 20

GMb 30.72

GSDb 2.46

95% confidence level valueb 132.11

90% confidence level valueb 95.55

50% confidence level valueb 30.22

a. Data from AEC (1961, p.33). 

b. Statistics calculated using a lognormal distribution and the Crystal Ball program. 

Table 3-8. Average breathing-zone samples for 3.6% uranium recovery from slag.

a

Operation

Number of 

samples

Average concentration 

(dpm/m3)

Operating jaw crusher 4 230

Loading top of Williams roll mill with crushed slag 3 500

Removing drum of 325 mesh MgF2 from hopper 2 70

Loading and unloading pot (oxidation furnace) with slag metal 

heavies

2 140

Charging digester with slag 3 150

Cleaning ADU from plate and frame filter press 3 130

Placing tray of ADU inside of oven 2 84

Removing tray of oxide from furnace-placing inside of glove box and 

transferring U3O8

4 97

Micropulverizing U3O8 (bag not working properly) 3 590

GMb 167.83

GSDb 2.12

95% confidence level valueb 578.38

90% confidence level valueb 440.08

50% confidence level valueb 167.83

a. Data from AEC (1961, p. 34).

b. Statistics calculated using a lognormal distribution and the Crystal Ball program. 

As seen in Table 3-4, the uranium exposure by occupation differed with the column and ceramic 

building operators with the largest daily weighted exposures. This exposure variability between 

occupations is also seen in Table 3-7, where the Williams roll mill operators were clearly the highest 

exposed group, with the micropulverizer-drifter, digestion and ammonium diuranate (ADU) 

precipitation operators being the next highest groups. 
Document No. ORAUT-TKBS-0043 Revision No. 01 Effective Date: 07/16/2008 Page 21 of 46

The general air sample results in Tables 3-6 and 3-9 result in the lowest calculated intakes. This is to 

be expected because general air samples are typically lower than breathing-zone samples. Bioassay 

results are expected to be at levels between breathing-zone and general air sample results. 

The breathing rate is based on the default for light work, 1.2 m3

/hr, as indicated in International 

Commission on Radiological Protection (ICRP) Publication 66 (ICRP 1994). This category assumes 

Table 3-9. Average general air-sampling concentrations for 3.6% uranium 

recovery from slag.

a

Area

Number of 

samples

Average concentration 

dpm/m3

Crushing room 5 110

Digester area 6 11

Leach area 3 20

Solvent extraction area–all levels 8 6.4

Boil-down feed tank area 3 17

O.K. Liquor storage 3 9

Raffinate storage area 3 9

Neutralizer tank storage 4 10

ADU precipitation tanks 3 16

ADU filter area 4 17

U3O8 transfer hood area 3 11

Micropulverizer area 4 62

Accountability room 5 21

Chemistry lab 4 4.3

Spectrographic lab 2 9.3

Research and development lab 4 6.6

Laundry 4 2.3

Locker room–clean side 2 17

Locker room–dirty side 2 45

Lunch room 4 21

Health physics room 2 1.5

GMb 12.57

GSDb 2.7

95% confidence level valueb 64.52

90% confidence level valueb 44.95

50% confidence level valueb 12.57

a. Data from AEC (1961, p.35). 

b. Statistics calculated using a lognormal distribution and the Crystal Ball program.

an activity distribution of one-third sitting and two-thirds light exercise. The intakes in picocuries are 

calculated by dividing the 95th-percentile value of the air concentration (578 dpm/m3

) by 2.22 dpm/pCi 

and multiplying this result by the breathing rate and the assumed number of hours of exposure at a 

given concentration. The organ doses are assumed to be a constant distribution. Several 

assumptions in the intake and dose reconstruction are likely to be overestimating assumptions. This 

includes the use of a lognormal distribution, the 95% confidence level concentration, the task-related 

versus occupation time-weighted average, and the assumption of constant work activity and worker 

exposure during the entire work period. 

Air sampling for 233U has also been identified. Airborne concentrations for enriched UO2 decladding 

and dissolution of 233U3O8 pellets from the immediate work area resulted in an average concentration 

of less than 1.8% of the maximum air concentration (1 × 10-10 μCi/ml) (Householder 1963b, p.51). 

This would result in calculations of intake less than that calculated above.
Document No. ORAUT-TKBS-0043 Revision No. 01 Effective Date: 07/16/2008 Page 22 of 46

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

to be considered. NIOSH (2004) indicates that the ingestion rate, in terms of picocuries for an 8-hour 

workday, can be estimated by multiplying the air concentrations in picocuries per cubic meter by a 

factor of 0.2. The uranium ingestion rate based on air concentration of 260 pCi/m3 would be 

52 pCi/workday. The daily inhalation and ingestion intake rate from LEU recovery is estimated from 

the 95% confidence level air concentration in Table 3-8. Table 3-10 lists the inhalation intake rate per 

year. Table 3-11 lists the ingestion intake rate by year. 

Table 3-10. Estimated uranium inhalation intake rates based on measured time-weighted air 

concentrations during uranium recovery operations. 

Work 

period

Number 

of years

Number of potential AEC work 

hours per work period

Air concentration 

(pCi/m3)

Breathing 

rate (m3/hr)

Intake 

(pCi)

1 yr 1 2,000 260 1.2 6.24 × 105

Table 3-11. Estimated uranium ingestion rates based on measured time-weighted air concentrations 

during uranium recovery operations. 

Work 

period

Number of 

years

Number of potential AEC 

work days per work period

Ingestion rate 

pCi/workdaya Intake (pCi)

1 yr 1 250 52 1.30 × 104

a. Ingestion values were calculated according to NIOSH (2004). Choose the same f1-value as that used for inhalation in 

accordance with NIOSH (2004)

Bioassay data should be used first rather than the intakes that were calculated based on air sampling 

results unless the bioassay data are deficient. Most potentially exposed workers at W.R. Grace/NFS 

were monitored. Skin contamination was a recurrent issue at W.R. Grace/NFS. The contamination 

levels in the claimant records should be bounded by the assumption of 10% of the skin contamination 

equilibrium activity levels being ingested in accordance with Estimation of Ingestion Intakes (NIOSH 

2004). A study at the Oak Ridge Gaseous Diffusion Plant to determine the intake from hand 

contamination indicated that the amount of uranium that was transferred from the hand to the 

cigarette while smoking was approximately 1% of the material on the surface of the hand (Bailey 

1959, p.166). The ingestion calculation included both contamination of food or drink from 

contaminated air settling and the transfer from contaminated surfaces to the hands to food or drink on 

a chronic basis. 

.

3.1.2

For a given uranium process, the mass of (long-lived) uranium that is released to air does not change 

because of enrichment (ORAUT 2006a). Because the AEC air samples were counted with alpha 

detectors, which detect radioactivity rather than mass, there is no need to adjust measured air 

concentration results for assumed uranium enrichment. AEC air sample results taken during the 

recovery of HEU were below the measured concentrations during the recovery of LEU. Because 

various levels of enrichment occurred during the processing and recovery operations of uranium, this 

document assumes that intakes are 234U for the purpose of internal dose calculation.

Enriched and Recycled Uranium

It should be assumed that exposure to RU potentially occurred during the entire operational period 

(January 1,1958, to December 31, 1970) [1]. When assigning dose due to contaminants in RU, the 

same material type should be applied to the contaminants as that for the uranium (ORAUT 2007a, 

p.21). For RU, the dose reconstructor should add the intakes from Table 3-12 (ORAUT 2005a), or 

those derived from bioassay results. The RU contaminant ratios are bounding for recycled 

components in uranium 233 and as such should be applied regardless of whether 233U or 234U is 

assumed.(Tomlinson 1964, p.6) 

Table 3-12. Intakes of contaminants in RU as 

fractions of uranium intake.

a
Document No. ORAUT-TKBS-0043 Revision No. 01 Effective Date: 07/16/2008 Page 23 of 46

Radionuclide

Activity fraction of contaminant 

(e.g., pCi X/pCi U)

Uranium 1.0

Pu-238 0.06

Np-237 0.005

Tc-99 0.4

Th-232 0.02

Ru-106 0.04

a. Source: Based on ORAUT (2005a, p.11). 

3.1.3

Urinalysis for uranium started in approximately October 1964. Samples were sent to Eberline until 

about 1974. From about that time, onsite analysis was performed; details of the analysis are not 

known. The recording level was 1 dpm/L at the 2-sigma or 95% confidence level as indicated from 

the review of employee documents as indicated from the NFS Health Physics department (Tester 

2005a, p.2). This is likely the assumed decision level (DL). The minimum detectable level (MDL) 

should be twice the critical level or 2 dpm/L. Some random samples were sent off the site for fecal 

analysis of uranium and urine analysis for EU. Some bioassay records from W.R. Grace/NFS have 

some of the radiometric uranium bioassay results with a minimum recording level of <10 dpm/L. 

Other records have bioassay results reported on two forms: one handwritten form that indicated 

laboratory results with actual results recorded down to 1.0 dpm/L, and another typed report in which 

the lowest values are recorded as <10 dpm/L (i.e., all values <10 dpm/L were recorded as <10 dpm/L 

on the typed form). If the recorded result is between 2 and 10 dpm/L, 2 dpm/L should be used as the 

MDL. If zeros or <10 dpm/L were recorded for uranium results, then the dose reconstructor should 

assume the MDL is 10 dpm/L [2]. When bioassay data are available, dose reconstructors should use 

bioassay data to estimate worker intakes. Because the uranium urine sample results are reported in 

units of activity (dpm/L), it is not necessary to adjust these results to a particular uranium enrichment. 

Intakes of the RU components in Table 3-12 should be added based on the given ratios.

Uranium Bioassay

Lung counts have been performed from 1970 to the present for uranium, as indicated from a cursory 

review of claimant documents. The minimum detectable activity (MDA) for 235

U was about 120 μg

(Tester 2005a, p.2). 

3.2 PLUTONIUM EXPOSURES

Many forms of plutonium were possible over the years including metal and oxide. Because not 

enough information for the recovery or manufacture of plutonium was found, the exact chemical forms 

are not known. It is possible that MOX fuel of a plutonium and thorium mixture was processed at 

some point because UO2 and ThO2 mixtures were produced on the site and plutonium production 

capacity existed at the same time. Most of the manufacture of plutonium occurred in Buildings 234A, 

234B, and 234C. 

In general, plutonium oxides, carbides, and hydroxides are absorption type S; nitrates and other 

compounds are type M (ICRP 1995). Older materials, even when starting out as soluble, can have a 

tendency to oxidize when left in contact with air. Oxides, metals, and old contamination should be 

treated as type S. If the chemical form of plutonium is unknown, either type M or S can be used to 

maximize the dose to the organ of concern (ORAUT 2007a). Americium-241 is a component of 

plutonium contamination and should be modeled in the lung the same as the plutonium matrix in 

which it has grown. In other words, the americium should be treated as absorption type S if the 

plutonium is type S (ORAUT 2007a). It is possible that some forms of plutonium at W.R. Grace could 

be retained in the lung longer than is predicted by the normal material type S model (commonly 

referred to as type Super S). Organ doses based on intake of plutonium should be evaluated in 
Document No. ORAUT-TKBS-0043 Revision No. 01 Effective Date: 07/16/2008 Page 24 of 46

accordance with the guidance in the technical information bulletin Estimating Doses for Plutonium 

Strongly Retained in the Lung (ORAUT2007c).

There are essentially three types of plutonium-based material – reactor, fuel, and weapons grade – 

with fuel grade falling between reactor and weapons grade. Although the origin of the plutonium at 

W.R. Grace is currently unknown, without any specific information on the actual composition of the 

processed plutonium an assumption of 10-year-old fuel-grade plutonium is favorable to claimants and 

reasonable [3]. This is noted in the Hanford site profile (ORAUT 2007b) because Hanford processed 

much of the DOE complex plutonium. A summary of the activity composition of the default reference 

fuel-grade plutonium mixture is presented in Table 3-13. 

Table 3-13. Activity composition of Hanford 

Site reference fuel-grade plutonium mixture 

(12%).a

Radionuclides: 

10 year agedb

Specific activity 

in mixture (Ci/g)a

Pu-238 1.58E-02

Pu-239 5.26E-02

Pu-240 2.72E-02

Pu-241 1.91E+00

Pu-242 3.93E-06

Am-241 3.89E-02

Pu-239+240 7.98E-02

Activity ratios

Pu239+240:Am-2412.05

Pu239+240:Pu-2385.05

Pu-241:Pu239+24024.0

a. Source: ORAUT (2007b, p.19)

b. Time since separation of the Am-241 from the 

plutonium mix. 

3.2.1

A review of plutonium records during the operational period showed that 239Pu was analyzed in urine 

from 1967 to approximately 1973 [4]. The years of plutonium exposures most likely occurred from 

1967 though February 1973; 

Plutonium Bioassay 

however, a plutonium exposure period of 1966 through 1970 should be 

assumed (for the operational period), because it is possible that some startup plutonium operations 

could have occurred in 1966 (Claimant records; Congress 1986, p.107; and NFS 2005, p.2). 

Historical detection limits for W.R. Grace are not available, so a review of detection limits for the 

1960s and 1970s at other AEC sites was made. The internal dosimetry section of the DOE Hanford 

site profile (ORAUT 2007b) reports an MDA of 0.05 dpm/sample. The Savannah River Site profile 

(ORAUT 2005b) has a plutonium recording level and MDA of 0.1 dpm/1.5 L. These values are 

reasonably consistent with the observed reporting limits from W.R. Grace. From a review of bioassay 

records, the lowest observed nonzero recorded plutonium results at W.R. Grace was 0.03 dpm/L, 

although some results were reported in units of dpm/sample, with a given sample volume. If 

additional information on detection capabilities is not available from the records, dose reconstructors 

should assume a DL of 0.03 dpm/L and an MDA of 0.06 dpm/L [5]. 

Bioassay data should be used for assessing a worker’s plutonium dose. No attempt should be made 

to estimate plutonium dose for unmonitored workers during the operational period. 
Document No. ORAUT-TKBS-0043 Revision No. 01 Effective Date: 07/16/2008 Page 25 of 46

Chest counting for 239Pu started at W.R. Grace around 1987. A germanium detector system had the

ability to detect 239Pu but had a variable MDA of about 168 nCi (July 6, 1989) to 481 nCi (October 22, 

1987). The chest-counting detection limits are based on a cursory review of claimant files and are not 

useful because the MDAs are high. Most results were nondetections or perhaps near the MDA. 

3.3 URANIUM-233 CONSIDERATIONS

Thorium oxide as ThO2 and uranium oxide as 233UO2 were the finished products for the production of 

233U/232Th/228Th MOX fuel. The 233U was received from Oak Ridge in the form of uranyl nitrate, then 

precipitated to (NH4)2-U2O7 (ADU), then calcined to UO2. Blending, pressing, and grinding were also 

a part of the process in addition to chemical processing. Particle sizes varied. Because of criticality 

concerns, a safe-dry batch did not exceed 4 kg of 233U during blending operations; this amount of 233U 

is nuclearly safe with a safety factor of 2.3 for a H/233U ratio of 1.5 and in a nominally reflected system. 

It should be noted that the limitation of 4 kg of 233U at low H/233U ratios is for a metal-water system and 

the campaign tested was for a low-density oxide-water system. With 235U, where more data are 

available, almost three times as much low-density 235U than high density 235U is nuclearly safe. No 

enrichment in 233U was stated in the Feasibility Report, but the ZrO2-

233UO2 pellet Feasibility Report 

mentioned 93% enrichment (Housholder 1963b). Isotopic composition of 232Th/233U fuel is given in 

Table 3-3.

3.4 INTERNAL DOSE RECONSTRUCTION ASSUMPTIONS 

Because of the SEC determination (NIOSH 2007b) that it is infeasible to adequately reconstruct 

internal thorium dose during the AWE operational period that is covered by the SEC (January 1, 1958 

through December 31, 1970), dose estimates for this period are considered partial dose estimates. 

Uranium and plutonium (and associated radionuclides) intakes should be based on actual bioassay 

monitoring results for the individual. When uranium bioassay results are not available, the values in 

Table 3-14 should be assigned for estimation of uranium dose. If a worker has positive thorium 

bioassay results during the operational period, these positive results can be used to reconstruct a 

partial thorium dose, as an approach that is favorable to claimants; however, this would still be 

considered a partial dose assessment because such dose is known to not be the entire possible

thorium dose. In addition, if an individual was not monitored for plutonium exposures through 

bioassay, no attempt should be made to estimate plutonium dose during the operational period, and 

the dose reconstruction is to be considered to be a partial internal assessment.

Table 3-14. Uranium internal exposure summary for operational period January 

1, 1958, to December 31, 1970.

a

Internal Start End Exposure Intake (pCi/d) IREP distribution

Uranium 1/1/1958 12/31/1970 Inhalation

Ingestion

1.71E+03

3.56E+01

Constant

Pu-238 1/1/1958 12/31/1970 Inhalation

Ingestion

1.03E+02

2.14E+00

Constant

Np-237 1/1/1958 12/31/1970 Inhalation

Ingestion

8.55E+00

1.78E-01

Constant

Tc-99 1/1/1958 12/31/1970 Inhalation

Ingestion

6.84E+02

1.43E+01

Constant

Th-232 1/1/1958 12/31/1970 Inhalation

Ingestion

3.42E+01

7.13E-01

Constant

Th-228b 1/1/1958 12/31/1970 Inhalation

Ingestion

3.42E+01

7.13E-01

Constant

Ru-106 1/1/1958 12/31/1970 Inhalation

Ingestion

6.84E+01

1.43E+00

Constant
Document No. ORAUT-TKBS-0043 Revision No. 01 Effective Date: 07/16/2008 Page 26 of 46

a. Ingestion values were calculated according to NIOSH (2004). Choose the same f1-value as 

that for inhalation in accordance with NIOSH (2004). 

b. Th-232 is assumed to be in equilibrium with Th-228.

The assumed occupational exposure period ran from January 1, 1958, to December 31, 1970. 

Table 3-14 lists the intake assumptions for uranium and associated compounds. These intake rates 

are based on information from Tables 3-10 and 3-11, which were derived from the 95th-percentile 

value of the average breathing-zone samples for 3.6% uranium recovery from slag (578.38 dpm/m3

), 

from Table 3-8, a breathing rate of 1.2 m3

/hr (ICRP 1994), and an annual occupational exposure of 

2,000 hours. The intake mode is chronic. 

4.0

During operations at the facility, occupational exposure occurred from handling of received material, 

standing near stored fissile material (either as feed or product), and airborne radioactivity with the 

resultant buildup of surface contamination. The fissile material was enriched UF6 that arrived in 

approved shipping and storage cylinders, solid fissile material in various forms (liquid, powder, or 

metal) to be converted into or made into nuclear fuel, or scrap material containing LEU or HEU.

ESTIMATION OF EXTERNAL EXPOSURE

4.1 LIMITATIONS

The potential for external radiation dose existed at all locations where radioactive materials were 

handled or stored. Based on site operations, sources of potential exposure included beta, photon,

and neutron radiation emitted from materials containing uranium, thorium, and plutonium. Personnel 

beta/photon dosimeters were assigned to workers. Claimant-specific data should be used for analysis 

of the external radiation dose. No attempt should be made to estimate external radiation doses for 

unmonitored workers during the operational period and, if this is done, the dose reconstruction should 

be considered to be a partial dose estimate. The only exception to assessing the unmonitored worker 

dose concerns the estimation of neutron doses to chemical operators using a neutron-to-photon dose 

ratio as discussed further in this section.

4.2 RECORDED EXTERNAL DOSE

Beginning in the late 1950s, Chicago Nuclear supplied all dosimetry badges and performed the 

necessary calibrations. At one time, everyone in the W.R. Grace plant was badged. At a later time, 

office workers were not assigned dosimeters (ORAUT 2005c). Landauer supplied dosimeter services 

beginning in 1961. The reported doses were normally adjusted for results of transit control 

dosimeters (i.e. control dosimeters were sent with dosimeters to be assigned to workers, and the 

control dosimeter results, were deducted from the personnel dosimeter results to arrive at the 

reported occupational dose). Minimal beta or soft X-ray nonpenetrating doses were not routinely 

reported until a positive skin dose was recorded (Gordon 2004, p.3). In the approximate 1989

timeframe, Landauer supplied W.R. Grace/NFS with TLDs instead of film. 

External radiation exposure records at W.R. Grace/NFS utilized AEC form AEC 5, which is similar to 

the NRC form 5 of today. Gamma, beta, and neutron doses are listed in separate columns. In 1961, 

Landauer started to report the external dose in the same format as AEC form 5. In 1964, Landauer 

still reported the gamma, beta, and neutron dose components in separate columns but had the same 

forms filled out separately for whole body, skin, and extremity dose components. Code 1 was for the 

whole body, code 2 for the skin of the whole body, and code 3 for the extremities (i.e. hands and 

forearms). Beginning January 1, 1970, the reported skin dose included the shallow dose from 

nonpenetrating and penetrating dose components. Prior to 1970, the reported skin dose represented 

only the nonpenetrating dose component. As such, prior to 1970, the total shallow dose is best 

estimated by summing the reported skin (i.e. nonpenetrating dose) and whole body dose (i.e. 

penetrating dose component) doses. 
Document No. ORAUT-TKBS-0043 Revision No. 01 Effective Date: 07/16/2008 Page 27 of 46

Table 4-1 lists a more detailed description of the Nuclear Chicago or Landauer dosimeters used at 

W.R. Grace and subsequently, NFS. W.R. Grace/NFS’ distribution of low-energy photons is 

dependent on the amounts, separation and enrichments of uranium, and the age and type of 

plutonium used at the site. 

Table 4-1. Dosimetry for external whole-body, wrist, and extremity exposures. 

Period Monitoring technique Dosimeter description

Beta/photon dosimeters

1958–June 1974

whole body

Photographic film 

badge 

Nuclear-Chicago or similar film badges. Nuclear-Chicago 

film badge contained single film packet. Three filters (front 

and back) were incorporated into film badge for energy 

dependence: cadmium, aluminum, and lead. 

July 1974–about 1983a

whole body

Landauer J (beta and 

gamma)

Type J dosimetry were film badges. Gamma and X-ray: 

30 keV to 20 MeV; beta: over 1.5 MeV.

About 1974a–April 1990

whole body

Landauer G1 Film emulsion packaged placed in standard Gardray 

holder/badge for monitoring beta, X-ray, and gamma 

exposure. Insensitive to neutron radiation. 

Gamma and X-ray: 30 keV to 20 MeV; beta: over 1.5 MeV.

May 1990–present

whole body

Landauer Z1 dosimeter 3 TLD-700 chips for monitoring beta, X-ray, and gamma 

exposure. Insensitive to neutron radiation. Replaced 

Landauer G1.

July 1974–about 1983a

wrist

Landauer Type M 

(wrist 

beta-gamma) badges.

Type M dosimetry was a film badge. Gamma and X-ray: 30 

keV to 20 MeV; beta: over 1.5 MeV.

About 1983a–1990

wrist

G5 wrist film badge Responded to beta, X-ray, and gamma exposure to provide 

data on extremity dose. 

Gamma and X-ray: 30 keV to 20 MeV; beta: over 1.5 MeV.

1991–present

wrist

K5 TLD wrist badge 3 TLD-100 chips.

1958–1974 Film badge-finger Nuclear-Chicago or similar film badges. Nuclear-Chicago 

film badge contained single film packet. Three filters (front 

and back) were incorporated into film badge for energy 

dependence: cadmium, aluminum, and lead. 

About 1983a–present

finger ring

U3 TLD (LiF) Responded to beta, X-ray, and gamma exposure to provide 

data on extremity dose.

Gamma and X-ray: 30 keV to 20 MeV; beta: over 1.5 MeV.

a. This analysis found no documentation that shows the start of G1, U3, G5, I8, and E1 dosimeter use and the end of 

Type K, J, and M badge use.

The NRC has reported the annual occupational radiation dose for nuclear power reactors since 1974. 

Consistent annual dose reporting could not be found in NRC records for the W.R. Grace Plant until 

1982. The only information found prior to 1982 was for whole-body doses greater than 1.25 rem for 

employees with employment of less than 90 days. The reporting format varied over the years. In the 

early years, the format included: the number of monitored individuals, measurable dose, collective 

dose, and average measurable dose in rem. In later years, the format included the annual wholebody doses segregated into dose bins with the number of workers in each bin. 

4.2.1

Dosimeters used at W.R. Grace would measure beta and low-energy photon nonpenetrating dose. 

Prior to 1970, the reported skin dose represents only the nonpenetrating dose component and 

beginning January 1, 1970, the reported skin dose includes the shallow dose from the penetrating 

dose component. For this reason, determination of the skin dose beginning on January 1, 1970 must 

be calculated as the difference between the reported skin and total-body dose [6]. For low energy 

photon irradiation such as to 16-keV X-rays and 59-keV photons typical of plutonium facilities, there is 

Nonpenetrating Dose
Document No. ORAUT-TKBS-0043 Revision No. 01 Effective Date: 07/16/2008 Page 28 of 46

an over-response by factors of 8.5 to 12 and 14 to 19, respectively based on measurements at 

Hanford (Wilson et al, 1990) when the dosimeter shallow dose is calibrated to uranium. 

4.2.2

Commercially provided personnel dosimeters used at W.R. Grace would measure the photon 

penetrating dose. Exposure to photon radiation from natural uranium metals and solutions would be 

comparatively low compared to the nonpenetrating exposure. The photon radiation is predominantly

in the 30-to-250 keV energy range. As uranium becomes more enriched, the photon spectrum 

increases because of the higher specific activity of 235U, and the energy of the photon emissions (DOE 

2004). Photon radiation from plutonium would be expected to include <30 and 30-250 keV 

components. 

Penetrating Dose

4.3 NEUTRON DOSE

There were no documented neutron exposures at W.R. Grace. However, the use of uranium 

hexafluoride and uranium fluoride, and the possible presence of plutonium can generate neutrons 

through an alpha-neutron reaction between the uranium and the fluorine (DOE 2004). Neutrons can 

also arise from highly enriched uranium and during plutonium processing operations (DOE 2006). 

Personnel exposure records are used to evaluate radiation exposure. No attempt should be made to 

estimate neutron dose for workers not monitored for neutrons during the operational period. If a dose 

reconstruction is performed, the dose reconstruction is considered to be a partial dose reconstruction 

and should be noted as such. 

4.3.1

Extremity dosimeters for monitoring exposures to the forearm and hand were used at W.R. 

Grace/NFS from at least the 1960s to the 1970s. The MDL information in Table 4-2 for wrist (forearm) 

and finger (hand or extremity) is from Landauer. Ring and wrist badges were calibrated for highenergy gamma (for 137Cs at 0.662 MeV) and high-energy beta (1.5 MeV) unless special arrangements 

were made (Gordon 2004, p.3). DOE (2006) provided results for measured extremity photon dose 

rates from plutonium glovebox operations and discusses that: “Doses to the extremities are usually 

dominated by gamma rays in typical glovebox operations. The extremity dose is more limiting than a 

whole body dose if the dose gradient is greater than 10:1 over a distance of 1 meter, the maximum 

distance from the fingers to the trunk of the body. In most cases, the source is not at arm’s length and 

the dose gradient needs to be 10:1 or 20:1 for the extremity dose to be limiting (NUREG/CR-4297; 

Reece et al. 1985). But in highly shielded gloveboxes, it is possible to have a very high extremity 

dose from dust layer on gloves; the dose to the torso can be much lower because of shielding applied 

to the glovebox.” Dose reconstructors should use guidance from DOE (2006) if the calculation of

extremity dose is needed.

Extremity Doses

4.4 MISSED DOSE

(MDL) information for photon and beta whole-body dosimeter systems are listed in Table 4-2. The 

potential annual missed doses listed in Table 4-2 are considered default values and are to be used, 

unless claimant dosimetry records suggest that a different dosimeter exchange frequency occurred 

[7]. 
Document No. ORAUT-TKBS-0043 Revision No. 01 Effective Date: 07/16/2008 Page 29 of 46

4.5 DOSE RECONSTRUCTION RECOMMENDATIONS

4.5.1

Measured and missed non-penetrating doses from beta radiation should be corrected to account for 

attenuation by clothing or personal protective equipment (PPE), if applicable, based on the location of 

the cancer and the workplace practices. Nonpenetrating dose can be considered for IREP input 

Beta Dose

Table 4-2 MDL and maximum potential annual missed photon or beta dose.a

Dosimeter Period of use MDL (rem)

Maximum annual missed dose (rem)

(MDL/2 × frequency)

Nuclear Chicago film–

whole body

1957–5/1959b 0.04 photons

0.04 beta

0.520 beta-photons (every 2 weeks)

Nuclear Chicago film–

whole body

6/1959–12/1960b 0.04 photons

0.04 beta

0.240 beta-photons (monthly)

Landauer film–

whole body

1/1961–12/1963b 0.04 photons

0.04 beta

0.240 beta-photons (monthly)

Landauer film–

whole body

1/1964–12/1988c 0.04 photons

0.04 beta

0.240 beta-photons (monthly)

Landauer TLD–

whole body

1/1989–12/1998c 0.04 photons

0.04 beta

0.080 beta-photons (quarterly)

Landauer OSL–

whole body

1/1/1999–12/2004c 0.001 photons

0.001 beta

0.008 beta-photons (quarterly)

Landauer TLD–

whole body

1/2005–presentc 0.02 photons

0.02 beta

0.040 beta-photons (quarterly)

Film-badge–

wrist

1957–June 1974b 0.04 photons

0.04 beta

0.240 beta-photons (monthly)

Landauer Type M–

wrist

July 1974–about 1983c 0.02 photons 

0.04 beta

0.120 photons (monthly)

0.240 beta (monthly)

G5 wrist film badge–

wrist

About 1983–1990c 0.02 photons

0.04 beta

0.120 photons (monthly)

0.040 photons (quarterly)

0.240 beta (monthly)

0.080 beta (quarterly)

K5 TLD–

wrist 

1991–presentc 0.04 photons

0.04 beta

0.240 photons (monthly)

0.240 beta (monthly)

Film badge–

finger

1957–1982c 0.04 photons

0.04 beta

0.240 beta -photons (monthly)

U3 TLD (LiF)–

finger ring

About 1983–presentc 0.03 photons

0.04 beta

0.180 photons (monthly)

0.060 photons (quarterly)

0.240 beta (monthly)

0.080 beta (quarterly)

a. TLD = thermoluminescent dosimeter.

b. MDLs from Tester (2005a, p.2-3). 

c. Koperski (2004). MDL information as communicated by Landauer.

categories as either electrons greater than 15 keV or photons <30 keV, depending upon which 

selection is more reasonable based on claim information and favorability to the claimant. Typically, 

nonpenetrating doses for plutonium workers are assigned as <30 keV photons for deep organ 

cancers (e.g. nonskin cancers) using the dose conversion factors (DCFs) for plutonium workers in 

accordance with the External Dose Reconstruction Implementation Guide (NIOSH 2007a). If a 

nonpenetrating dose is assigned as <30 keV photons, attenuation factors for clothing are not 

applicable (ORAUT 2005d). A Dose Reconstructor should apply the guidance in ORAUT-OTIB-0017 

for assignment of skin dose from penetrating and nonpenetrating radiation (ORAUT 2005d).
Document No. ORAUT-TKBS-0043 Revision No. 01 Effective Date: 07/16/2008 Page 30 of 46

4.5.2

Measured penetrating doses from photon radiation should be reasonably correct. It is possible to read 

a photon dose of 100 mrem to within ±15 mrem if the exposure involved photons with energies 

between several keV and several MeV (Morgan 1961, p.13). The estimated standard error in 

recorded film badge doses from photons of any energy is ±30% (ORAUT 2006c). Dose

Reconstructors should include an uncertainty factor of 1.3 multiplied by the measured photon dose, 

assume a constant distribution for overestimates, and utilize the Exposure ( R ) to Organ Dose (HT), 

[8] dose conversion factors from IG-001. To ensure favorability to claimants, a photon energy range 

of 100% 30 to 250 keV should be applied (ORAUT 2006b).

Photon Dose

4.5.3

Consideration of glovebox geometry is necessary for workers with significant glovebox or similar 

benchtop geometry work activities with plutonium, as noted in OCAS-TIB-0010, “Special External 

Dose Reconstruction Considerations for Glovebox Workers.” (NIOSH 2005). This would include the 

Chemical Operators. Indicators of hands-on plutonium work may be plutonium bioassay results

and/or extremity monitoring. For these work activities, adjustment to assigned doses for organs of the 

lower torso in relation to the measured whole-body dosimeter dose is necessary to account for 

potential geometry effects. 

Glovebox Geometry Assumptions

4.5.4

In the spring of 1961, NFS processed 28 kg of 233U in used pellets. The radiation levels reached 

3,000 mR/hr at 1 in. for 15 g of U3O8 pellets. Twenty operators were exposed to these operations with 

none exceeding the 10 CFR Part 20 limits; the average exposure for the operation was 377 mrem. 

Rubber gloves were worn during operations in shielded gloveboxes with filtration. Birdcages were 

used for storage of the materials. The exposure rate for the surface of each container did not exceed 

200 mR/hr, and the contamination level did not exceed 500 dpm/100 cm2

. The reading from each 

birdcage did not exceed 1 mR/hr at 1 meter (Householder 1963b, p.51-55). Doses from the 

processing of 233U were measured through external monitoring. Therefore, dose reconstructors 

should use reported dosimetry results to assign measured dose, along with the assignment of missed 

dose. 

Uranium-233

4.6 OCCUPATIONAL MEDICAL DOSE 

To date, no site-specific information is available for W.R. Grace workers in relation to the type or 

frequency of occupationally required medical X-ray examinations. The type and projection of X-ray 

examination should be based on current Oak Ridge Associated Universities (ORAU) Team guidance, 

which is a posterior-anterior (PA) radiographic chest examination [9]. Preemployment and annual PA 

chest x-rays are assumed for all site workers. Organ doses can be obtained from the current revision 

of ORAUT-OTIB-0006, Dose Reconstruction from Occupationally Related Diagnostic X-Ray 

Procedures (ORAUT 2005e). Photofluorography (PFG) could have been possible, but unless there is 

evidence in the claimant files that PFG was performed; it is reasonable to assume it was not [10]. 

Occupational medical X-ray doses should be entered into IREP as the annual dose to an acute 

exposure to photons with an energy range of 30 to 250 keV. The distribution is assumed to be normal 

with a standard deviation of 30%. 
Document No. ORAUT-TKBS-0043 Revision No. 01 Effective Date: 07/16/2008 Page 31 of 46

4.7 SUMMARY OF EXTERNAL DOSE RECONSTRUCTION RECOMMENDATIONS 

Guidance presented in the foregoing sections to determine IREP input categories for annual external 

doses to be assigned is summarized in Table 4-3. 

Table 4-3. IREP input summary. 

Dose 

category

Period of 

exposure Exposure category

Exposure 

type Basis

Annual 

exposure

IREP

distribution

Photon missed

dose

1/1/1958–

12/31/1970

Photons, 30-250 keV, AP 

acute

Penetrating Table 4-2 Use MDL/2 × 

exchange 

frequency) 

Lognormal 

GSD 1.52

Photon missed 

dose

1/1/1958–

12/31/1970

Photons, <30 keV or 

electrons >15 keV, AP, 

acute

NonpenetratingTable 4-2 Use MDL/2 × 

exchange 

frequency & 

OTIB-0017

Lognormal 

GSD 1.52

Photon 

measured 

dose

1/1/1958–

12/31/1970

Photons, 30-250 keV, AP 

acute

Penetrating Recorded 

value

1.3 * 

recorded 

dose

Constant

Photon 

measured 

dose

1/1/1958–

12/31/1970

Photons, <30 keV or 

electrons >15 keV, AP, 

acute

NonpenetratingRecorded 

value

1.3 * 

recorded 

dose

Constant

Medical X-ray 1/1/1958–

12/31/1970

PA radiographic chest 

exam

Penetrating Initial plus 

annual 

examinations

See Section 

4.4.6

See ORAUT

(2005e) 

5.0

The W.R. Grace/NFS plant continued to operate after the covered AWE period, which ended on 

December 31, 1970. The Report on Residual Radioactive and Beryllium Contamination at Atomic 

Weapons Employer Facilities and Beryllium Vendor Facilities, from December 2006 noted that a 

significant potential for residual contamination remained after 1970. The residual contamination 

period at W.R. Grace is from January 1, 1971, to the present (NIOSH 2006). Differentiation between 

residual contamination during the operational period at W.R. Grace and contamination due to ongoing 

operations is not feasible.

ESTIMATION OF EXPOSURE TO RESIDUAL ACTIVITY

5.1 INTERNAL EXPOSURE TO RESIDUAL ACTIVITY

5.1.1

During the residual radioactivity period at W.R. Grace (1971 to the present), personnel might have 

been internally exposed to uranium, plutonium, and thorium in the workplace. This section describes 

the methdology for the assignment of internal dose. 

Introduction

Internal exposures should be estimated using bioassay data, if they exist. For situations where no 

internal monitoring was performed or monitoring information is not available, the default intakes for 

uranium, plutonium, and thorium should be assigned, unless the potential for exposure to these 

radionuclides can be ruled out. In situations where there are bioassay data, but gaps exist, the 

default intakes should be assigned from the day following the date of the last bioassay sample, 

through the end of the potential exposure period (ORAUT 2007a, p. 14).

The following information provides a method for estimating exposures during the residual radiation 

period due to uranium, thorium, and plutonium contamination, in cases where bioassay data is 

unavailable.
Document No. ORAUT-TKBS-0043 Revision No. 01 Effective Date: 07/16/2008 Page 32 of 46

5.1.2

If bioassay data are available for the residual period, then that data should be used to estimate and 

assign internal dose for the worker. Internal doses from bioassay data for this period should be 

evaluated with the guidance in Section 3.0. Intakes that are calculated in this manner are considered 

an overestimate because only the contribution from residual contamination should be applied. If a 

best estimate must be calculated, source term depletion factors from Table 3-1 of ORAUT (2008) may 

be applied as appropriate.

Uranium

When bioassay data are not available, and the worker might have worked in areas with residual 

activity, the following method may be used to assign internal dose. 

The uranium intakes (and associated RU components) in Section 3.0 (Table 3-14) may be used as a 

basis for intakes during the residual period. These tabulated values are used to calculate an 

associated average daily air concentration (assuming 250 workdays per calendar year and a 

breathing rate of 9.6 m3

/d).

An average daily air concentration is used to estimate the residual surface concentration with

guidance from ORAUT (2008). The annual deposition amount is estimated with a deposition velocity 

of 0.00075 m/s and with deposition assumed to occur for 1 year. Using this approach a surface 

concentration of uranium is estimated as follows: 

 air concentration pCi/m3

 × 31,536,000 s/yr × 0.00075 m/s = surface activity dpm/m2 (Eq. 1)

This mean surface concentration is favorable to claimants at the end of the operating period. The 

deposited material is assumed to have been resuspended and inhaled during the residual period. 

The amount of resuspension is assumed to reduce with time due to fixing of the material on surfaces 

and to also due to depletion (ORAUT 2008). The depletion factors for each year are described in 

Table 3-1 of ORAUT (2008). The depletion factors indicate that the residual concentration at the end 

of the operational period is to be used for the first year, the second year is reduced by a factor of 0.03, 

and the third and remaining years are reduced by a factor of 0.0007 (which represents a constant 

concentration after the third year).

The intakes in Table 5-1 can be used to estimate the internal dose to the target organ for the years of 

employment for the worker. The estimated internal doses are assigned as a constant distribution. 

Table 5-1. Uranium (and RU components) air concentration and annual 

intake in the residual period. 

Nuclide

Intake (pCi/yr)

1971 1972 1973–present

U-234 1.5E+04 Inhalation

3.12E+02 Ingestion

4.4E+02 Inhalation

9.2E+00 Ingestion

1.0E+01 Inhalation

2.1E-01 Ingestion

Pu-238 8.9E+02 Inhalation

1.9E+01 Ingestion

2.7E+01 Inhalation

5.6E-01 Ingestion

6.2E-01 Inhalation

1.3E-02 Ingestion

Np-237 7.4E+01 Inhalation

1.5E+00 Ingestion

2.2E+00 Inhalation

4.6E-02 Ingestion (a)

Th-232 3.0E+02 Inhalation

6.2E+00 Ingestion

8.9E+00 Inhalation

1.9E-01 Ingestion (a)

Th-228 3.0E+02 Inhalation

6.2E+00 Ingestion

8.9E+00 Inhalation

1.9E-01 Ingestion (a)

Tc-99 (a) (a) (a)

Ru-106 (a) (a) (a)

a. No intake assigned, <0.001 rem/yr.
Document No. ORAUT-TKBS-0043 Revision No. 01 Effective Date: 07/16/2008 Page 33 of 46

The internal dose analysis should include the potential inadvertent ingestion of uranium activity and 

be based on guidance in NIOSH (2004). The daily intake rate in activity per day is estimated as 0.2 

times the average daily air concentration expressed in units of activity/m3

. Using the air concentration 

for the first year after the end of operations (1971) from Table 5-1, this value should be applied for all 

years of the residual period using the depletion factors as described in Table 3-1 of ORAUT (2008). 

For ingestion intakes, it is not necessary to include intakes of 228Th and 237Np from 1973 through the 

present, because these intakes result in < 0.001 rem per year. It is also not necessary to include 

inhalation and ingestion intakes of 99Tc and 106Ru from 1971 to the present, because these intakes 

result in < 0.001 rem per year.

5.1.3

Plutonium urine sample results from the operational period were reviewed, sorted from lowest to 

highest, ranked and statistically analyzed (ORAUT 2005f) [11]. A summary of the urine sample 

results is presented in Table 5-2. The highest reported level was 2.856 dpm/d (after normalizing the 

result) from October 3, 1967. Seventy-five sample results were taken during the operational period 

(October 1, 1958, to December 31, 1970). Thirty-six (approximately 49%) of these sample results 

were less than the assumed DL of 0.03 dpm/L. A review of the available plutonium data (to date) 

suggests that they are representative of the types of workers most likely to be exposed to plutonium 

and includes: chemical operators, operators, production and maintenance foreman, front line 

supervisors, and radiation technicians. Table 5-3 summarizes the 95th-percentile excretion rate, 

which was derived from the plutonium urinalysis results during the operational period. 

Plutonium

Table 5-2. Plutonium urinalysis results from 

operational period (January 1, 1958 to 

December 31, 1970) as of April 17, 2007. 

Sample date(s)

Result 

(dpm/d)a,b Sample #’s

10/11/67–12/05/70 <LODc 1–36

12/4/1967 0.042 37

12/5/1967 0.042 38

11/3/1967 0.056 39

9/3/1968 0.056 40

5/29/1968 0.070d 41

5/19/1969 0.070 42

12/7/1967 0.084 43

10/15/1968 0.084 44

7/31/1969 0.084 45

12/7/1967 0.098 46

12/13/1967 0.098d 47

8/30/1970 0.098 48

12/5/1967 0.126 49

5/20/1969 0.126 50

11/17/1970 0.126 51

8/6/1968 0.140 52

9/7/1967 0.154 53

8/27/1970 0.154 54

12/28/1967 0.182 55

2/20/1969 0.182 56

8/23/1970 0.182 57

11/10/1967 0.224 58

2/5/1968 0.238 59

11/19/1967 0.266d 60

2/21/1969 0.266 61
Document No. ORAUT-TKBS-0043 Revision No. 01 Effective Date: 07/16/2008 Page 34 of 46

Sample date(s)

Result 

(dpm/d)a,b Sample #’s

8/1/1969 0.266 62

10/5/1967 0.280 63

10/8/1967 0.350 64

10/5/1967 0.392 65

12/13/1967 0.392 66

11/10/1967 0.434d 67

3/4/1968 0.476d 68

8/30/1968 0.602 69

10/4/1970 0.742 70

8/6/1968 0.966 71

8/5/1968 1.344 72

7/28/1968 1.400 73

8/5/1968 1.624 74

10/3/1967 2.856d 75

a. Results were normalized using 1.4 L/d. 

b. Assume DL = 0.03 dpm/L (normalized = 0.042 dpm 

/d). 

c. LOD = limit of detection.

d. Positive sample results all from one individual (a 

production worker/laboratory technician), who also 

had the highest result. This same individual had two 

<LOD results on October 11, 1967, and August 8, 

1968. 

Table 5-3. Summary of excretion rates from plutonium 

urinalysis results from operational period (dpm/d).

a

GM GSD 95th percentile

0.058 5.6 1.01

a. Excretion rates were determined by log-transforming the data, 

calculating the z-score, plotting the data, associated R2 fit 

parameter to estimate a 50th-and 95th-percentile values 

(ORAUT 2005f).

The excretion rates from Table 5-3 were modeled in IMBA, assuming a chronic intake from the date of 

the first bioassay sample on September 7, 1967, through the date of the last bioassay sample on 

November 21, 1970, and using an effective bioassay sample date of the midpoint of this intake period. 

Based on this analysis, an intake rate of 563 dpm/d was calculated (based on the GM). The intake 

rate at the 95th-percentile value was 9,760 dpm/d. 

The plutonium intake during the operational period as calculated above was used as the basis for 

plutonium intakes at the start of the residual period (1971). The 95th-percentile intake rate (dpm/d) 

was used to calculate an associated daily air concentration (assuming 250 workdays per calendar 

year and a breathing rate of 9.6 m3

/d). This daily air concentration was then used to estimate the 

residual surface concentration with guidance from the technical information bulletin (ORAUT 2008). 

The annual deposition amount was estimated using a deposition velocity of 0.00075 m/s, with 

deposition assumed to occur for 1 year. Using this approach a surface concentration of plutonium is 

estimated as follows.

 air concentration pCi/m3

 × 31,536,000 s/yr × 0.00075 m/s = surface activity dpm/m2 (Eq. 2)

The deposited material was assumed to have been resuspended and inhaled during the residual 

period. The calculated intake rate was 84,000 dpm/yr. Plutonium intakes during subsequent residual 

period years was calculated based on air monitoring data available for the period from 1976 through 

1989 (Table 5–4). The highest sitewide air concentration for the period was used to estimate an 
Document No. ORAUT-TKBS-0043 Revision No. 01 Effective Date: 07/16/2008 Page 35 of 46

intake rate starting in 1976 that was assumed to remain constant. Because the data is in fractions of 

the maximum permissible concentration (MPC), it was necessary to assume that all of the gross alpha 

air activity was plutonium and using the plutonium MPC value (2 × 10-12 μCi/ml). The intake rate that 

was calculated for 1976 and subsequent years was 5,168 dpm/yr. Intake rates for 1972 through 1975 

were calculated based on the 1971 and 1976 values by fitting them to an exponential function. 

Plutonium intakes during the residual period are summarized in Table 5-5. 

The internal dose analysis should include the potential inadvertent ingestion of plutonium activity 

based on the guidance in NIOSH (2004). The daily intake rate (dpm/d) was estimated as 0.2 times 

the daily air concentration. The estimated internal doses are assigned as a constant distribution in 

IREP. 
Document No. ORAUT-TKBS-0043 Revision No. 01 Effective Date: 07/16/2008 Page 36 of 46

Table 5-4. Air-sampling results in percentage of MPCsa during the residual period, 1976 to 1989.

b 

Building/area/process 1976 1977 1978 1979 1980 1981 1982 1983 1984 1985 1986 1987 1988 1989 Average

105 (302) Laboratories 3.90 4.60 3.20 3.90 4.55 5.15 3.00 1.75 1.48 0.58 1.42 1.46 0.71 0.62 2.59

100 Decontamination Laundry 1.40 0.90 1.50 2.20 1.70 1.75 1.85 1.44 1.76 0.83 1.25 1.14 0.31 0.43 1.32

105 Hall 1.39 0.68 2.09 2.18 1.01 0.94 1.38

110A, B LEU Scrap Rec, Store 

& Staging 

1.60 2.20 2.10 3.20 4.00 4.56 0.95 0.95 1.01 0.56 0.60 1.48 0.24 0.14 1.68

110C, D Pu U-233 Labs 19.20 22.00 22.60 14.30 24.00 15.35 6.30 6.25 17.13 11.88 22.57 8.19 2.76 1.64 13.87

110D-1 U Spectrographic lab 4.70 3.00 1.50 5.00 2.40 1.80 0.65 0.41 0.43 0.43 0.64 0.40 2.39 1.27 1.79

111 Chemical LEU scrap 

recovery

3.10 4.60 3.60 5.10 5.65 4.45 4.35 1.58 3.64 3.38 2.24 1.21 0.15 0.13 3.08

120 Maintenance 1.40 1.30 3.70 3.20 2.10 2.45 1.95 2.02 1.22 0.50 1.42 0.77 0.19 0.14 1.60

130 (Metals) LEU UF6 Cyl 

wash

6.50 1.70 3.80 5.60 16.05 43.80 9.80 0.59 0.69 0.95 1.92 1.95 0.32 0.85 6.75

131 Pilot Plant–R & D 2.20 0.60 1.90 3.00 1.85 2.25 0.95 0.51 0.51 0.34 0.57 0.40 0.15 0.05 1.09

220 NDA Scanning 1.10 1.10 1.80 4.20 1.35 1.90 1.35 1.16 0.81 0.67 1.98 1.09 1.54

230 Complex 2.11 0.70 1.86 1.76 0.66 0.79 1.31

233 HEU Scrap Recovery 31.70 14.60 15.20 12.50 10.40 9.35 7.50 5.54 4.31 3.10 5.50 3.61 3.61 3.27 9.30

234A Pu Production Areas 3.20 6.30 18.00 20.40 10.00 11.90 8.40 7.15 26.13 11.88 10.63 0.00 0.00 0.00 9.57

234B U-233 Process Area 0.30 0.30 0.50 0.40 0.45 0.50 0.35 0.45 0.52 0.25 0.00 0.00 0.00 0.00 0.29

234C Pu Dissolution Area 26.00 6.40 21.00 12.70 10.50 10.00 7.15 8.75 18.38 11.46 9.00 0.00 0.00 0.00 10.10

234C Fabrication Area 48.50 6.40 18.10 13.50 10.75 11.55 8.55 10.13 22.38 11.47 9.75 0.00 0.00 0.00 12.22

234 building 8.25 7.79 3.62 1.63 5.32

300 Nuclear Materials 

Warehouse

0.60 0.80 1.20 2.40 1.55 2.10 0.85 0.60 0.47 0.33 0.33 0.31 0.00 0.00 0.82

301 LEU Oxide Production 0.40 0.50 4.00 10.70 2.95 3.30 1.20 1.39 0.66 0.46 2.00 0.98 0.36 0.58 2.10

302 Finished Fuel Production 38.60 25.10 20.40 24.70 12.25 14.35 13.90 16.06 6.52 3.59 7.46 4.39 5.97 7.92 14.37

302 Retort Tube Cleaning 32.30 32.70 40.00 31.20 23.05 22.90 22.15 31.36 11.75 5.47 0.00 0.00 0.00 0.00 18.06

302 Incinerator 9.80 10.50 10.80 30.00 15.55 31.40 10.50 9.19 4.27 1.90 0.00 0.00 0.00 0.00 9.56

302 Lunchroom 9.50 8.50 5.00 4.80 3.25 2.75 0.00 0.00 0.00 0.00 3.38

303 Finished Fuel Production 27.20 27.00 14.50 29.90 18.05 25.80 18.55 13.04 5.16 7.63 11.11 9.68 8.07 10.22 16.14

304 Finished Fuel Warehouse 3.50 3.00 4.50 7.40 6.40 7.70 3.60 3.27 1.98 1.22 2.21 1.58 0.95 1.06 3.45

304 Warehouse excluding 

storage

2.15 1.22 3.34 3.54 0.00 0.00 1.71

305 Lunchroom 1.78 0.69 0.86 0.57 0.26 0.18 0.72

306 Complex 0.99 1.07 1.74 2.18 0.00 0.00 1.00

310 Scrap Metal Warehouse 0.50 0.40 1.80 4.00 2.75 3.65 1.10 0.39 0.51 0.35 0.51 0.29 0.00 0.00 1.16

a. MPC is considered to be the most restrictive NRC limit whether soluble or insoluble (i.e., 2 × 10-12 μCi/ml for plutonium). 

b. Data adapted from Tester (2005b,c,d,e).
Document No. ORAUT-TKBS-0043 Revision No. 01 Effective Date: 07/16/2008 Page 37 of 46

 Table 5-5. Plutonium air concentration and annual intake in the 

residual period. 

Year Air concentration

(dpm/m3)

Inhalation

(dpm/yr)

Ingestion

(dpm/yr)

1971 35 8.4E+04 1.8E+03

1972 29 7.0E+04 1.5E+03

1973 20 4.8E+04 1.0E+03

1974 11 2.8E+04 5.5E+02

1975 5.4 1.3E+04 2.7E+02

1976 2.1 5.2E+03 1.1E+02

1977–present 2.1 5.2E+03 1.1E+02

 

5.1.4

Air-sampling data from thorium processing facilities, which is summarized in ORAUT-OTIB-0070 

(ORAUT 2008), were used as the basis for airborne thorium concentrations at the end of the 

operational period. Data from the Lindsay Light facility were used as a surrogate for W.R. Grace. 

The Lindsay Light facility was selected as a basis for the upper bound of the radiological conditions in 

thorium process areas at W.R. Grace due to the nature of the activities at Lindsay (operations with 

thorium compounds such as thorium nitrate and thorium oxide) and the likely presence of these types 

of materials at W.R. Grace. Radiological conditions during the meteorological processing of thorium 

are bounded by those during the processing of the precursor materials. Since W.R. Grace handled 

both the source materials (thorium compounds) and produced thorium metals, the use of the data 

from Lindsay is believed to be bounding for all of the activities performed at W.R. Grace. The 95thpercentile air concentration (411 dpm/m3

) was assumed to be bounding for the general area air 

concentration at the end of the operational period (1971). This air concentration was then used to 

estimate the residual surface concentration with the guidance from the technical information bulletin 

(ORAUT 2008). The annual deposition amount was estimated using a deposition velocity of 

0.00075 m/s, with deposition assumed to occur for 1 year. Using this approach, a surface 

concentration of thorium was estimated as follows.

Thorium

air concentration pCi/ m3 × 31,536,000 s/yr × 0.00075 m/s = surface activity dpm/m2 (Eq. 3)

This mean surface concentration is favorable to claimants at the end of the operating period. The 

deposited material was assumed to have been resuspended and inhaled during the residual period. 

The amount of resuspension was assumed to be reduced with time due to fixing of the material on 

surfaces and to also due to depletion (ORAUT 2008). The depletion factors that were applied to each 

year are described in Table 3-1 of ORAUT (2008). The depletion factors indicate the residual 

concentration at the end of the operational period is to be used for the first year, the second year 

concentration is reduced by a factor of 0.03, and the concentration for the third and remaining years is 

reduced by a factor of 0.0007 (representing a constant concentration after the third year). 

Tabulated values represent total thorium activity; as such, they should be distributed between 232Th 

and 228Th in a manner favorable to claimants (West 1965; GGA 1971) [12]. An equal intake of 228Ra 

should be added.

The intakes in Table 5-6 can be used to estimate the internal dose to the target organ for the years of 

employment for the worker. The estimated internal doses should be assigned as a constant 

distribution. 
Document No. ORAUT-TKBS-0043 Revision No. 01 Effective Date: 07/16/2008 Page 38 of 46

Table 5-6. Thorium air concentration and annual intake in the residual period. 

Nuclide

Intake (pCi/yr)

1971 1972 1973–present

Th-232/Th228 1.1E+04 Inhalation

2.3E+02 Ingestion

3.2E+02 Inhalation

6.7E+00 Ingestion

7.4E+00 Inhalation

1.5E-01 Ingestion

Ra-228 1.1E+04 Inhalation

2.3E+02 Ingestion

3.2E+02 Inhalation

6.7E+00 Ingestion

7.4E+00 Inhalation

1.5E-01 Ingestion

The internal dose analysis should include the potential inadvertent ingestion of thorium activity using 

the guidance in NIOSH (2004). The daily intake rate is estimated as 0.2 times the average daily air 

concentration. Using the air concentration for the first year after the end of operations (1971) from 

Table 5-6, this value should be applied for all years of the residual period using the depletion factors 

as described in Table 3-1 of ORAUT (2008). 

5.2 ESTIMATION OF EXTERNAL EXPOSURE FROM RESIDUAL RADIOACTIVITY 

Radiation exposure records from 1971 to the present should be used to estimate external exposure to 

residual activity, as described in Section 4.0. The measured dose includes the NFS non-DOE 

operational dose plus whatever DOE residual dose may exist. As such, the measured dose will overestimate the dose from DOE operations only. 

The lack of a neutron personnel dosimetry program during the residual period is an indication of the 

low potential for neutron exposure. It is possible, however, that some minimal exposure to neutron 

radiation (from contaminated surfaces or held-up material) did occur. Any additional dose from 

unmonitored neutron exposure is likely bounded by the overall measured photon dose which includes 

dose from non-EEOICPA activities in addition to any residual dose covered under the EEOICPA 

program.

In the absence of claim-specific information, the following assumptions should be made about 

external exposures [13]:

Residual period Sources of external exposure Default assumptions

1971–to the present Photons, 30–250 keV

Electrons, > 15 keV or <30 keV

Exposure (R) to Organ Dose (HT)

Follow ORAUT-OTIB-0017 

(ORAUT 2005d) requirements

For unmonitored workers, dose information from site external radiation summary reports (available 

from 1976 through 1988) can be used as a basis for external dose. The penetrating exposure of 

100 mrem (Table 5-7 under the average row for 1978) was the maximum listed value from 1976 

through 1988. This 100-mrem penetrating exposure value between the external dose equivalent 

values was calculated from possible contamination source terms. It is not likely that unmonitored 

coworkers would have received more external dose from residual AEC sources than that recorded for 

coworkers who were monitored for NRC- or DOE-related activities [14].

For unmonitored workers the dose reconstructor should take the following steps:

1. For the penetrating dose estimate, take the highest annual coworker average penetrating 

exposure and add a missed dose calculation. The highest annual exposure of 100 mrem 

should be applied to all years of unmonitored residual years. The 100 mrem is based on the 

recorded coworker exposure history from Table 5-7 for 1978.

An example calculation is as follows for the year:
Document No. ORAUT-TKBS-0043 Revision No. 01 Effective Date: 07/16/2008 Page 39 of 46

Use the coworker penetrating exposure estimate of 100 mrem/yr, and add the missed dose 

calculation (40 mrem per badge cycle is the limit of detection from Table 4-2 for the 

Landauer film whole-body period of January 1964 to December 1988), which is 

40 mrem ÷ 2 × 11 months = 220 mrem (The 100-mrem coworker dose counts as a 

recorded dose and therefore there are only 11 months of unrecorded or missed dose.)

100 mrem (high average coworker) + 220 mrem (missed dose) = 320 mrem

A value of 320 mrem/yr for penetrating dose should be assigned to unmonitored workers, 

assigned as a constant distribution, and a 30-250 keV photons energy range. 

2. For the nonpenetrating dose estimate, take the highest average annual coworker average 

nonpenetrating exposure and do not add a missed dose calculation. 

An example calculation is as follows for the year:

Coworker nonpenetrating exposure estimate, which is 390 mrem/yr (390 mrem based on 

the highest average recorded coworker nonpenetrating exposure history in Table 5-7, 1976 

row). There is no necessity to add a missed dose calculation to the nonpenetrating dose 

estimate because it has already been accounted for in the missed dose penetrating 

estimate (ORAUT 2005d). 

390 mrem (high average coworker) 

A value of 390 mrem/yr for nonpenetrating dose should be assigned to unmonitored workers, 

assigned as a constant distribution, and as electrons > 15 keV or < 30 keV photons, consistent 

with the approach described in Section 4.4.1.
Document No. ORAUT-TKBS-0043 Revision No. 01 Effective Date: 07/16/2008 Page 40 of 46

Table 5-7. W.R. Grace/NFS external exposure summary (NFS 1980, 1980-1989).

a

Penetrating 

exposure range 1976 1977 1978 1979 1980 1981 1982 1983 1984 1985

DDE 

range 1986 1987 1988

No measurable exposure 320 322 297 364 149 316 147 353 278 309 M 500 532 603

<100 mrem 175 144 165 150 504 496 943 546 594 560 >M-50 253 319 348

>100 mrem <250 mrem 7 44 60 1 101 107 70 44 31 2 >50-100 127 24 56

>250 mrem <500 mrem 0 1 8 0 5 6 1 1 1 0 >100-150 19 2 2

>500 mrem <1,500 mrem 0 0 3 0 1 0 0 0 0 0 >150 5 4 1

Max 180 340 160

Total 502 511 533 515 760 925 1,161 944 904 871 Total 904 881 1,010 Total average

High 0.25 0.50 1.50 0.25 1.50 0.50 0.50 0.50 0.50 0.25 0.18 0.34 0.16 0.53

Average 0.05 0.08 0.10b 0.05 0.07 0.08 0.06 0.06 0.06 0.05 0.04 0.03 0.03 0.06

Nonpenetrating 

exposure range

SDE 

range

No measurable exposure 381 404 417 431 149 316 946 799 277 307 M 493 522 592

<600 mrem 117 107 113 84 609 606 218 145 627 564 >M-50 237 320 358

>600 mrem <1,500 mrem 1 0 3 0 2 3 0 0 0 0 >50-100 147 33 57

>1,500 mrem <3,000 

mrem

1 0 0 0 0 0 0 0 0 0 >100-150 19 2 2

>3,000 mrem <4,500 

mrem

1 0 0 0 0 0 0 0 0 0 >150 8 4 1

>4,500 mrem 1 0 0 0 0 0 0 0 0 0

Max 290 340 160

Total 904 881 1,010 Total average

High 4.52 0.60 1.50 0.60 1.50 1.50 0.60 0.60 0.60 0.60 0.29 0.34 0.16 1.03

Average 0.39c 0.30 0.32 0.30 0.30 0.30 0.30 0.30 0.30 0.30 0.04 0.03 0.03 0.25

a. W.R. Grace did not provide averages and collective dose before 1977; the average has to be approximated. This will sometimes overestimate the dose. NFS Radiation Exposure 

Summary Reports are available for 1977-2005. 

b. This represents the highest value of the average penetrating dose for years 1976-1988. 

c. This represents the highest value of the average non-penetrating dose for years 1976-1988.
Document No. ORAUT-TKBS-0043 Revision No. 01 Effective Date: 07/16/2008 Page 41 of 46

6.0

It is not necessary to include an environmental dose component for W.R. Grace worker dose because 

all workers are assumed to have been exposed to operational conditions, and dose has been 

assigned accordingly. 

ENVIRONMENTAL DOSE 

7.0

Where appropriate in this document, bracketed callouts have been inserted to indicate information, 

conclusions, and recommendations provided to assist in the process of worker dose reconstruction. 

These callouts are listed here in the Attributions and Annotations section, with information to identify 

the source and justification for each associated item. Conventional references, which are provided in 

the next section of this document, link data, quotations, and other information to documents available 

for review on the Project’s Site Research Database.

ATTRIBUTIONS AND ANNOTATIONS

[1] Guido, Joseph. ORAU Team Member. March 2008 

Because it is unknown, it is favorable to assume RU radionuclides were present during the 

entire operational period.

[2] Demopoulos, Paul J. ORAU Team. Senior Health Physicist. March 2006. 

Based on personal observations during review of claimant uranium urine results, if the 

recorded result is between 2 and 10 dpm/L, it is reasonable to assume 2 dpm/L as the MDL. If 

there are zeros or “<10 dpm/L” recorded for uranium results, then the dose reconstructor 

should assume the MDL is 10 dpm/L. 

[3] Demopoulos, Paul J. ORAU Team. Senior Health Physicist. March 2006. 

An assumption of 10-year-old fuel-grade plutonium is favorable to claimants and reasonable in 

the absence of claimant-specific information. This is also noted in the Hanford site profile

(ORAUT 2007b). 

[4] Demopoulos, Paul J. ORAU Team. Senior Health Physicist. March 2006. 

Personal observations during review of plutonium bioassay records show in general that 

plutonium urine sampling occurred from 1967 through approximately 1973.

[5] Demopoulos, Paul J. ORAU Team. Senior Health Physicist. March 2006. 

Personal observations during a review of plutonium bioassay records and plutonium DLs and 

MDAs from other sites show that a plutonium DL of 0.03 dpm/L and a MDA of 0.06 dpm/L is 

reasonable.

[6] Demopoulos, Paul J. ORAU Team. Senior Health Physicist. March 2006. 

The observation that the reported skin dose represents only the shallow dose component 

before 1970, and the dose after January 1, 1970, includes the penetrating gamma component, 

is from a cursory review of claimant dosimetry records.

[7] Demopoulos, Paul J. ORAU Team Member. Senior Health Physicist. March 2006. 

Based on personal observations during a review of dosimetry records. Actual recorded badge 

frequencies should always be used in lieu of these default assumptions. 

[8] Smith, Matthew, and Winslow, Rob, ORAU Team Members. June 2007. 

In the absence of further information, it is appropriate to apply these DCFs. 
Document No. ORAUT-TKBS-0043 Revision No. 01 Effective Date: 07/16/2008 Page 42 of 46

[9] Thomas, Elyse M. ORAU Team. Principal Medical Dosimetrist, February 2007. 

The default assumption for the frequency of chest X-rays for AWE sites is a PA chest X-ray at 

preemployment, annually, and at termination, as stated in an e-mail from E. Thomas to C. 

Bloom (February 23, 2007). 

[10] Thomas, Elyse M. ORAU Team. Principal Medical Dosimetrist. February 2007 

It is reasonable to assume that PFG was not performed unless there is information in the 

claimant files to suggest otherwise, as stated in an e-mail from E. Thomas to C. Bloom 

(February 23, 2007). 

[11] Labone, Thomas. ORAU Team. Deputy Principal Internal Dosimetrist. April 2007. 

Statistical analysis was performed by Thomas Labone. 

[12] Olsen, Bernard M. ORAU Team. Senior Health Physicist. February 2008.

The ratio of 232Th to 228Th is variable and ranges from equal activity concentrations minimum 

condition of 228Th being present at approximately 40% the activity concentration of 232Th (West 

1965; GGA 1971). The relative ratio of 232Th to 228Th is significant for dose reconstruction. 

Organ dose that is determined by applying the DCFs of ICRP Publication 68 (ICRP 1995) is 

maximized for nonrespiratory cancers by assuming all thorium is 232Th, while the dose is 

maximized for respiratory cancers by assuming that all thorium is 228Th. For a nonrespiratory 

cancer, it is favorable to the claimant to assume that all the thorium is 232Th. For a respiratory 

cancer, it is favorable to the claimant to assume that all the thorium is 228Th. For a bestestimate evaluation, dose reconstructors should assume that 232Th and 228Th are present in 

ratios of 1 to 1 for a respiratory cancer, or 7 to 3for a nonrespiratory cancer (the minimum 

condition of 228Th present at 40% that of 232Th) . 

[13] Fix, Jack, Smith, Matthew, and Winslow, Rob. ORAU Team. ORAU Team Members. March 

2008. 

Default DCFs from NIOSH (2007a) were identified for photons and electrons for the respective 

operational and residual periods. 

[14] Demopoulos, Paul J. ORAU Team. Senior Health Physicist. April 2007. 

The residual period AEC source term was not likely to cause exposures above those during 

the AEC operational period because contamination levels would likely decrease, not increase, 

with the cessation of AEC-related operations. The amount of this decrease at this time cannot 

be defined.
Document No. ORAUT-TKBS-0043 Revision No. 01 Effective Date: 07/16/2008 Page 43 of 46

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