ORAU TEAM

Dose Reconstruction

Project for NIOSH

Oak Ridge Associated Universities I Dade Moeller I MJW Technical Services

Page 1 of 43

DOE Review Release 09/26/2011

Document Title:

An Exposure Matrix for W.R. Grace and 

Company in Erwin, Tennessee

Document Number: ORAUT-TKBS-0043

Revision: 02

Effective Date: 09/16/2011

Type of Document: TBD

Supersedes: Revision 01

Subject Expert(s): Mutty M. Sharfi

Approval: Signature on File Approval Date: 09/14/2011

Karen S. Kent, Document Owner

Concurrence: Signature on File Concurrence Date: 09/14/2011

John M. Byrne, Objective 1 Manager

Concurrence: Signature on File Concurrence Date: 09/14/2011

Edward F. Maher, Objective 3 Manager

Concurrence: Vickie S. Short Signature on File for Concurrence Date: 09/14/2011

Kate Kimpan, Project Director

Approval: Signature on File Approval Date: 09/16/2011

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. 02 Effective Date: 09/16/2011 Page 2 of 43

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.

09/16/2011 02 This revision was initiated to modify the residual period guidance. 

This includes a recommendation to not assign plutonium and thorium 

exposures during the residual period; modified the methodology for 

assigning external dose and uranium exposure during the residual 

period; and revised the use of monitoring records during the residual 

period. Also, an error was corrected for 238U in Table 3-3. In 

addition, references were updated as appropriate and various minor 

technical edits were made throughout this document. The SEC 

language in Section 1.0 was moved to a newly created Sections 1.3. 

These changes were made to Sections 3.1, 3.1.3, 4.7, 5.1, 5.2, and 

the references. Incorporates formal internal and NIOSH review 

comments. Constitutes a total rewrite of the document. Training 

required: As determined by the Objective Manager. Initiated by 

Karen S. Kent.
Document No. ORAUT-TKBS-0043 Revision No. 02 Effective Date: 09/16/2011 Page 3 of 43

TABLE OF CONTENTS

SECTION TITLE PAGE

Acronyms and Abbreviations ...............................................................................................................6

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

1.1 Purpose........................................................................................................................8

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

1.3 Special Exposure Cohort Petition Information for W. R. Grace ................................... 10

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 

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........................................................................................... 22

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 Non-Penetrating Dose..................................................................................... 28

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

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

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

4.4 Missed Dose...............................................................................................................29

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.........................................................................32

5.2 Estimation of External Exposure from Residual Radioactivity ..................................... 35

6.0 Environmental Dose............................................................................................................... 37

7.0 Attributions and Annotations .................................................................................................. 37

References ........................................................................................................................................39
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LIST OF TABLES

TABLE TITLE PAGE

2-1 W.R. Grace and NFS AEC period and locations .................................................................... 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, total core....................................... 18

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

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

recovery ................................................................................................................................. 18

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............................................... 19

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 ................................................................................................................................20

3-10 Estimated uranium inhalation intake rates based on measured timeweighted 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 ................................................ 23

3-13 Chest count conversions for uranium source terms................................................................ 23

3-14 Activity composition of Hanford Site reference fuel-grade plutonium 

mixture...................................................................................................................................24

3-15 Uranium internal exposure summary for operational period January 1, 

1958, to December 31, 1970.................................................................................................. 26

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 Source-term depletion rates during residual period ................................................................ 32

5-2 Uranium (and RU components) annual inhalation rates for the residual 

period..................................................................................................................................... 33

5-3 Uranium (and RU components) annual ingestion rates for the residual 

period.....................................................................................................................................34

5-4 External default assumptions during the residual period ........................................................ 35

5-5 External annual exposure rates for the residual period........................................................... 36

LIST OF FIGURES

FIGURE TITLE PAGE

2-1 Location of Erwin, Tennessee ................................................................................................ 11

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

Bq becquerel

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

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

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

MOX mixed oxide

mR milliroentgen
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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

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
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1.0 INTRODUCTION

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

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

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

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

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

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

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

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

employer [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 2010):

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

conventional structures

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

reasons

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 2010). 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.

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

Grace and Company plant. 

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 
Document No. ORAUT-TKBS-0043 Revision No. 02 Effective Date: 09/16/2011 Page 9 of 43

in W.R. Grace facilities and to occupationally required medical X-ray examinations performed for 

screening. 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 describe 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 U.S. Atomic Energy Commission (AEC) contract years from the late 1950s through 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 were collected in the 1960s 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.

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 

used Nuclear Chicago Company and then Landauer, Inc., 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.
Document No. ORAUT-TKBS-0043 Revision No. 02 Effective Date: 09/16/2011 Page 10 of 43

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.

1.3 SPECIAL EXPOSURE COHORT PETITION INFORMATION FOR W. R. GRACE

January 1, 1958 through December 31, 1970

NIOSH has determined, and the Secretary of the U.S. Department of Health and Human Services 

(DHHS) has concurred, that it is not feasible to reconstruct internal radiation dose for (Leavitt 2007): 

Atomic Weapons Employer (AWE) employees who were monitored or should have 

been monitored for potential exposure to thorium while working in any of the 100 series 

buildings or Buildings 220, 230, 233, 234, 301, or 310 at the W. R. Grace site at Erwin, 

Tennessee for a number of work days aggregating at least 250 work days from 

January 1, 1958, through December 31, 1970, or in combination with work days within 

the parameters established for one or more other classes of employees in the Special 

Exposure Cohort. 

Therefore, only internal dose from uranium and plutonium is considered in Section 3.0 of this site 

profile. Because of the identified dose reconstruction infeasibility, all dose reconstructions for all 

workers having employment during the SEC class period are considered partial dose reconstructions. 

If monitoring data are available for workers included in the SEC class, this data may be used to refine 

the dose reconstruction; however, such dose reconstructions are still considered partial dose 

reconstructions.

2.0 SITE DESCRIPTION, HISTORY, AND PROCESS

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). 

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

in 1986 (Congress 1986, p.108).

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

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

License SNM-124, as amended. 

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 
Document No. ORAUT-TKBS-0043 Revision No. 02 Effective Date: 09/16/2011 Page 11 of 43

Figure 2-1. Location of Erwin, Tennessee. 

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

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 a 

series of Feasibility Reports (W.R. Grace 1959; Glauberman 1962; Housholder and Runion 1962; 

Housholder 1963a,b). Inherent in all the operations is nuclear criticality safety that governs not only 

the operations and storage but also the movement of material in 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 

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

a,b

Operations Period of AEC work 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 fuel and MOX 1/1/1966 through 12/31/1970 234A, 234B, 234C, 110C, 110D

HEU scrap recovery 1/1/1958 through 12/31/1970c 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

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

b. Listed activities include both weapons-related work and non-weapons-related work during the AEC contract period. 

However, dose from non-weapons-related work is only considered during the AEC contract years, 1958 through 1970. 

c. 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).

whenever possible. For birdcage units, the two control parameters are geometry (birdcage 

dimensions) and a 235U mass limit that can be placed in 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 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-in. × 2-in. metal frame 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 review of the available literature shows that no 

criticality accidents have occurred during W.R. Grace and later NFS operations.

2.1.1 Production of Uranium Metal and Uranium Metal Alloys Enriched to 

12% Uranium-235

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.

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, pp. 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. As 

many as six cylinders could be placed in 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 entrained UF4 dust was collected in additional product 

cans. The offgas was then filtered to remove the last traces of UF4 dust and 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.
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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, derbies were broken into smaller pieces on a large hydraulic press before shipment. 

On other occasions, derbies were remelted and cast into various shapes such as slugs, rods, and 

plates. In addition, the uranium could be alloyed with other metals during remelting and casting. 

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

above. 

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 Scrap Recovery Operations

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

process steps were assumed to be 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. 

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 g of 235U per liter. There were 

solutions ranging from 2.8 to 9.9% enrichment at a concentration of 5 g of 235U per liter. From these 

two types of solutions, a third solution was formed that did not exceed 3% enrichment, and contained

a maximum of 88 lb of uranium (39.9 kg) 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 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)
Document No. ORAUT-TKBS-0043 Revision No. 02 Effective Date: 09/16/2011 Page 14 of 43

A summary of the HEU scrap recovery follows (Housholder and Runion 1962, pp. 122-123).

2.1.3 Highly Enriched Uranium from Scrap Recovery

2.1.3.1 Receiving and Storage

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, 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 apparently

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, 

pp. 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. 02 Effective Date: 09/16/2011 Page 15 of 43

2.1.4 Production of Uranium Oxide Mixed with Thorium Oxide and Zirconium Oxide 

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

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

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

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

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. 02 Effective Date: 09/16/2011 Page 16 of 43

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 in 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 Production of Mixed Oxide Fuel

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 in shielded 55-galdrum birdcages for temporary storage and shipment to the customer. 

2.1.6 Uranium Hexafluoride Cylinder Washing

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.

3.0 ESTIMATION OF INTERNAL EXPOSURE

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. 

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, 232Th, 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 enrichments and chemical forms of the processed radionuclides. 
Document No. ORAUT-TKBS-0043 Revision No. 02 Effective Date: 09/16/2011 Page 17 of 43

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.

3.1 URANIUM EXPOSURES

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

uranium occurred in most of the buildings, with the exception of Buildings 110 and 234 where 

plutonium was primarily processed.

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

could have been exposed versus another. 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). 

International Commission on Radiological Protection (ICRP) Publication 68 (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 (0.72%) 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

DU (0.2%) IMBA 0.402 0.155 0.011 0.0005 0.834

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 

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

because it is more favorable to claimants. However, if only chest count data is available, then 233U 

fuel mixture should not be used since chest counts would not be used to detect the 233U fuel mixture 

given the small abundance of 235U. In this case the most favorable to claimant mixture (between DU, 

NU, LEU, and HEU) should be used.
Document No. ORAUT-TKBS-0043 Revision No. 02 Effective Date: 09/16/2011 Page 18 of 43

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

Isotope Mass (g)

Specific activity 

(Bq/g) Total activity % activity % weight

U-232 3.7795 7.93E+11 3.00E+12 1.63 0

U-233 500,600.60 3.57E+08 1.79E+14 97.5 2

U-234 6,547.90 2.31E+08 1.51E+12 0.82 0

U-235 440.90 8.01E+04 3.53E+07 0.00 0

U-236 112.80 2.40E+06 2.70E+08 0.00 0

U-238 1,900.60 1.25E+04 2.38E+04 0.00 0

Th-232 23,481,000.00 4.06E+03 9.54E+10 0.05 98

Total 23,990,606.58 1.84E+14 100

3.1.1 Uranium Air Sampling

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

Laboratory 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, pp. 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. 

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.

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
Document No. ORAUT-TKBS-0043 Revision No. 02 Effective Date: 09/16/2011 Page 19 of 43

Operation

Number of 

samples

Average concentration 

(dpm/m3)

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 laboratory 3 3

Spectrographic laboratory 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.

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.
Document No. ORAUT-TKBS-0043 Revision No. 02 Effective Date: 09/16/2011 Page 20 of 43

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 glovebox 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. 

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 laboratory 4 4.3

Spectrographic laboratory 2 9.3

Research and development 

laboratory

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.
Document No. ORAUT-TKBS-0043 Revision No. 02 Effective Date: 09/16/2011 Page 21 of 43

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) by assuming a 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 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 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 task-oriented, timeweighted, average air concentrations is favorable to claimants.

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. 

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 ICRP Publication 

66 (ICRP 1994). This category assumes 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.

For the case in which 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 
Document No. ORAUT-TKBS-0043 Revision No. 02 Effective Date: 09/16/2011 Page 22 of 43

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 6.24E+5

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/workday)

a Intake (pCi)

1 yr 1 250 52 1.30E+4

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 Enriched and Recycled Uranium

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.

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 233U and as such should be applied regardless of whether 233U or 234U is assumed 

(Tomlinson 1964, p. 6).

3.1.3 Uranium Bioassay

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

Services 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 
Document No. ORAUT-TKBS-0043 Revision No. 02 Effective Date: 09/16/2011 Page 23 of 43

 Table 3-12. Intakes of contaminants in RU as

 fractions of uranium intake.

a

Radionuclid

e

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).

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

2005, 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 less than 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, the dose reconstructor 

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

use them 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.

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 235U was about 120 μg 

(Tester 2005, p. 2).

A summary of the chest count conversions is shown in Table 3-13.

Table 3-13. Chest count conversions for uranium source terms. 

Mixture

Detected

isotope

U-235 

MDA

(ug)

pCi total U 

per ug U235b

Total U 

MDA

(pCi)

dpm total 

U per ug 

U-235b

Total U 

MDA

(dpm)

Uranium–depleted U-235 120 201 24,120 446 53,546

Uranium–natural U-235 120 95 11,377 210 25,257

Uranium–2% enriched U-235 120 81 9,696 179 21,525

Uranium–3.5% enriched U-235 120 63 7,543 140 16,745

Uranium–93% enriched U-235 120 73 8,787 163 19,507

Uranium-233 fuel mixture U-235 120 1.18E+07 1.41E+09a 2.62E+07 3.14E+09a

a. If only chest count measurements exist in a claimant’s records, then U-235 chest count measurements should not be 

used to assess U-233 fuel mixture intakes. Due to the low U-235 content in the U-233 fuel mixture, the site would not 

have used U-235 chest counting to detect U-233 intakes.

b. These specific activities can also be used to convert chest counts results reported in U-235 mass to total uranium 

activity.

3.2 PLUTONIUM EXPOSURES

Many forms of plutonium were possible over the years including metal and oxide. The specific 

chemical forms of plutonium at the WR Grace site 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 
Document No. ORAUT-TKBS-0043 Revision No. 02 Effective Date: 09/16/2011 Page 24 of 43

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 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 accordance 

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

Retained in the Lung (ORAUT 2010a).

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 considered reasonable

and should be used in the absence of claimant-specific information [3]. A summary of the activity 

composition of the default reference fuel-grade plutonium mixture is presented in Table 3-14. 

 Table 3-14. 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

Pu-239+240:Am-241 2.05

Pu-239+240:Pu-238 5.05

Pu-241:Pu-239+240 24.0

a. Source: ORAUT (2010b) 

b. Time since separation of the Am-241 from

 the plutonium mix.

3.2.1 Plutonium Bioassay 

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 through February 1973; 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; NFS 2005, p. 2). 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. 02 Effective Date: 09/16/2011 Page 25 of 43

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 site profile for 

the Hanford Site (ORAUT 2010b) reports an MDA of 0.05 dpm/sample. The site profile for the 

Savannah River Site (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, an 

assumed DL of 0.03 dpm/L and an MDA of 0.06 dpm/L should be used [5]. 

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 non-detections 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, 

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

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 is 

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

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 reconstruct internal thorium 

dose adequately 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 during the operational period (1958-

1970) should be based on actual bioassay monitoring results for the individual. When uranium 

bioassay results are not available, the values in Table 3-15 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 a partial internal assessment.

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

Table 3-15 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. 
Document No. ORAUT-TKBS-0043 Revision No. 02 Effective Date: 09/16/2011 Page 26 of 43

 Table 3-15. Uranium internal exposure summary for

operational period January 1, 1958, to

December 31, 1970.

a

Internal Exposure Intake (pCi/d) IREP distribution

Uranium Inhalation

Ingestion

1.71E+03

3.56E+01

Constant

Pu-238 Inhalation

Ingestion

1.03E+02

2.14E+00

Constant

Np-237 Inhalation

Ingestion

8.55E+00

1.78E-01

Constant

Tc-99 Inhalation

Ingestion

6.84E+02

1.43E+01

Constant

Th-232 Inhalation

Ingestion

3.42E+01

7.13E-01

Constant

Th-228b Inhalation

Ingestion

3.42E+01

7.13E-01

Constant

Ru-106 Inhalation

Ingestion

6.84E+01

1.43E+00

Constant

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.

4.0 ESTIMATION OF EXTERNAL EXPOSURE

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.

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.

4.2 RECORDED EXTERNAL DOSE

Beginning in the late 1950s, Nuclear Chicago 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). In the approximate 1989 timeframe, Landauer supplied W.R. Grace/NFS with 

TLDs rather than 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 
Document No. ORAUT-TKBS-0043 Revision No. 02 Effective Date: 09/16/2011 Page 27 of 43

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. Before 1970, the reported skin dose represented only 

the non-penetrating dose component. As such, before, the total shallow dose is best estimated by 

summing the reported skin (i.e., non-penetrating dose) and whole-body (i.e., penetrating dose 

component) doses. 

Table 4-1 lists a more detailed description of the Nuclear Chicago Company or Landauer, Inc.,

dosimeters used at W.R. Grace and, later, NFS. The 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 types 

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 before 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.
Document No. ORAUT-TKBS-0043 Revision No. 02 Effective Date: 09/16/2011 Page 28 of 43

4.2.1 Non-Penetrating Dose

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

Before 1970, the reported skin dose represents only the non-penetrating 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 

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

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

4.2.2 Penetrating Dose

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 

low in comparison with non-penetrating exposure. The photon radiation is predominantly in the 30-to250-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. 

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 HEU 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.

4.3.1 Extremity Doses

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 high-energy 

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.
Document No. ORAUT-TKBS-0043 Revision No. 02 Effective Date: 09/16/2011 Page 29 of 43

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

Dosimetera Period of use MDL (rem)

Maximum annual missed dose (rem)

(MDL/2 × frequency)

Nuclear Chicago film–

whole body

1957–5/1959b 0.04 photons 0.520 beta-photons (every 2 weeks)

0.04 beta

Nuclear Chicago film–

whole body

6/1959–12/1960b 0.04 photons 0.240 beta-photons (monthly)

0.04 beta

Landauer film–

whole body

1/1961–12/1963b 0.04 photons 0.240 beta-photons (monthly)

0.04 beta

Landauer film–

whole body

1/1964–12/1988c 0.04 photons 0.240 beta-photons (monthly)

0.04 beta

Landauer TLD–

whole body

1/1989–12/1998c 0.04 photons 0.080 beta-photons (quarterly)

0.04 beta

Landauer OSL–

whole body

1/1/1999–12/2004c 0.001 photons 0.008 beta-photons (quarterly)

0.001 beta

Landauer TLD–

whole body

1/2005–presentc 0.02 photons 0.040 beta-photons (quarterly)

0.02 beta

Film badge–

wrist

1957–June 1974b 0.04 photons 0.240 beta-photons (monthly)

0.04 beta

Landauer Type M–

wrist

July 1974–about 1983c 0.02 photons 0.120 photons (monthly)

0.04 beta 0.240 beta (monthly)

G5 wrist film badge–

wrist

About 1983–1990c 0.02 photons 0.120 photons (monthly)

0.04 beta 0.040 photons (quarterly)

0.240 beta (monthly)

0.080 beta (quarterly)

K5 TLD–

wrist 

1991–presentc 0.04 photons 0.240 photons (monthly)

0.04 beta 0.240 beta (monthly)

Film badge–

finger

1957–1982c 0.04 photons 0.240 beta-photons (monthly)

0.04 beta

U3 TLD (LiF)–

finger ring

About 1983–presentc 0.03 photons 0.180 photons (monthly)

0.04 beta 0.060 photons (quarterly)

0.240 beta (monthly)

0.080 beta (quarterly)

a. TLD = thermoluminescent dosimeter.

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

b. MDL information communicated by Landauer. 

4.4 MISSED DOSE

MDL information for photon and beta whole-body dosimeter systems is 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 a different dosimeter exchange frequency occurred [7]. 

4.5 DOSE RECONSTRUCTION RECOMMENDATIONS

4.5.1 Beta Dose

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

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

cancer and the workplace practices. Non-penetrating dose can be considered for IREP input 

categories as either electrons greater than 15 keV or photons <30 keV, depending on 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 
Document No. ORAUT-TKBS-0043 Revision No. 02 Effective Date: 09/16/2011 Page 30 of 43

with the External Dose Reconstruction Implementation Guide (NIOSH 2007a). If a non-penetrating 

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 non-penetrating radiation (ORAUT 2005d).

4.5.2 Photon Dose

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 2006b). Dose 

reconstructors should use the Exposure (R) to Organ Dose (HT), [8] dose conversion factors from 

NIOSH for employment during AEC operations, (e.g. January 1, 1958, through December 31, 1970) 

(2007a). To ensure favorability to claimants, a photon energy range of 100% 30-to-250 keV should

be applied (ORAUT 2006c).

4.5.3 Glovebox Geometry Assumptions

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

benchtop geometry work activities with plutonium, as noted in DCAS-TIB-0010, Special External Dose 

Reconstruction Considerations for Glovebox Workers (DCAS 2010a). This would include chemical 

operators. Indicators of hands-on plutonium work could be plutonium bioassay results 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.

4.5.4 Uranium-233

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 m (Householder 1963b, pp. 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. 

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 Oak Ridge Associated Universities (ORAU) Team guidance, which is 

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

chest X-rays are assumed for all site workers based on information from the Technical Information 

Bulletin: Guidance on Assigning Occupational X-Ray Dose under EEOICPA for X-Rays Administered 

Off Site (ORAUT 2011c). Organ doses can be obtained from Dose Reconstruction from 

Occupationally Related Diagnostic X-Ray Procedures (ORAUT 2011b). 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. 02 Effective Date: 09/16/2011 Page 31 of 43

4.7 SUMMARY OF EXTERNAL DOSE RECONSTRUCTION RECOMMENDATIONS 

Guidance in the above sections to determine 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

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

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 

& OTIB0017

Lognormal 

GSD 1.52

Measured 

dose

1/1/1958–

12/31/1970

Photons, 30-250 

keV, AP, acute

Penetrating Recorded 

value

recorded 

dose

Measured 

dose

1/1/1958–

12/31/1970

Photons, <30 keV or 

electrons >15 keV, 

AP, acute

NonpenetratingRecorded 

value

recorded 

dose

Medical X-ray 1/1/1958–

12/31/1970

PA radiographic 

chest exam

Penetrating Initial plus 

annual 

examinations 

See Section 

4.4.5

See ORAUT 

(2011b) 

5.0 ESTIMATION OF EXPOSURE TO RESIDUAL ACTIVITY

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

December 31, 1970. The October 2009 Report on Residual Radioactive and Beryllium Contamination 

at Atomic Weapons Employer Facilities and Beryllium Vendor Facilities, noted that a significant 

potential for residual contamination remained after 1970. The residual contamination period at W.R. 

Grace is from January 1, 1971, through October 2009 (NIOSH 2009). 

Remediation of the Burial Grounds

Remediation activities of the burial grounds occurred during the residual period at WR Grace. The 

North site area was used in the past for waste storage and disposal activities related to its nuclear 

work. NFS has excavated and removed buried wastes and debris. Excavated wastes, debris, and 

contaminated soils have been packaged for shipment to and disposal at an off-site licensed disposal 

facility. Three surface impoundments, Ponds 1, 2, and 3 are located within the North site. These 

impoundments received liquid waste from on-site processing operations from 1957 until 1978. Also 

low-level, contaminated solid wastes were disposed of in the North site Burial Ground area from 1966 

until 1977, as authorized under 10 CFR 20.302. The contents and locations of most disposal pits are 

well documented. Another area previously used for solid waste disposal is the former Pond 4 area 

which is located west of the three impoundments. NFS removed waste materials from Ponds 1, 2, 

and 3 and the Pond 4 area from 1991 through 1996. NFS has also excavated waste and 

contaminated soil from the North site Burial Ground. Each former disposal area at the site has been 

identified as a Solid Waste Management Unit (SWMU) in the Hazardous and Solid Waste 

Amendments (HSWA) permit issued to NFS by the Environmental Protection Agency (EPA 1993a).

Based on this information, these remediation activities are assumed to have occurred between 1991

and 2009 (MACTEC 2009). The residual approach discussed below is to be applied to those workers 

who had the potential for internal radiation exposure, since they could have worked in the Burial 

Grounds and Ponds Area. This is a favorable to claimant assumption.
Document No. ORAUT-TKBS-0043 Revision No. 02 Effective Date: 09/16/2011 Page 32 of 43

5.1 INTERNAL EXPOSURE TO RESIDUAL ACTIVITY

During the residual radioactivity period at W.R. Grace (1971 through October 2009 ), personnel might 

have been internally exposed to uranium in the workplace. Internal monitoring data could exist for 

plutonium or thorium during the residual period; however, only operations involving uranium were part 

of the nuclear weapons-related program, and are covered during the residual period for dose 

reconstruction under EEOICPA (Leiton 2011). 

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

period due to uranium contamination.

Unless the potential for exposure can be ruled out, the default intakes below should be assigned. 

However, monitoring data can be used instead of the default intake assumptions to limit dose, as 

appropriate (NIOSH 2008a). These internal exposures can be limited using bioassay data for 

uranium, if they exist. Internal doses assessed from bioassay data during this period should be 

evaluated in accordance with the guidance in Section 3.0 for workers with the potential for internal 

radiation exposure and the depletion factors applied using the values in Table 5-1 of this site profile 

document. 

The uranium intakes (and associated RU components) in Section 3.0 (Table 3-15) can 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 (2008a). 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 also to depletion (ORAUT 2008a). NIOSH has previously determined the activity removal rates 

during the residual radiation periods for the four sites as summarized in Table 5-1 (DCAS-TKBS-0002

(DCAS 2010b); Battelle-TBD-6000 Appendix C (NIOSH 2008b); ORAUT-TKBS-0045 (ORAUT 

2008b); ORAUT-TKBS-0032 (ORAUT 2011a). In these cases, contemporary estimates of airborne 

radioactivity at the beginning and end of each site’s residual radiation period were used to estimate 

the effective exponential clearance of the contamination over an extended period of time for each site 

(ORAUT-OTIB-0070 (ORAUT 2008a)). 

The source-term depletion rates calculated for these sites are provided in Table 5-1.

Table 5-1. Source-term depletion rates during residual period. 

Facility Depletion rate (per day)

Blockson 0.00076

Dow Madison 0.00027

General Atomics 0.00116

Simonds Saw and Steel 0.00049

Average 0.00067
Document No. ORAUT-TKBS-0043 Revision No. 02 Effective Date: 09/16/2011 Page 33 of 43

Because these values were empirically derived at operating facilities, NIOSH considers them more 

appropriate than a value based solely on theoretical considerations. As such, the average value of 

0.067% per day, as calculated using the data in Table 5-1, will be used to determine source term 

depletion effects. The intakes in Table 5-2 and Table 5-3 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. The internal dose analysis should include the potential 

inadvertent ingestion of uranium activity and be based on guidance in NIOSH (2004). 

Table 5-2. Uranium (and RU components) annual inhalation rates for the residual

period.

Year

Inhalation (pCi/day)

U234 Pu238 Np237 Tc-99 Th232 Th228 Ru-106

1971 1.71E+03 1.03E+02 8.55E+00 6.84E+02 3.42E+01 3.42E+01 6.84E+01

1972 1.34E+03 8.03E+01 6.70E+00 5.36E+02 2.68E+01 2.68E+01 5.36E+01

1973 1.05E+03 6.29E+01 5.24E+00 4.19E+02 2.10E+01 2.10E+01 4.19E+01

1974 8.21E+02 4.93E+01 4.11E+00 3.28E+02 1.64E+01 1.64E+01 3.28E+01

1975 6.43E+02 3.86E+01 3.21E+00 2.57E+02 1.29E+01 1.29E+01 2.57E+01

1976 5.03E+02 3.02E+01 2.52E+00 2.01E+02 1.01E+01 1.01E+01 2.01E+01

1977 3.94E+02 2.37E+01 1.97E+00 1.58E+02 7.88E+00 7.88E+00 1.58E+01

1978 3.09E+02 1.85E+01 1.54E+00 1.23E+02 6.17E+00 6.17E+00 1.23E+01

1979 2.42E+02 1.45E+01 1.21E+00 9.67E+01 4.83E+00 4.83E+00 9.67E+00

1980 1.89E+02 1.14E+01 9.46E-01 7.57E+01 3.79E+00 3.79E+00 7.57E+00

1981 1.48E+02 8.89E+00 7.41E-01 5.93E+01 2.96E+00 2.96E+00 5.93E+00

1982 1.16E+02 6.96E+00 5.80E-01 4.64E+01 2.32E+00 2.32E+00 4.64E+00

1983 9.09E+01 5.45E+00 4.54E-01 3.64E+01 1.82E+00 1.82E+00 3.64E+00

1984 7.12E+01 4.27E+00 3.56E-01 2.85E+01 1.42E+00 1.42E+00 2.85E+00

1985 5.57E+01 3.34E+00 2.79E-01 2.23E+01 1.11E+00 1.11E+00 2.23E+00

1986 4.36E+01 2.62E+00 2.18E-01 1.75E+01 8.73E-01 8.73E-01 1.75E+00

1987 3.42E+01 2.05E+00 1.71E-01 1.37E+01 6.83E-01 6.83E-01 1.37E+00

1988 2.68E+01 1.61E+00 1.34E-01 1.07E+01 5.35E-01 5.35E-01 1.07E+00

1989 2.10E+01 1.26E+00 1.05E-01 8.38E+00 4.19E-01 4.19E-01 8.38E-01

1990 1.64E+01 9.85E-01 8.20E-02 6.56E+00 3.28E-01 3.28E-01 6.56E-01

1991 1.28E+01 7.71E-01 6.42E-02 5.14E+00 2.57E-01 2.57E-01 5.14E-01

1992 1.01E+01 6.04E-01 5.03E-02 4.02E+00 2.01E-01 2.01E-01 4.02E-01

1993 7.88E+00 4.73E-01 3.94E-02 3.15E+00 1.58E-01 1.58E-01 3.15E-01

1994 6.17E+00 3.70E-01 3.08E-02 2.47E+00 1.23E-01 1.23E-01 2.47E-01

1995 4.83E+00 2.90E-01 2.42E-02 1.93E+00 9.66E-02 9.66E-02 1.93E-01

1996 3.78E+00 2.27E-01 1.89E-02 1.51E+00 7.57E-02 7.57E-02 1.51E-01

1997 2.96E+00 1.78E-01 1.48E-02 1.18E+00 5.92E-02 5.92E-02 1.18E-01

1998 2.32E+00 1.39E-01 1.16E-02 9.28E-01 4.64E-02 4.64E-02 9.28E-02

1999 1.82E+00 1.09E-01 9.08E-03 7.27E-01 3.63E-02 3.63E-02 7.27E-02

2000 1.42E+00 8.53E-02 7.11E-03 5.69E-01 2.84E-02 2.84E-02 5.69E-02

2001 on 1.11E+00 6.68E-02 5.57E-03 4.46E-01 2.23E-02 2.23E-02 4.46E-02
Document No. ORAUT-TKBS-0043 Revision No. 02 Effective Date: 09/16/2011 Page 34 of 43

Table 5-3. Uranium (and RU components) annual ingestion rates for the residual 

period. 

Year

Ingestion (pCi/day)

U234 Pu238 Np237 Tc-99 Th232 Th228 Ru-106

1971 3.56E+01 2.14E+00 1.78E-01 1.42E+01 7.12E-01 7.12E-01 1.42E+00

1972 2.79E+01 1.67E+00 1.39E-01 1.12E+01 5.58E-01 5.58E-01 1.12E+00

1973 2.18E+01 1.31E+00 1.09E-01 8.73E+00 4.37E-01 4.37E-01 8.73E-01

1974 1.71E+01 1.03E+00 8.55E-02 6.84E+00 3.42E-01 3.42E-01 6.84E-01

1975 1.34E+01 8.03E-01 6.69E-02 5.35E+00 2.68E-01 2.68E-01 5.35E-01

1976 1.05E+01 6.29E-01 5.24E-02 4.19E+00 2.10E-01 2.10E-01 4.19E-01

1977 8.21E+00 4.92E-01 4.10E-02 3.28E+00 1.64E-01 1.64E-01 3.28E-01

1978 6.43E+00 3.86E-01 3.21E-02 2.57E+00 1.29E-01 1.29E-01 2.57E-01

1979 5.03E+00 3.02E-01 2.52E-02 2.01E+00 1.01E-01 1.01E-01 2.01E-01

1980 3.94E+00 2.36E-01 1.97E-02 1.58E+00 7.88E-02 7.88E-02 1.58E-01

1981 3.09E+00 1.85E-01 1.54E-02 1.23E+00 6.17E-02 6.17E-02 1.23E-01

1982 2.42E+00 1.45E-01 1.21E-02 9.67E-01 4.83E-02 4.83E-02 9.67E-02

1983 1.89E+00 1.14E-01 9.46E-03 7.57E-01 3.78E-02 3.78E-02 7.57E-02

1984 1.48E+00 8.89E-02 7.41E-03 5.93E-01 2.96E-02 2.96E-02 5.93E-02

1985 1.16E+00 6.96E-02 5.80E-03 4.64E-01 2.32E-02 2.32E-02 4.64E-02

1986 9.09E-01 5.45E-02 4.54E-03 3.63E-01 1.82E-02 1.82E-02 3.63E-02

1987 7.11E-01 4.27E-02 3.56E-03 2.85E-01 1.42E-02 1.42E-02 2.85E-02

1988 5.57E-01 3.34E-02 2.79E-03 2.23E-01 1.11E-02 1.11E-02 2.23E-02

1989 4.36E-01 2.62E-02 2.18E-03 1.74E-01 8.72E-03 8.72E-03 1.74E-02

1990 3.42E-01 2.05E-02 1.71E-03 1.37E-01 6.83E-03 6.83E-03 1.37E-02

1991 2.67E-01 1.60E-02 1.34E-03 1.07E-01 5.35E-03 5.35E-03 1.07E-02

1992 2.09E-01 1.26E-02 1.05E-03 8.38E-02 4.19E-03 4.19E-03 8.38E-03

1993 1.64E-01 9.84E-03 8.20E-04 6.56E-02 3.28E-03 3.28E-03 6.56E-03

1994 1.28E-01 7.71E-03 6.42E-04 5.14E-02 2.57E-03 2.57E-03 5.14E-03

1995 1.01E-01 6.03E-03 5.03E-04 4.02E-02 2.01E-03 2.01E-03 4.02E-03

1996 7.88E-02 4.73E-03 3.94E-04 3.15E-02 1.58E-03 1.58E-03 3.15E-03

1997 6.17E-02 3.70E-03 3.08E-04 2.47E-02 1.23E-03 1.23E-03 2.47E-03

1998 4.83E-02 2.90E-03 2.41E-04 1.93E-02 9.66E-04 9.66E-04 1.93E-03

1999 3.78E-02 2.27E-03 1.89E-04 1.51E-02 7.56E-04 7.56E-04 1.51E-03

2000 2.96E-02 1.78E-03 1.48E-04 1.18E-02 5.92E-04 5.92E-04 1.18E-03 

2001 on 2.32E-02 1.39E-03 1.16E-04 9.27E-03 4.64E-04 4.64E-04 9.27E-04

The daily ingestion rate in activity per day is estimated as 0.2 times the average daily air 

concentration expressed in units of activity/m3 (NIOSH 2004). The annual ingestion rates for the 

residual period are shown in Table 5-3.

Remediation of the Burial Grounds

Remediation activities of the burial grounds should be assessed based on individual bioassay records

per guidance in Section 3. 
Document No. ORAUT-TKBS-0043 Revision No. 02 Effective Date: 09/16/2011 Page 35 of 43

5.2 ESTIMATION OF EXTERNAL EXPOSURE FROM RESIDUAL RADIOACTIVITY 

The default exposure rates below should be assigned, during the applicable years. External 

dosimetry data might exist during the residual period; however, operations were not part of the 

nuclear weapons-related program, and are not covered for dose reconstruction under EEOICPA 

(Leiton 2011). However, these monitoring data can be used to limit dose, as appropriate (NIOSH 

2008a). 

In the absence of claim-specific information, the following assumptions in Table 5-4 should be made 

about external exposures [11]:

Table 5-4. External default assumptions during the residual period. 

Residual period Sources of external exposure Default assumptions

1971–Present

(Applicable to all 

workers pre-1991 and 

all workers 1991 on 

with no monitoring 

data) 

Photons, 30–250 keV Ambient H * (10) to organ dose (HT)

Electrons, > 15 keV or <30 keV Follow ORAUT (2005d) requirements

1991–2009 

(Applicable to Burial 

Grounds Remediation 

Workers with 

monitoring data)

Photons, 30–250 keV Deep Dose Equivalent (Hp(10)) to Organ Dose (HT)

Electrons, > 15 keV or <30 keV Follow ORAUT (2005d) requirements

The 95th-percentile value of the maximum average breathing-zone air concentration for the 1961 AEC 

air sample results (from Table 3-8) was calculated using Crystal Ball and assuming a lognormal 

distribution calculation yielded a distribution with a GM of 167 dpm/m3 and a GSD of 2.12. This 

resulted in a 95th-percentile value of 578 dpm/m3

. 

It is assumed that the material was deposited on the floor with a deposition velocity of 0.00075 m/s for 

a period of 1 year without cleanup. This would result in a contamination level of 1.37 × 107 dpm/m2

. 

External dose rates were calculated for exposure to contaminated surfaces and submersion from 

resuspended materials based on the DCFs in Federal Guidance Report No. 12 (EPA 1993).

The external doses are from penetrating photons with energies between 30 keV and 250 keV and 

>15 keV for penetrating exposures. NIOSH has previously determined the activity removal rates 

during the residual radiation periods for the four sites as summarized in Table 5-1 (DCAS-TKBS-0002

(DCAS 2010b); Battelle-TBD-6000 Appendix C (NIOSH 2008b); ORAUT-TKBS-0045 (ORAUT 

2008b); ORAUT-TKBS-0032 (ORAUT 2011a)). In these cases, contemporary estimates of airborne 

radioactivity at the beginning and end of each site’s residual radiation period were used to estimate 

the effective exponential clearance of the contamination over an extended period of time for each site 

(ORAUT-OTIB-0070 (ORAUT 2008a)). The source-term depletion rates calculated for these sites are 

provided in Table 5-1. Because these values were empirically derived at operating facilities, NIOSH 

considers them more appropriate than a value based solely on theoretical considerations. As such, 

the average value of 0.067% per day, as calculated using the data in Table 5-1, will be used to 

determine source term depletion effects.

Table 5-5 provides the external exposure rates for the residual period using the average source term 

depletion value.
Document No. ORAUT-TKBS-0043 Revision No. 02 Effective Date: 09/16/2011 Page 36 of 43

Table 5-5. External annual 

 exposure rates for the residual

period. 

Year

External

(rem/year)

Pen Non-Pen

1971 0.007 1.444

1972 0.005 1.131

1973 0.004 0.885

1974 0.003 0.693

1975 0.003 0.543

1976 0.002 0.425

1977 0.002 0.333

1978 0.001 0.261

1979 0.001 0.204

1980 0.001 0.160

1981 0.001 0.125

1982 0.001* 0.098

1983 0.001* 0.077

1984 0.001* 0.060

1985 0.001* 0.047

1986 0.001* 0.037

1987 0.001* 0.029

1988 0.001* 0.023

1989 0.001* 0.018

1990 0.001* 0.014

1991 0.001* 0.011

1992 0.001* 0.008

1993 0.001* 0.007

1994 0.001* 0.005

1995 0.001* 0.004

1996 0.001* 0.003

1997 0.001* 0.003

1998 0.001* 0.002

1999 0.001* 0.002

2000 0.001* 0.001

2001 -

2009 0.001* 0.001

*As a favorable to claimant assumption,

 values were rounded up to 0.001 rem per

year. 

Remediation of the Burial Grounds

Remediation activities of the burial grounds should be assessed based on individual monitoring 

records per guidance in Section 4 and Table 5-4. 
Document No. ORAUT-TKBS-0043 Revision No. 02 Effective Date: 09/16/2011 Page 37 of 43

6.0 ENVIRONMENTAL DOSE 

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. 

7.0 ATTRIBUTIONS AND ANNOTATIONS

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.

[1] Guido, Joseph. ORAU Team. Senior Health Physicist. 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, the dose reconstructor should 

assume the MDL is 10 dpm/L.

[3] Sharfi, Mutty M. ORAU Team. Principal AWE Dosimetrist. June 2011. 

An assumption of 10-year-old fuel-grade plutonium is considered reasonable and should be 

used in the absence of claimant-specific information.

[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 an 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 that 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. 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, M., and Winslow, R.. ORAU Team. Senior Health Physicists. June 2007. 

In the absence of further information, it is appropriate to apply these DCFs. 
Document No. ORAUT-TKBS-0043 Revision No. 02 Effective Date: 09/16/2011 Page 38 of 43

[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] Fix, J. J., Smith, M., and Winslow, R. ORAU Team. Senior Health Physicists. March 2008. 

Default DCFs from NIOSH (2007a) were identified for photons and electrons for the 

operational and residual periods. 
Document No. ORAUT-TKBS-0043 Revision No. 02 Effective Date: 09/16/2011 Page 39 of 43

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