DOE Review 04/05/2011

Division of Compensation Analysis and Support

Technical Basis Document for the DuPont Deepwater Works

Deepwater, New Jersey

Document Number:

DCAS-TKBS-0006

Effective Date: 12/13/2013

Revision No.: 01

Page 1 of 17

Subject Expert: David Allen

Approval: Signature on file Date: 12/13/2013

Supersedes:

DCAS-TKBS-0006

Rev. 00 PC-1

James W. Neton, Associate Director for Science

TABLE OF CONTENTS

Section Page

1.0 Introduction ............................................................................................................................4

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

3.0 Process Description ................................................................................................................6

4.0 Internal Dose...........................................................................................................................6

5.0 External Dose .........................................................................................................................7

6.0 Residual Contamination .......................................................................................................13

7.0 Occupational Medical Dose..................................................................................................15

8.0 References ............................................................................................................................16

LIST OF TABLES

Table Page

Table 1: Daily Intakes of Uranium ........................................................................................................7

Table 2: External Dose Conversion Factor for Air Submersion ............................................................8

Table 3: External Dose Conversion Factor for Surface Contamination.................................................8

Table 4: Uranium dose rates from drums of yellowcake .......................................................................9

Table 5: Beta Surface Exposure Rates from Uranium Metal and Compounds......................................9

Table 6: Dose rates from drums of uranium compounds.....................................................................10

Table 7: Annual Doses at Deepwater Works.......................................................................................12

Table 8: Annual Whole Body External Dose from Residual Contamination ......................................14

Table 9: Annual Shallow External Dose from Residual Contamination .............................................14

Table 10: Daily Intakes from Residual Contamination (Solubility Type S or M)...............................15
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LIST OF FIGURES

Figure Page

Figure 1: Beta dose rate from yellowcake separated from ore.............................................................11

Figure 2: Approximation of beta dose rates in Figure 1 ......................................................................11
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ISSUE

AUTHORIZATION

DATE

EFFECTIVE

DATE REV. NO. DESCRIPTION

02/14/2011 02/15/2011 00 Changes Battelle-TBD-6001 Appendix to 

a standalone document. Revises external 

dose model to eliminate dependence on 

Battelle-TBD-6001. Provides more 

detailed description of other dose models. 

Incorporate review comments.

03/08/2011 03/08/2011 00 PC-1 Corrected values in Table 1. Corrected 

error in table of contents.

12/13/2013 12/13/2013 01 Revised ingestion and external dose 

estimates to incorporate comments from 

the Advisory Board Working Group.
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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 for which 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 received 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 2007). 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 nonweapons-related work, if applicable, will be covered elsewhere.
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The following information from the Department of Energy’s Office of Health, Safety and 

Security EEOICPA Find Facilities webpage defines the EEOICPA covered periods for DuPont 

Deepwater.

Site: DuPont Deepwater Works

Location: Deepwater, New Jersey

Covered Period: AWE 1942-1949, Residual Radiation 1950-1995; 1997-March 2011; DOE 

1996 (remediation)

This document contains a summary of the description of the site as well as the Atomic Energy 

Commission activities performed there, and provides the technical basis to be used to evaluate 

the occupational radiation doses for EEOICPA claims.

2.0 Site Description and Operational History

DuPont Deepwater Works was a DuPont facility located in Deepwater NJ. The name of the 

facility was officially changed from “Dye Works” to “Chambers Works” on 4/7/1944 (Chambers 

Works 1945). DuPont Deepwater Works conducted work on several projects for the MED. 

Several involved producing non-radioactive chemicals (Chambers Works 1945). These include 

project number 9595 (under Letter Contract W-7412 Eng. 2), project number 9757 (under Letter 

Contract W-7412 Eng. 6), and project number 9233 (under Letter Contract W-7412 Eng. 8). 

Project number 9634 was conducted under Letter Contract W-7412 Eng. 3. The letter contract 

was dated 11/20/1942 and the project was approved by DuPont’s Executive Committee on 

12/23/1942. Construction was completed in three stages which started production on 2/13/1943, 

4/28/1943 and 6/5/1943. The scope of work under this contract included converting U3O8 to 

UO2, converting UO2 to UF4, and converting UF4 to uranium metal.

Letter Contract W-7412 Eng 3 indicated the U3O8 would be supplied by the Government. 

However, on 12/30/1942, Letter Contract W-7412 Eng. 22 was issued to direct DuPont to build a 

facility to produce the U3O8 from various types of uranium scrap. This became Project number 

9803 and was approved by DuPont’s Executive committee on 3/31/1943. The 100 Section of the 

plant was operational on 8/16/1943 and the 200 Section of this plant was operational on 

10/1/1943 (Chambers Works 1945). 

The original research work was conducted at the Jefferson Lab in Building J-16. This building 

was demolished and several feet of earth removed sometime between 1943 and 1945. Building 

J-26 was eventually built at that location (DOE 1978). 

The other two projects were located in buildings 708 and 845. A portion of building 708 was 

demolished in 1945. The rest of the building along with several feet of earth was removed in 

1953. Radiological surveys of building 845 were conducted in 1977 and 1983 (DOE 1978, 

Bechtel 1983).
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3.0 Process Description 

Operations involving uranium at the DuPont Deepwater Works began early in 1942 when 

DuPont was conducting experiments with uranium hexafluoride (UF6) under contract to the 

Office of Scientific Research and Development (OSRD). The method employed utilizing 

natural uranium oxide and converted it to uranium tetrafluoride (UF4) and then to UF6. When 

the MED was chartered, it took over the OSRD contracts. DuPont operations for MED included 

conversion of black oxide (U3O8) and sodium diuranate to orange oxide (UO3) and then to brown 

oxide (UO2), production of uranium tetrafluoride (UF4) from uranium oxide (UO2 and UO3), 

production of uranium peroxide (UO42H2O) from scrap uranium for subsequent production of 

UO2, production of UF6 from UF4, production of uranium metal using the magnesium process 

and various related research activities. DuPont continued its research activities for AEC until 

late 1947 (Chambers Works 1945). No documentation was found indicating there were other 

sources of radiation at Deepwater Works. 

4.0 Internal Dose

Air samples were collected at the Deepwater Works plant on various occasions at a variety of 

locations between 4/3/1944 and 6/7/1945 (DuPont Dust Reports). A total of 252 air samples 

were collected. These air samples were analyzed by assuming they fit a lognormal distribution. 

The geometric mean of that distribution was 181 dpm/m3 with a geometric standard deviation of 

5.73. These air samples included primarily operational areas but also included some general 

areas of the facility as well as operational areas while equipment was shutdown. The distribution 

would therefore not necessarily be representative of operational personnel. Therefore, exposure 

estimates will rely on three categories of workers. People routinely working with uranium 

(Operators) will be given the 95th percentile of the air concentration distribution. People 

working in the vicinity but not normally operating equipment (Supervisors) will be given the 

50th percentile of the distribution. People not routinely in the vicinity of the uranium (Clerical) 

will be given the 5th percentile of the distribution. These values were used to determine an 

ingestion intake per OCAS-TIB-0009 (OCAS 2004) and are summarized in the table below. 

At the end of the weapons related work at DuPont Deepwater Works, the buildings were 

decontaminated and turned over to DuPont. The last building was turned over in 1949; however 

the decontamination was performed in 1948. The last building was surveyed after 

decontamination on 12/30/1948 (DuPont 19449). Therefore, this estimate will cover the time 

period of 1942 through 1948. Intakes associated with 1949 will be the same as those for the 

residual contamination period addressed in Section 6.
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Table 1: Daily Intakes of Uranium

Category Years Description

Inhalation 

(dpm/day)

Ingestion 

(dpm/day)

Operators 1942-1948 Routinely working 

with uranium 25245 438

Supervisors/Laborers 1942-1948 Routinely in the 

area

1428 25

Clerical 1942-1948 Not routinely in 

the area 81 1.4

Dose calculated from these intakes is entered into IREP as alpha radiation with a “constant” 

distribution. 

5.0 External Dose

No External dosimetry results were found for the DuPont Deepwater plant. Therefore, the 

external dose to workers at the plant was modeled. Radiation can be emitted not only from 

uranium but from its short lived decay products. Since the Deepwater Works plant did not 

process any uranium ores, all the uranium present at the plant had been processed previously and 

any decay products removed. While decay products will be produced immediately after the 

processing, the long-lived decay products can take hundreds or thousands of years to reach an 

appreciable level. However, short –lived decay products can reach a value near equilibrium in a 

much shorter period of time. Anderson and Hertel (Anderson and Hertel 2005) showed that the 

short lived nuclides (Th234, Pa234m, Pa234and Th231) are very close to equilibrium (adjusted for 

branching ratios) at 100 days following separation. Therefore, for modeling external dose, 

uranium isotopes (U238, U235, and U234) were assumed to exist in their natural ratios and their 

short-lived decay products (Th234, Pa234m, Pa234and Th231) were assumed to have reached 

equilibrium. 

External sources of radiation at a uranium facility could include beta and photon radiation from 

the material being produced, from exposure to contaminated surfaces and from submersion in air 

contaminated with uranium dust. 

When workers are enveloped in a cloud of radioactive dust, they will receive a small amount of 

external dose. External exposure rates from uranium and its radioactive decay products are 

shown in Table 2. The doses were calculated using the computer code MicroShield version 6.02

(Grove Engineering 2003). The calculated dose rates are for natural uranium and include the 

dose contribution from the radioactive decay products of U

238, U235, and U234

. Radioactive decay 

product ingrowth of 100 days was assumed for these calculations. Air concentration from 

DuPont Deepwater Works air samples was used to determine an external dose rate from this 

route of exposure. This calculation resulted in a dose rate of 4.45x10-7 mR/hr. This is 

negligible when compared to other sources of external radiation.
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Table 2: External Dose Conversion Factor for Air Submersion

External Dose Conversion factor

Time since separation (mR/h per dpm(α)/m3)

100 d 2.46E-09

When workers are working on a contaminated surface, they will receive a small amount of 

external dose. External dose rates from uranium and its radioactive progeny are shown in Table 

3. The doses were calculated using the computer code MicroShield version 6.02 (Grove 

Engineering 2003). The calculated dose rates are for natural uranium and include the dose 

contribution from the radioactive decay products of U238, U235, and U234. 

The quantity of uranium on the floor surface was estimated from measured air concentrations. 

The level of surface contamination was determined by first calculating a terminal settling 

velocity for 5-μm activity mean aerodynamic diameter (AMAD) particles. The calculated 

terminal settling velocity was 0.00075 meters per second. It was assumed that the surface 

contamination level was due to 30 days of constant deposition from the constant air 

concentration with no removal. Using this surface concentration and the conversion factors in 

Table 3, the external dose rates from contaminated surfaces were calculated. This results in a 

calculated value of 0.000139 mR/hr photon and 0.0134 mrad/hr beta. These values are small 

when compared to the dose rate directly from uranium. Combined with the large uncertainty

applied to the direct values, these values are considered negligible and not included in the 

estimate.

Table 3: External Dose Conversion Factor for Surface Contamination

Surface contamination dose conversion factor

Time since separation

Photon Exposure Rate

(mR/h per dpm(α)/m2)

Beta Dose Rate

(mrad/h per dpm(α)/m2)

100 d 3.94E-10 3.82E-8

Next, the external dose rate from direct handling of uranium compounds was considered. 

Several different chemical compounds of uranium were handled at the Deepwater plant. Also 

the material was handled in various-sized containers or equipment. However, the external dose 

rate from uranium is not very sensitive to these variations once a sufficient quantity has 

accumulated. Table 4 below shows the dose rates calculated for a drum of U3O8 using MCNPX 

(version 2.5.0). The density of the drum was increased allowing for a greater amount of uranium 

to be contained by the drum. It can be seen that once the density reaches a realistic value, the 

dose rate does not change appreciably. The same is true of an array of drums or large amounts of 

uranium contained in equipment. 
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Table 4: Uranium dose rates from drums of uranium oxide

Density of 

U3O8 (g cm-3)

Activity of U in 

drum (Ci)

Photon 

emission 

dose(rad/hr)

Bremsstrahlung 

dose (rad/hr)

Total dose rate 

at 30 cm 

(rad/hr)

0.5* 3.121E-02 3.96E-04 3.20E-4 7.16E-4

1 6.242E-02 5.00E-04 3.60E-04 8.60E-04

2 1.248E-01 5.54E-04 3.76E-04 9.30E-04

4 2.497E-01 5.84E-04 3.84E-04 9.69E-04

6 3.745E-01 5.84E-04 3.64E-04 9.48E-04

7 4.182E-01 5.81E-04 3.74E-4 9.56E-4

*The drum begins to noticeably impact the dose rates at low material concentration.

Also varying the chemical makeup of the uranium compound has little effect on the external 

dose rates. Table 5 shows the surface beta dose rates from various chemical forms of uranium. 

Uranium metal exceeds the dose rates from other uranium compounds. However, the dose rates 

from uranium oxides and UF4 are sufficiently similar in magnitude to the dose rates from 

uranium metal so that uranium metal dose rates can be assumed to be representative of the dose 

rates from all uranium compounds.

Table 5: Beta Surface Exposure Rates from Equilibrium Thickness of Uranium Metal and 

Compounds (DOE-STD-1136-2004)

Source

Beta Surface 

Exposure

Rate, mrad h-1

U-Nat metal slab 233

UO2 207

UF4 179

UO2(NO3)26H20 111

UO3 204

U3O8 203

UO2F2 176

Na2U2O7 167

a. Beta surface exposure rate in air through a 

polystyrene filter 7mg/cm2thick.

The geometry of the uranium can also have an effect on the external dose rate from the uranium. 

Table 6 shows the calculated dose rate from several sizes of drums. The dose rates from drums 

were calculated using MicroShield Version 5.01. The calculations assumed that the time of 

decay was 100 days, which allows the ingrowth of uranium decay products which will increase 

the dose rate. The contents of the drums were modeled as soil at a density of 1.6 (Eckerman and 

Ryman 1993). The calculations did not account for Bremsstrahlung that may have been 

generated by the interactions of beta particles with the contents of the drum. Calculations 

performed by others (Anderson and Hertel 2005) indicate that the dose rate due to 
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Bremsstrahlung may be equal to the photon dose rate. Therefore, the values shown in Table 6

are twice the dose rate that was calculated for photons alone. The one foot (30 cm) values for the 

55 gallon drum compare well to the values in Table 4. In Table 6, the photon dose was doubled 

to account for Bremsstrahlung radiation. In Table 4, the Bremsstrahlung radiation was 

accounted for separately. In comparing the two tables, it can be seen that the Bremsstrahlung 

treatment in Table 6 is favorable. It can also be seen that the same treatment would result in total 

dose values of approximately 1.2 mR/hr from Table 4 which compares well with the 1.3 mR/hr

value in Table 6. Values from Table 6 will be used for external photon dose calculations for the 

Deepwater plant. 

Table 6: Dose rates from drums of uranium compounds

Drum Size (gal)

Dose Rates (mR/h)

1 cm 10 cm 30 cm 100 cm

100 day decay

5 3.7 1.4 0.4 0.1

30 4.4 2.5 1.1 0.2

55 4.5 2.8 1.3 0.28

The exact external exposure scenarios at the Deepwater plant vary making it impossible to model 

each task. However, based on the fact that the external photon dose rates do not vary 

significantly with most parameters, the external exposure estimate assumes an operator spent 

100% of the time one meter from a 55 gallon drum of uranium. This is considered the median of 

a lognormal distribution and a geometric standard deviation of 5 is assigned to account for the 

variability of distance.

While the photon dose rates do not vary significantly the beta dose rates do vary a great deal with 

distance due to air attenuation. The beta dose rates are estimated using the beta dose rates from 

aged yellowcake (U3O8) at various distances from the uranium presented in Figure 1. Using the 

same distance variation as the gamma dose rate (geometric standard deviation of 5), Figure 1 can 

be reasonably approximated by a lognormal distribution with a geometric mean of 1 mrem/hr. 

This distribution is show in Figure 2 superimposed over Figure 1.

The median distance of the gamma dose rate distribution was the dose rate at 1 meter. Since the 

distribution is assumed to be associated with the operators distance from the source, the same 

distribution will be used for the beta dose rate. The one meter beta dose rate of the lognormal 

distribution in Figure 2 is 1 mrem/hr. Therefore, as with the gamma dose estimate, the operator 

will be based on the beta dose distribution with a GSD of 5 and the GM of 1 mrem/hr.
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Figure 1: Beta dose rate from yellowcake separated from ore for more than 100 days as 

a function of distance from the surface. [Reproduced from US NRC 2002a]

Figure 2: Approximation of beta dose rates in Figure 1
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For extremity dose (dose to the hands and forearms), the beta dose distribution would 

underestimate the dose since operators could be handling the material directly at times. To 

account for this, it is assumed the operator could be touching the material half the time he is 

within one foot of the material. The dose rate for used for the estimate will be the uranium metal 

contact dose rate from Table 5. The amount of time spent within 1 foot is determined from the 

gamma dose rate distribution. With a geometric mean of 0.28 mr/hr and a geometric standard 

deviation of 5, the dose rate of 1.3 mr/hr will be exceeded 17% of the time. This is the gamma 

dose rate at 1 foot (30 cm) shown to Table 6. Therefore, the estimate assumes the operators 

hands are in contact with the material 8.5% of the time and at 1 foot another 8.5% of the time. 

The remaining time will be assumed to be at 1 meter. Unlike the other estimates, this estimate is 

considered bounding and will therefore not be represented by a distribution. Instead, this 

estimate will be considered a constant.

For exposure estimates, each claim will be evaluated to determine the most appropriate job 

category to utilize for the external dose estimate. The “operator” job category consists of

personnel that were routinely and directly involved in operations with uranium. The “laborer” 

job category consists of personnel that supported these operations. The “Supervisor” job 

category consists of personnel that were in the operations area but were not routinely involved in 

hands on activities with uranium. The “Other” job category consists of personnel that did not 

routinely enter the testing area. 

External exposure scenarios for operators have already been described. The typical work year is 

assumed to be 2400 hours during the operational period. Laborers will be assumed to receive 

half the external dose that operators receive. Supervisors are assumed to receive half the dose 

that laborers receive. Others are assumed to receive one tenth the external radiation that 

supervisors receive. External doses to the different job categories during operational years are

listed in Table 7. Extremity doses should be entered into IREP as a constant distribution. Other 

doses should be entered into IREP as a lognormal distribution with a GSD of 5.0. Photon doses 

should be assigned as 50% 30 to 250 keV photons and the remaining 50% as greater than 250 

keV photons. Skin and Hands and forearm dose should be assigned as 100% greater than 15 keV 

electrons.

Table 7: Annual Doses at Deepwater Works

Years Photon

(mR/yr)

a

Skin

(mrad/yr) a

Hands and forearms

(mrad/yr) b

Operators 1942-1948 672 2400 50000

Laborers 1942-1948 336 1200 25000

Supervisors 1942-1948 168 600 12500

Other 1942-1948 17 60 1250

a GM of a lognormal distribution with a GSD of 5.

b Constant distribution
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6.0 Residual Contamination

Even though the last building was released to DuPont in 1949, the last decontamination survey 

(described below) was completed in late 1948 (DOE 1978). After 1948, there was still the 

potential for dose from residual contamination remaining in the buildings. The dose estimate 

from this residual contamination is described below. 

The initial research conducted in 1942 at Deepwater Works was conducted in building J-16. 

This building was demolished and several feet of earth removed sometime between 1943 and 

1945 (DOE 1978). 

Building 708 was partially demolished in 1945. This building was eventually shutdown, 

decontaminated and released to DuPont in 1949. The final survey of the building was conducted 

on 12/30/1948. This survey indicated the northwest tile wall was the most contaminated location 

in the building. The survey measured beta and gamma dose rates along the wall at several 

distances from the center of the contaminated area including the center of the contaminated area. 

The measurements were taken at these locations on contact with the wall as well as one foot 

away, two feet away, four feet away, six feet away and twenty feet away from the wall. These 

measurements were conducted on all five floors of the building with the exception of the second 

and third floors. That section of the wall was not accessible from those floors (DuPont 1949).

The survey also recorded measurements on the floor on all five floors of the building. All of 

these measurements indicated direct alpha results of less than 500 dpm/100cm2and the highest 

beta gamma dose rate three feet above the floor as 0.05 mrep/hr. It should be noted that mrep is 

an outdated unit of measure equivalent to the more modern mrad. Building 708 was released to 

DuPont and in 1953 the building was demolished and several feet of soil were removed. 

Building 845 was released to DuPont on 11/15/1948 after decontamination (DuPont 1948c). The 

decontamination effort consisted of removing all apparatus, contaminated ducts, pipes, tanks, 

concrete bases, and wood floors as well as sandblasting the concrete floors. The whole interior 

was washed with water under pressure. A survey conducted on 10/6/1948 indicated all direct 

alpha measurements were less than 500 dpm/100cm2. The survey also indicated beta and gamma 

radiation levels three feet above the floor were less than 0.03 mrep/hr (DuPont 948b).

The building remained standing and was again surveyed in 1977 (DOE 1978) and 1983

(Bechtel 1983). The 1983 survey provided only a range of values with no indication of the 

average or typical contamination levels. The 1977 survey however, indicated average 

measurements as well as maximum measurements. This survey also indicated areas of 

maximum measurements were typically small areas. 

The survey indicated beta gamma direct contact dose rates typically around 0.1 mrad/hr on most 

floors, walls, and ceilings. These readings were not corrected for background radiation so they 

are slightly high. Every floor of the plant had higher dose rates in small areas. This estimate 

will assume a dose rate of 0.2 mrad/hr as a favorable average dose rate. Since this is a contact 

beta plus gamma dose rate, corrections must be made to determine a whole body gamma and 

whole body beta dose rate. 
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As part of a test to determine the effectiveness of sandblasting, a survey was conducted in 

building 845 on 8/30/1948 and 8/31/1948 (DuPont 1948a). Part of this survey measured the 

open window dose rates (beta plus gamma) from three spots on the floor. Measurements were 

taken on contact with the floor and at waist height (three feet above the floor) prior to 

decontamination. The contact readings were 2.6, 2.6, and 15 mrep/hr. The corresponding three 

foot readings were 0.5, 0.5, and 1.0 mrep/hr. These measurements imply the readings taken 

three feet above the floor were 5.2, 5.2 and 15 times lower than the contact readings. Based on 

this, this estimate will use a value of five to estimate the whole body beta plus gamma dose rates 

from the contact dose rates. Therefore, the average whole body beta plus gamma dose rate is 

assumed to be 0.04 mrad/hr. 

The ratio of beta to gamma from uranium contamination can vary from one to one to as high as 

100 to one depending on the geometry, the amount of self shielding, and a number of other 

factors. For the purposes of this estimate, a ratio of 10 beta to 1 gamma will be used. Therefore, 

this appendix will assume the 0.04 mrad/hr is composed of a whole body gamma dose rate of 

0.0036 mR/hr and a beta whole body dose rate of 0.0364 mrad/hr. Skin of the extremities (hands 

and forearms) will be assumed to be exposed to the contact dose rate of 0.2 mrad per hour. 

While work schedules and locations can vary throughout a large facility such as Deepwater 

Works, a favorable assumption would be that individuals were exposed to these levels for 2000 

hours per year. While it is recognized that some individuals would have worked overtime, it is 

also very likely they worked in other areas of the site since building 845 was primarily used for 

storage in later years. This also makes the use of 2000 hours per year a bounding estimate. As 

such, this portion of the dose estimate will be considered a constant value with no statistical 

distribution. 

With these assumptions in place, this dose estimate will assume for the residual period, an annual 

whole body dose of 80 mrem. This is divided into 7.3 mrem deep dose and 72.7 mrem shallow 

dose (beta). Extremities will be estimated with an annual dose of 400 mrem assumed to be beta 

plus 7.3 mrem deep dose.

Table 8: Annual Whole Body External Dose from Residual Contamination

Operation Phase Years Whole Body

(mrem/year)(a)

Job Category GSD

Residual 1949-Mar. 2011 7.3 All Constant

Applied as Photons 30-250 keV. Whole body photon doses are to be converted to organ doses using the Exposure to Organ Dose 

Conversion Factors (US DHHS 2007).

Table 9: Annual Shallow External Dose from Residual Contamination

Operation Phase Years Shallow Dose

(mrem/year)(a)

Extremity 

Dose 

(mrem/year)(a)

Job 

Category

GSD

Residual 1949-Mar. 2011 72.7 400 All Constant

Applied as Electrons > 15 keV. 
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Direct alpha contamination measurements of the floor were conducted after decontamination of 

buildings 708 and 845. These measurements indicated the values were less than 500 

dpm/100cm2. The 1977 survey of building 845 confirmed this was still the case even though 

isolated spots of higher contamination were also found. The isolated spots were primarily less 

than 500 cm2in area but six spots on the first floor ranged up to 5000 cm2. These readings 

indicate fixed contamination that cannot be easily resuspended into the air. The fact that the 

1948 survey results are near the same values of the 1977 survey results indicates that the 

contamination is generally not being resuspended or removed in any other fashion. However, in 

order to account for small amounts of uranium that may have become airborne, a resuspension 

factor was applied to the 500 dpm/100cm2value. This resulted in an estimated airborne 

concentration of 0.05 dpm/m3. Again, it will be assumed that an individual was exposed to this 

level of airborne contamination for 2000 hours per year. This estimate will therefore be 

considered a bounding estimate and no statistical distribution will be associated with it. 

As with the airborne estimate, the fixed contamination will be assumed to be transferable to the 

point of causing inadvertent ingestion intakes. Again, the 500 dpm/100cm2value will be used 

and combined with a value of 1.1E-4 m2/hr (NUREG/CR 5512) to estimate an ingestion rate. 

This results in an estimated ingestion rate of 30.1 dpm/day. 

For internal dose estimates, the uranium will be considered to be type M or type S solubility. 

The dose estimate should be based on the one that produces the highest dose. 

Table 10: Daily Intakes from Residual Contamination (Solubility Type S or M)

Operation 

Phase

Years Radionuclide Inhalation

(dpm/day)

Ingestion

(dpm/day)

Job 

Category

GSD

Residual 1949-Mar. 

2011

U-234 0.329 30.1 All Constant

Note: intakes rates are normalize to calendar days

7.0 Occupational Medical Dose

No documentation regarding occupational medical dose specific to DuPont Deepwater Works

was found. Information to be used in dose reconstructions, for which no specific information is 

available, is provided in ORAUT-OTIB-0006, Technical Information Bulletin: Dose 

Reconstruction from Occupationally Related Diagnostic X-Ray Procedures (ORAUT 2005c). 

The assumed frequency in this document is PA chest X-ray for pre-employment, annual, and 

termination examinations between the years 1942 and 1949 (the covered period). Annual organ 

doses are entered into the NIOSH-IREP program as the annual dose due to an acute exposure to 

photons (E=30-250 keV). The distribution is assumed to be normal with a standard deviation of 

30%. 
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8.0 References

Anderson and Hertel 2005, Anderson, J.L., and Hertel, N.E., “Bremsstrahlung Doses from 

Natural Uranium Ingots,” Radiation Protection Dosimetry, 115 (1-4): 298-301, 2005

Bechtel National, Inc., 1983, Radiological Survey Report for the DuPont Chamber Works Plant 

Deepwater, New Jersey, Oak Ridge, Tennessee [ Ref ID 6391]. 

Chambers Works 1945, Report: History Chambers Works Special Construction, February 9, 

1945 [Ref ID 16272]. 

DOE 1978, Radiological Survey of the E.I DuPont DeNemours and Co., Deepwater, New Jersey, 

December 1978 [Ref ID 6395]. 

DOE, Office of Health, Safety and Security, EEOICPA web site.

http://www.dol.gov/owcp/energy/regs/compliance/EEOICPA_FacilityList.pdf

DuPont 1948a, Letter from R.E. Hayden to A.R. Piecot regarding Decontamination of Concrete 

Floor at DuPont, September 9, 1948 [Ref ID 6410 pg 8]. 

DuPont 1948b, Letter from R.E. Ryan to B. S. Wolf regarding Contamination of DuPont 

Building, November 15, 1948 [Ref ID 6410 pg 6].

DuPont 1948c, Memo from G. L. Ryan to B. S. Wolf regarding Contamination of DuPont 

Building, December 17, 1948 [Ref ID 6410 pg. 5].

DuPont 1949, Letter from Merril Eisenbud to E.I. DuPont de Nemours & Co. regarding 

Radiation Intensities in Building 708, January 18, 1949 [Ref ID 6410 pg 1].

DuPont Dust Reports. Various Air Sample Reports, 1944 through 1945. [Ref IDs 5882, 6420, 

6423, 6425, 6432, 6434, 6436, 6438, 6439, 8253, 8256, 17136].

Eckerman, KF and JC Ryman. 1993. External exposure to radionuclides in air, water, and soil. 

Federal Guidance Report No. 12. EPA-402-R-93-081. Environmental Protection Agency. 

Washington, DC.

Grove Engineering. 2003. MicroShield Software Program, Version 6.02. Copyright 1992-2003.

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

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

Compensation Program Act, Rev 1, OCAS-IG-003, National Institute for Occupational Safety 

and Health, Office of Compensation Analysis and Support, Cincinnati, Ohio.

NUREG/CR 5512 Vol. 1, Residual Radioactive Contamination From Decommissioning, 

Prepared for the Nuclear Regulatory Commission, October 1992
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OCAS (Office of Compensation Analysis and Support) 2004, Estimation of Ingestion Intakes, 

Rev 0, OCAS-TIB-0009, National Institute for Occupational Safety and Health, Office of 

Compensation Analysis and Support, Cincinnati, Ohio. 

ORAUT (Oak Ridge Associated Universities Team) 2005, ORAUT-OTIB-0006, Technical 

Information Bulletin: Dose Reconstruction from Occupationally Related Diagnostic X-Ray 

Procedures, Rev 03 PC-1, December 21, 2005.