ORAU TEAM

Dose Reconstruction

Project for NIOSH

 

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

Page 1 of 77

Document Title:

Nevada Test Site – Occupational 

Environmental Dose

Document Number: ORAUT-TKBS-0008-4 

Revision: 01

Effective Date: 05/27/2008

Type of Document: TBD

Supersedes: Revision 00 PC-1 

Subject Expert(s): Eugene M. Rollins

Site Expert(s): N/A

Approval: Signature on File Approval Date: 04/23/2008

Eugene M. Rollins, Document Owner

Approval: Signature on File Approval Date: 04/22/2008

John M. Byrne, Task 3 Manager

Concurrence: Signature on File Concurrence Date: 04/22/2008

Edward F. Maher, Task 5 Manager

Concurrence: Signature on File Concurrence Date: 05/14/2008

Kate Kimpan, Project Director

Approval: Signature on File Approval Date: 05/27/2008

James W. Neton, Associate Director for Science

 New Total Rewrite Revision Page Change

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

REVISION AND DISCARD / DESTROY ALL COPIES OF THE PRIOR REVISION.
Document No. ORAUT-TKBS-0008-4 Revision No. 01 Effective Date: 05/27/2008 Page 2 of 77

PUBLICATION RECORD

EFFECTIVE 

DATE

REVISION 

NUMBER DESCRIPTION

04/20/2004 00 New technical basis document for the Nevada Test Site– Occupational 

Environmental Dose. First approved issue. Initiated by Eugene M. 

Rollins.

12/08/2006 00 PC-1 Approved page change revision as a result of biennial review. Adds 

acronyms and abbreviations on pages 4 and 5. Updates required 

language on page 6 in the Introduction. Adds Purpose and Scope 

sections on page 7. Added instructions to dose reconstructors on 

pages 44 and 45 in Section 4.5. As a result of formal internal review, 

deletes Section 4.5.2 from page 45. Completes references on page 

47. This revision results in no change to the assigned dose and no 

PER is required. Training required: As determined by the Task 

Manager. Initiated by Eugene M. Rollins. Approval:

Eugene M. Rollins, Document Owner

Signature on File 12/05/2006

John M. Byrne, Task 3 Manager

Signature on File 10/24/2006

Edward F. Maher, Task 5 Manager

Signature on File 10/24/2006

Kate Kimpan, Project Director

Signature on File 11/15/2006

James W. Neton, Associate Director for Science

Brant A. Ulsh Signature on File for 12/08/2006

05/27/2008 01 Approved revision initiated to apply only to post 1962 employment 

periods and all covered employees including those identified by job 

classification as a drillback operator prior to 1965. Also applies to 

covered employees involved with any of the 10 underground tests that 

resulted in unexpected release of radioactive materials. Revised Table 

4.2.1.2.2.-2 (new Table 4-3) to reflect 2,000 hr/yr and 2400 m3/yr 

breathing rate. Deleted Tables 4.2.1.2.2-3 and 4.2.1.2.2-4. Deleted 

Sections 4.2.1.2.3 and 4.2.1.2.4. Added Sections 4.2.1.2.3 and 4.1.2.4 

to introduce revised method for assigning environmental ambient 

inhalation intakes. Replaced Section 4.2.2 with ambient ingestion 

intake section. Revised Section 4.4 to include dose from radon and 

other alpha and beta emitting radionuclides and to include discussion 

of radon exposure in Gravel Gerties. Added Section 4.4.3. Revised 

Section 4.4.4, Table 4.21 to increase radon WLM for G-Tunnel prior to 

1984 and to maximize radon exposures for unidentified work locations. 

Added discussion of radon in Gravel Gerties. Added Section 4.4.5 to 

provide method for assigning inhalation and ingestion intakes for 

underground workers. Added Section 4.5.2 to provide instructions to 

dose reconstructors for assigning inhalation and ingestion ambient 

environmental intakes for all workers at the NTS. Occupational 

environmental dose will increase for all claims not previously worked 

using ORAUT-OTIB-0018 or ORAUT-OTIB-0002. Also, cases of lung, 

ET1, and ET2 will need to be re-evaluated for increased radon 

exposures. Attachments A and B (formerly Rollins 2007, SRDB Ref 
Document No. ORAUT-TKBS-0008-4 Revision No. 01 Effective Date: 05/27/2008 Page 3 of 77

EFFECTIVE 

DATE

REVISION 

NUMBER DESCRIPTION

ID: 37805) were added to support the changes discussed in Section 

A.6.0 which included enriching the near-field environment with 

refractory elements. The 01-C version only added refractory elements 

back to the depleted refractory near-field environment given by the 

Hicks data (i.e., multiplied the Hicks refractory values by 2). The 01-D 

version enriched the near-field refractory environment by multiplying 

the Hicks refractory values by 4. Assumption of the enriched refractory 

near-field environment resulted in larger fission and activation product 

inhalation and ingestion correction factors (Tables 4-8 and 4-12) and 

increased number of affected organs and the resultant doses (Tables 

4-9 and 4-14). Incorporates formal internal and NIOSH review 

comments. Constitutes a total rewrite of the document. Training 

required: As determined by the Task Manager. Initiated by Eugene M. 

Rollins. 
Document No. ORAUT-TKBS-0008-4 Revision No. 01 Effective Date: 05/27/2008 Page 4 of 77

TABLE OF CONTENTS

SECTION TITLE

4.1 Introduction ............................................................................................................................

PAGE

10

4.1.1 Purpose......................................................................................................................11

4.1.2 Scope.........................................................................................................................11

4.2 Internal Inhalation and Ingestion Dose to Aboveground Workers ........................................... 11

4.2.1 Onsite Releases to Air................................................................................................ 12

4.2.1.1 Source Description....................................................................................... 12

4.2.1.1.1 Weapons Testing........................................................................ 12

4.2.1.1.2 Ground Seepage of Noble Gases ............................................... 13

4.2.1.1.3 Tunnel Operations ...................................................................... 13

4.2.1.1.4 Containment Ponds .................................................................... 13

4.2.1.1.5 Drillbacks.................................................................................... 13

4.2.1.1.6 Laboratories................................................................................ 13

4.2.1.1.7 Radioactive Waste Management Sites ....................................... 14

4.2.1.2 Atmospheric Radionuclide Concentrations................................................... 14

4.2.1.2.1 Annual Inhalation Intakes of Tritium............................................ 15

4.2.1.2.2 Annual Intakes of Plutonium ....................................................... 18

4.2.1.2.3 Annual Intakes of Other Radionuclides ....................................... 21

4.2.1.2.4 Scaling Factors for Inhalation Intakes ......................................... 22

4.2.1.2.5 Correction for Resuspension for Early Times after 

Atmospheric Tests...................................................................... 23

4.2.1.2.6 Correction for Inhalation Dose from Short-Lived 

Fission and Activation Products.................................................. 24

4.2.2 Annual Intakes from Ingestion .................................................................................... 25

4.3 Occupational External Dose...................................................................................................28

4.3.1 Ambient Radiation ...................................................................................................... 28

4.3.2 Releases of Noble Gases ........................................................................................... 31

4.4 Internal Inhalation and Ingestion Dose to Underground Workers............................................ 33

4.4.1 Underground Activities................................................................................................ 33

4.4.2 Radon Measurements ................................................................................................ 33

4.4.3 Underground Radon Concentrations .......................................................................... 34

4.4.4 Underground Workers’ Exposure to Radon ................................................................ 35

4.4.5 Underground Workers’ Exposure to Other Radionuclides........................................... 38

4.5 Instruction to Dose Reconstructors for Assignment of Environmental Intakes ........................ 38

4.5.1 Assigning External Ambient Dose ............................................................................... 38

4.5.2 Assigning Internal Ambient Dose................................................................................ 38

4.6 Attributions and Annotations ..................................................................................................39

References ........................................................................................................................................40

ATTACHMENT A, AMBIENT ENVIRONMENTAL INTAKES AT THE NEVADA TEST SITE 

BASED ON AIR SAMPLING AND SOIL CONTAMINATION DATA....................... 47
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A.1 Introduction ............................................................................................................................48

A.2 Air Sampling Data and Intake Estimates ................................................................................49

A.3 Soil Contamination and Decay Corrections ............................................................................52

A.4 Scaling Factors for Inhalation Intake Estimates......................................................................54

A.5 Correction for Resuspension for Early Times After Atmospheric Tests................................... 54

A.6 Corrections for Inhalation Dose From Short-lived Fission and Activation Products................. 57

A.7 Ingestion Pathway..................................................................................................................65

A.8 Instruction to Dose Reconstructors for Assignment of Environmental Intakes ........................ 73

ATTACHMENT B, TOTAL ANNUAL ORGAN DOSES FROM 30 YEARS OF INHALATION 

AND INGESTION INTAKES AND SOURCES OF OVERESTIMATED 

ORGAN DOSE .....................................................................................................74
Document No. ORAUT-TKBS-0008-4 Revision No. 01 Effective Date: 05/27/2008 Page 6 of 77

LIST OF TABLES

TABLE TITLE

4-1 HTO atmospheric concentrations by year and area with estimated maximum and 

average annual organ dose....................................................................................................

PAGE

17

4-2 Atmospheric concentrations of 239Pu for sampled areas......................................................... 19

4-3 Annual inhalation intakes from 239Pu for sampled areas ......................................................... 20

4-4 Inventory of contaminated soil................................................................................................ 21

4-5 Radionuclide areal soil deposition decay-corrected to 1963................................................... 22

4-6 Abundance of radionuclides in NTS soils relative to 239Pu decay-corrected to 1963............... 22

4-7 Scaled inhalation intakes corrected for early resuspension .................................................... 24

4-8 Organ-specific inhalation dose fission and activation product correction factors..................... 25

4-9 Inhalation dose from short-lived fission and activation products ............................................. 25

4-10 Radionuclide soil concentration by area................................................................................. 26

4-11 Area-specific and maximum annual ingestion rates................................................................ 26

4-12 Organ-specific ingestion fission and activation correction factors........................................... 27

4-13 Organ-specific annual ingestion doses for the 90Sr intake of 55.42 Bq/year ........................... 27

4-14 Organ-specific doses from ingestion of fission and activation products .................................. 28

4-15 Ambient background radiations levels by year ....................................................................... 29

4-16 Ambient radiation by area ...................................................................................................... 29

4-17 85Kr average background and the Area 20 net concentrations and annual organ doses 

from 1983 through 1992......................................................................................................... 32

4-18 Results of experiment to determine the effects of ventilation conditions on RDCs in the 

G-Tunnel inclined drift ............................................................................................................ 35

4-19 Radon daughter concentrations for N-Tunnel in 1991 and 1992............................................. 35

4-20 Radon daughter concentrations for P-Tunnel in 1991 and 1992............................................. 36

4-21 Annual 222Rn exposures and uncertainties for internal dose reconstruction for miners and 

tunnel workers........................................................................................................................36

A-1 Atmospheric concentrations of 239Pu for sampled areas......................................................... 50

A-2 Annual inhalation intakes from 239Pu for sampled areas ......................................................... 51

A-3 Inventory of contaminated soil................................................................................................ 53

A-4 Radionuclide areal soil deposition.......................................................................................... 53

A-5 Radionuclide areal soil deposition decay-corrected to 1963................................................... 54

A-6 Abundance of radionuclides in the NTS soils relative to 239Pu decay-corrected to 1963......... 55

A-7 Scaled inhalation intakes corrected for early resuspension in 1963........................................ 56

A-8 Far-field refractory enrichment ratios...................................................................................... 58

A-9 Organ-specific inhalation dose fission and activation product correction factors..................... 61

A-10 Inhalation dose from short-lived fission and activation products ............................................. 65

A-11 Radionuclide soil concentration by area................................................................................. 66

A-12 Area-specific and maximum annual ingestion rates................................................................ 66

A-13 Annual organ doses from ingestion of 238,239Pu and 241Am...................................................... 68

A-14 Organ-specific ingestion fission and activation correction factors........................................... 72

A-15 Organ-specific annual ingestion doses for the 90Sr intake of 55.42 Bq/yr ............................... 72

A-16 Organ specific doses from ingestion of fission and activation products .................................. 73

B-1 Total annual organ doses greater than 0.0001 rem from 30 years of inhalation and 

ingestion intakes ....................................................................................................................75
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LIST OF FIGURES

FIGURE TITLE

4-1 Air sampling stations in 1997 ................................................................................................. 

PAGE

16

4-2 Resuspension factor as a function of time after initial deposition............................................ 24

A-1 Resuspension factor as a function of time after initial deposition............................................ 56

A-2 Small Boy – 90Sr fraction of total dose lungs........................................................................... 59

A-3 Little Feller I – 90Sr fraction of total dose to lungs ................................................................... 60

A-4 Turk – 90Sr fraction of total dose to lungs ............................................................................... 60

A-5 Strontium-90 fraction of total dose.......................................................................................... 61

A-6 Strontium-90 fraction of total dose.......................................................................................... 62

A-7 Strontium-90 fraction of total dose.......................................................................................... 62

A-8 Strontium-90 fraction of total dose.......................................................................................... 63

A-9 Strontium-90 fraction of total dose to the liver ........................................................................ 63

A-10 Strontium-90 fraction of total dose red marrow....................................................................... 64

A-11 Strontium-90 fraction of total dose to bone surface ................................................................ 64

A-12 Strontium-90 fraction of total ingestion dose .......................................................................... 69

A-13 Strontium-90 fraction of total ingestion dose to red bone marrow........................................... 69

A-14 Strontium-90 fraction of total ingestion dose to red bone marrow........................................... 70

A-15 Strontium-90 fraction of total ingestion dose .......................................................................... 70

A-16 Strontium-90 fraction of ingestion dose to the upper large intestine ....................................... 71

A-17 Strontium-90 fraction of total ingestion dose .......................................................................... 71
Document No. ORAUT-TKBS-0008-4 Revision No. 01 Effective Date: 05/27/2008 Page 8 of 77

ACRONYMS AND ABBREVIATIONS

Bq becquerel

CFR Code of Federal Regulations

Ci curie 

cm centimeter

d day

DOD U.S. Department of Defense

DOE U.S. Department of Energy

DOL U.S. Department of Labor

EEOICPA Energy Employees Occupational Illness Compensation Program Act

EPA U.S. Environmental Protection Agency

ft foot

GSD geometric standard deviation

hr hour

HT airborne tritium

HTO tritiated water vapor

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

km kilometer

L liter

LANL Los Alamos National Laboratory

lb pound

LLNL Lawrence Livermore National Laboratory

m meter

MeV megaelectron-volt, 1 million electron-volts

mg milligram

mi mile

min minute

ml milliliter

mR milliroentgen

mrem millirem

NIOSH National Institute for Occupational Safety and Health

NTS Nevada Test Site

pCi picocurie

PERM passive environmental radon monitor

POC probability of causation
Document No. ORAUT-TKBS-0008-4 Revision No. 01 Effective Date: 05/27/2008 Page 9 of 77

REECo Reynolds Electrical & Engineering Company

RDC radon daughter concentration 

RPISU radon progeny integrating sampling unit

s second

SRDB Ref ID Site Research Database Reference Identification (number)

Sv sievert

TLD thermoluminescent dosimeter

TRU transuranic

TTR Tonopah Test Range

U.S.C. United States Code

WEF Waste Examination Facility

WL working level

WLM working level month

yr year

μCi microcurie

§ section or sections
Document No. ORAUT-TKBS-0008-4 Revision No. 01 Effective Date: 05/27/2008 Page 10 of 77

4.1 INTRODUCTION

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

facility encompassed by EEOICPA.

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

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

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

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

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

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

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

determined in accordance with the POC [probability of causation1

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

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

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

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

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

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

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

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

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

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

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

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

occupationally derived (NIOSH 2007):

] guidelines established under 

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

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

DOE employees with a covered cancer or restrict the “duty” to nuclear weapons work (NIOSH 2007).

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

conventional structures

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

reasons

 

1 The U.S. Department of Labor (DOL) is ultimately responsible under the EEOICPA for determining the POC. 
Document No. ORAUT-TKBS-0008-4 Revision No. 01 Effective Date: 05/27/2008 Page 11 of 77

4.1.1

This technical basis document (TBD) discusses occupational environmental dose. Occupational 

environmental dose is the dose individuals received at the Nevada Test Site (NTS or Test Site) while 

outside operational facilities but on the site during work activities. These doses can be internal or 

external depending on the characteristics of the individual radionuclides. While inhalation of most 

radionuclides would cause a dose to various organs in the body, noble gases would primarily cause 

only an external dose because these inert radionuclides are not readily absorbed by the body. 

However, inhalation of radon will result in dose to the lungs and respiratory airways due to the 

subsequent deposition and decay of daughter products. Occupational environmental dose would be 

received by workers in both the aboveground and the underground working environments.

Purpose

With the cessation of atmospheric testing at NTS in 1962, the greatest potential for environmental 

intakes of radioactive material in the aboveground environment results from the inhalation of 

radioactive particles that were resuspended from NTS soils into the atmosphere and from ingestion of 

soils that were previously contaminated by atmospheric nuclear weapons tests, reactor tests, and 

safety tests. The potential inhalation intakes can be estimated from air sampling data provided in the 

NTS annual environmental reports (see the Environmental Reports list in the References section) 

coupled with extensive soil contamination data gathered between 1983 and 1991 (McArthur 1983, 

1985, 1987, 1988, 1989, 1991). Because the air monitoring data was limited to gross alpha and beta 

measurements, tritium, and isotopes of plutonium (e.g., 238Pu, 239Pu, 240Pu), inhalation intakes of other 

relatively long-lived radionuclides that have been identified in the NTS soils (e.g., 241Am, 60Co, 137Cs, 

90Sr, 152Eu, 154Eu, 155Eu) are scaled to those of plutonium based on their relative abundance in the 

NTS soils. Ingestion intakes can also be estimated by assuming consumption of the contaminated 

NTS soils. To ensure that both inhalation and ingestion intakes are not underestimated, the relative 

abundances of the long-lived radionuclides identified in the NTS soils determined from the 1991 soil 

contamination data (McArthur 1991) were decay-corrected back to 1963. In addition, to ensure that 

intakes and resultant doses are not underestimated, correction factors are developed to account for 

potential exposures to short-lived fission and activation products based on test-specific data provided 

by Hicks (1981a,b,c,d). In addition, a correction factor is developed for inhalation intakes that 

accounts for the phenomenon of early resuspension (Anspaugh 2002).

4.1.2

Section 4.2 discusses internal doses to aboveground workers from onsite releases to the air and 

resuspension of radioactive materials in soil, as well as from ingestion of contaminated soils. 

Section 4.3 describes external doses to workers from ambient radiation and releases of radioactive 

noble gases into air. Section 4.4 discusses internal dose to underground workers from inhalation of 

ambient concentrations of radioactive materials in the air, ingestion of contaminated materials, and 

from exposure to radon. Section 4.5 provides instructions to dose reconstructors for assignment of 

environmental intakes. Section 4.6 presents attributions and annotations, which are indicated by 

bracketed callouts and used to identify the source, justification, or clarification of the associated 

information.

Scope

4.2 INTERNAL INHALATION AND INGESTION DOSE TO ABOVEGROUND WORKERS 

Section 4.2.1 discusses the internal dose for workers outside the facilities, as determined from air 

concentrations resulting from ground-level releases. Section 4.2.2 discusses the internal dose from 

resuspension of radioactive materials in the soil. 
Document No. ORAUT-TKBS-0008-4 Revision No. 01 Effective Date: 05/27/2008 Page 12 of 77

4.2.1

4.2.1.1 Source Description

Onsite Releases to Air

4.2.1.1.1 Weapons Testing

NTS has been the primary location for the testing of nuclear explosives in the continental United 

States since 1951. Test programs have included atmospheric testing in the 1950s and early 1960s, 

earth-cratering experiments, and open-air nuclear reactor and rocket engine testing. Since the 

mid-1960s, testing of nuclear devices has occurred underground in drilled vertical holes or in mined 

tunnels. No nuclear tests have been conducted since September 1992 (Black 1995). 

In all, more than 900 nuclear tests have taken place at the Test Site as part of these programs. One 

result of these tests is that the surface soils in many parts of NTS contain measurable amounts of 

several long-lived radionuclides. Almost all of the more than 100 aboveground tests contaminated the 

soil near ground zero. In addition, several underground tests were cratering experiments that threw 

radioactive rock and soil hundreds of feet, and some deeper underground tests vented radioactive 

material to the surface. A few safety tests, in which a nuclear device was destroyed by conventional 

explosives, scattered plutonium (and in some cases uranium) over the nearby ground. Further, there 

was fallout of radioactive debris from many tests over the northern and eastern parts of the Test Site 

(McArthur 1991).

Radiation levels at NTS have been monitored regularly, and safety officials have identified and fenced 

off areas where the soil is heavily contaminated. In many other areas, radionuclide levels are not high 

enough to warrant closing the area, but they are still above background. 

Atmospheric weapon and safety tests from 1951 and 1963 resulted in the release of approximately 

1.2 × 1010 Ci to the atmosphere (DOE 1996). Much of this activity was from relatively short-lived 

radionuclides that decayed in a matter of days or weeks. The volatile radionuclides (such as 

radioiodines, noble gases, and tritium) were diluted in the atmosphere and transported off the site. 

However, much of the nonvolatile, long-lived radionuclides settled back into the soils at various 

locations at NTS (see Table 4-10 in Section 4.2.2). These contaminated soils continue to represent a 

potential inhalation pathway to workers from resuspension of soils by wind and such mechanical 

activities as cleanup and remediation (McArthur 1991).

In 1963, nuclear weapons testing was moved underground to prevent the release of radionuclides to 

the atmosphere and achieve containment. Releases of radioactive material following an underground 

test are generally categorized with terms that describe both the volume of material released and the 

conditions of the release (DOE 1996):

• Containment Failures: Containment failures are unintentional releases of radioactive matter to 

the atmosphere due to failure of the containment system. A prompt massive release, or one 

that occurs soon after a test, is a venting.

• Late-Time Seeps: Late-time seeps are small releases that occur days or weeks after a test 

when gases diffuse through pore spaces in the overlying rock and are drawn to the surface by 

decreases in atmospheric pressure.

• Controlled Tunnel Purging: A controlled tunnel purge is an intentional release to allow either 

recovery of experimental data and equipment or reuse of part of the tunnel system.

• Operational Release: Operational releases are small consequential releases that occur when 

core or gas samples are collected, or when a drillback hole is sealed.
Document No. ORAUT-TKBS-0008-4 Revision No. 01 Effective Date: 05/27/2008 Page 13 of 77

At the present time, processing of radioactive materials at NTS includes only laboratory analyses. 

Handling of these materials includes only the transport and storage of nuclear explosive devices and 

the operation of a radioactive waste management site for low-level radioactive and mixed wastes. 

Monitoring and evaluation of the various activities indicate that the potential sources of onsite 

radiation exposure are releases from the following sources (Black 1995): 

• Tritiated water vapor (HTO) from drainage containment ponds for E Tunnel in Area 12

• Onsite radioanalytical laboratories

• Area 3 and 5 waste facilities

• Other diffuse sources

The following sections describe effluent sources at NTS from information in Black (1995).

4.2.1.1.2 Ground Seepage of Noble Gases

Ground seepage can be enhanced when changes in ambient pressure pump small amounts of noble 

gases up through the overburden and into the atmosphere from the cavity that is created by a nuclear 

test. This process, sometimes referred to as atmospheric pumping, creates a diffuse source of 

radiological effluents. These area sources are rare and therefore not routinely monitored. The 

phenomenon is usually restricted to tests conducted in the Pahute Mesa region of NTS. These 

seepages are from nuclear tests before 1993. 

4.2.1.1.3 Tunnel Operations

Nuclear tests occurred in mined tunnel complexes in the Rainier Mesa region. Because some tunnels 

were sealed in the mid-1990s, currently small amounts of contaminated water continue to drain from 

only one tunnel (Section 4.2.1.1.4). 

4.2.1.1.4 Containment Ponds

Water contaminated with radionuclides seeps from the tunnels in Area 12 and collects in containment 

ponds where some evaporates and some seeps into the soil. The only radiological contaminant that 

produces a measurable air emission from evaporation of the water is tritium (3

H) in the form of HTO. 

4.2.1.1.5 Drillbacks

Following underground nuclear tests, core samples have been taken from the cavity that was formed 

by the nuclear detonation for analysis and diagnosis. This core sampling is accomplished by drilling 

into the area of interest and recovering samples using special equipment. Radioactive material can 

escape to the atmosphere during these operations. 

4.2.1.1.6 Laboratories

The Reynolds Electrical & Engineering Company (REECo) conducted radiological analyses in 

Building 650, and Los Alamos National Laboratory (LANL) conducted similar analyses in Building 701 

at Mercury. Because these facilities have processed primarily environmental samples, very little 

radioactivity has passed through them. However, there was potential for some radionuclides to be 

discharged to the atmosphere through the hood ventilation system during sample processing, 

particularly of spiked samples or from loss of radioactive standards that contained heavy water, 

radioiodines, or noble gases. 
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4.2.1.1.7 Radioactive Waste Management Sites

Areas 3 and 5 contain sites for the disposal of low-level radioactive waste, and Area 5 contains sites 

for storage of transuranic and mixed transuranic wastes, as well as the Greater Confinement Disposal 

Test Unit and 12 accompanying boreholes (only a few contain waste). Disposal occurs in pits and 

trenches; concrete pads provide temporary storage of certain wastes. Area 5 is for packaged waste 

disposal only. The Waste Examination Facility (WEF) houses a glovebox with high-efficiency 

particulate air filtration that is used to examine and repack transuranic (TRU) waste drums. No 

contamination has been released from glovebox operations to the environment. The drums, which 

have been sent to NTS from Lawrence Livermore National Laboratory (LLNL) in past years, are 

stored inside the TRU Pad Cover Building. Repacked drums will be sent to the Waste Isolation Pilot 

Plant. The facility is a diffuse source of radiological effluents. The only radioactive effluent that has 

been detected by the various types of samplers around the site is HTO in atmospheric moisture. The 

Area 3 low-level waste site is in a location where surface soil has been contaminated by deposited 

plutonium, and resuspension of this soil by wind and vehicular activity has resulted in detection of 

above-background levels of plutonium in nearby air samples.

4.2.1.2 Atmospheric Radionuclide Concentrations

In 1964, REECo established an environmental surveillance program at NTS that was designed to 

measure radiological conditions throughout the site without regard to nuclear tests (Glora and Brown 

1964a). That is, the collected data was not to relate to specific tests but to general conditions of 

radiation. The short-term objective of the program was to minimize casual personnel exposure to 

radiation by locating and identifying localized radiological environmental conditions by type and 

quantity of contamination. The long-range objective of the program was to establish baseline 

environmental data that could provide a reference for comparison with subsequent test activities and 

radiation measurements (Glora and Brown 1964a).

The initial surveillance program included, over time, 12 permanent air-sampling stations in the most 

populated areas at NTS. The air samplers were low-volume Filter Queen samplers with 8- by 10-in. 

(Gelman Type E) glass-fiber filter papers. Operating times were determined by integrated electric

timers with flow rates from calibrated rotometers. Typical flow rates varied from 3 to 6 ft3

/min; 

samples were collected weekly (Glora and Brown 1964b).

After the first reporting period (June 1964), positive-displacement Gast pumps, which were equipped 

with in-line total volume gas meters, replaced the Filter Queen samplers. In May and June of 1965, 

the 8- by 10-in. Gelman filters were replaced with a new sampler that used 4-in.-diameter 

Whatman #41 filter paper (Lewis, Glora, and Aoki 1965). The sampling rate of these samplers was 

about 3 ft3

/min. Therefore, the total volume of sampled air in a 7-day period was about 1 × 103 m3. 

During this period, the number of sampling stations increased to 13 and caustic scrubbers were 

added for the detection of radioiodines.

Early particulate samples were typically analyzed only for gross alpha and gross beta. However, if 

gross beta concentrations exceeded 1 × 10-5 μCi/m3

, researchers conducted an analysis for 227Ac (the 

most hazardous beta emitter present). Because no historical evidence exists that 227Ac has been 

detected in air or soil samples, the assumption that unidentified beta emitters were 227Ac would be 

unreasonable and inappropriate. Therefore, for purposes of dose reconstruction, 90Sr is assumed to 

be the unidentified beta-emitting radionuclide with the highest internal dose impact (see Section 

4.2.1.2.4 for details).

Environmental samples were analyzed for gross alpha and beta radioactivity by gas proportional 

counting. Air samples were analyzed with a Nuclear Chicago ULTRASCALER system with a 

9- by 10-in. detection chamber to accommodate the 8- by 10-in. filters. All other samples were 
Document No. ORAUT-TKBS-0008-4 Revision No. 01 Effective Date: 05/27/2008 Page 15 of 77

analyzed with a Beckman WIDEBETA system equipped with an automatic sample changer. This 

counting equipment and protocol allowed lower detection limits of about 2.9 × 10-9 and 

1.6 × 10-8 μCi/m3 for gross alpha and gross beta, respectively (Lewis, Glora, and Aoki 1965). By 

1978, the lower detection limit for gross beta counting had been reduced to 1 × 10-10 μCi/m3

 (Lewis,

Glora, and Aoki 1965).

Gross gamma screening was typically performed on samples that were prepared for gross beta 

counting. The screening was performed using a 5- by 5-in. NaI(Tl) detector that was coupled to a 

single-channel analyzer. Any samples that showed elevated gamma readings were sent to a 

multichannel analyzer for radionuclide identification (Lewis, Glora, and Aoki 1965). In the early 1970s, 

gamma spectroscopy was performed using germanium-lithium [Ge(Li)] detectors and 2,000-channel 

analyzers with a lower detection limit of 5 × 10-9 μCi/m3 (REECO 1978a). After cessation of 

atmospheric testing at NTS in 1963, fission products were frequently measured in the atmosphere but 

were typically correlated with foreign atmospheric weapons testing (see Section 4.2.1.2.4).

In 1971, weekly air samples from a given station were batched on a monthly basis and subjected to 

radiochemical analysis for 239Pu. The procedure included acid dissolution with ion exchange recovery 

and electroplating onto stainless-steel discs, which was followed by alpha spectroscopy using 

solid-state surface barrier detectors (REECO 1978a). This analysis provided a nominal minimum 

detection limit of 3 × 10-7 μCi/m3 (Lantz 1978a). Routine analysis for 238Pu started in 1989 (Wruble

and McDowell 1990a).

In 1977, a separate sampler was designed for the collection of airborne tritium (HT) and HTO. The 

sampler was portable and capable of unattended operation for up to 2 weeks in desert areas. A small 

electronic pump drew air into the sampler at about 0.5 L/min, and the HTO was removed from the air 

stream by a silica gel drying column. The dry air then passed through a catalytic converter with 

platinum to generate HTO from HT. Another drying column collected the vapor, in which a small 

volume of distilled water served as a trap for the HTO. Appropriate aliquots of condensed moisture 

were obtained by heating the silica gel. Counting via liquid scintillation techniques allowed for the 

determination of the HT and HTO activities. The typical minimum detection limit for this analysis was 

3 × 10-7 μCi/m3 (REECO 1978a). 

The total number of air sampling stations increased over the years to a peak of 52 in 1989 (Wruble 

and McDowell 1990a). This number remained fairly constant until a gradual reduction began in 1995 

(Black and Townsend 1996). This reduction occurred primarily because of a gradual strategy shift 

from environmental monitoring to demonstration of compliance with National Emission Standards for 

Hazardous Air Pollutants as approved by the U.S. Environmental Protection Agency (i.e., 40 CFR 

Part 61) (Black and Townsend 1997). Figure 4-1 shows the locations of the 48 sampling stations in 

1997 (Black and Townsend 1998).

Since the late 1980s, the environmental surveillance program has routinely monitored atmospheric 

concentrations of tritium, 238Pu, 239Pu, and 240Pu. These radionuclides were considered the most 

important to dose for workers and members of the public (Wruble and McDowell 1990a). In addition, 

since the mid-1960s, measurements have been reported for gross alpha and gross beta 

concentrations (Glora and Brown 1964b). The following sections discuss the history of these 

measurements in the NTS annual environmental reports (see Reference section) and their importance 

to dose reconstruction for unmonitored employees.

4.2.1.2.1 Annual Inhalation Intakes of Tritium

Atmospheric measurements at NTS began in 1977 with samplers in Areas 5, 10, and 23 (REECO 

1978a). The sampler near Building 650 (Mercury) was designated as the control station. During that 

year, the samples in Area 10 (near the Sedan Crater) demonstrated the highest tritium (HTO) 
Document No. ORAUT-TKBS-0008-4 Revision No. 01 Effective Date: 05/27/2008 Page 16 of 77

concentrations with a high of 3.0 × 10-4 pCi/m3 (REECO 1978a). The number of sampling locations 

increased over the following years to include all the areas in Table 4-1. These locations provided 

representative samples from the most populated areas on the site (Black and Townsend 1998).

Figure 4-1. Air sampling stations in 1997. 

The tritium concentrations in Table 4-1 are typically the average of the maximum concentrations for a 

given area in a given year, except when these values were determined to be adversely affected by 

nearby point-source releases (e.g., Building 790 in 1983; Scoggins 1983, p. 29). In cases where 

maximum values were not provided, the average concentrations were reported. 
Document No. ORAUT-TKBS-0008-4 Revision No. 01 Effective Date: 05/27/2008 Page 17 of 77

Table 4-1. HTO atmospheric concentrations (pCi/ml) by year and area with estimated maximum and average annual organ dose (mrem).a

Year

Area at NTS

Site 

average

(pCi/ml)

Site 

maximum

(pCi/ml)

Site 

maximum 

annual 

dose

(mrem)

Site 

average 

annual 

dose

1 3 5 6 10 12 15 16 18 20 23 25 (mrem)

1977 3.0E-04 1.2E-06 1.5E-04 3.0E-04 6.9E-02 3.5E-02

1978 2.9E-03 1.3E-04 1.9E-05 1.9E-05 3.0E-05 6.2E-04 2.9E-03 6.7E-01 1.4E-01

1979 6.2E-04b 2.9E-03b 6.7E-01b 1.4E-01b

1980 2.6E-04 7.1E-05 1.7E-04 2.6E-04 6.0E-02 3.8E-02

1981 3.6E-03 9.8E-04 2.3E-03 3.6E-03 8.3E-01 5.3E-01

1982 7.5E-04 2.3E-03 4.4E-03 7.6E-04 5.0E-03 2.6E-03 5.0E-03 1.2E+00 6.1E-01

1983 8.2E-05 1.1E-04 8.9E-05 3.9E-03 6.5E-04 9.6E-04 3.9E-03 8.9E-01 2.2E-01

1984 1.7E-04 5.3E-04 3.3E-05 3.5E-05 5.0E-03 1.2E-03 5.0E-03 1.2E+00 2.7E-01

1985 1.4E-04 8.6E-05 4.7E-03 3.5E-05 4.7E-04 1.1E-03 4.7E-03 1.1E+00 2.5E-01

1986 1.2E-04 5.6E-05 6.6E-05 6.0E-05 8.4E-04 2.3E-04 8.4E-04 1.9E-01 5.3E-02

1987 4.5E-05 1.7E-04 6.3E-05 4.6E-04 2.6E-04 2.0E-04 4.6E-04 1.0E-01 4.6E-02

1988 4.4E-05 3.9E-05 2.0E-05 5.2E-04 3.2E-05 1.3E-04 5.2E-04 1.2E-01 3.0E-02

1989 1.7E-04 3.1E-05 1.3E-05 1.9E-05 8.2E-05 4.6E-06 5.3E-05 1.7E-04 3.9E-02 1.2E-02

1990 7.4E-06 2.2E-05 1.3E-05 4.0E-06 2.3E-05 7.5E-05 2.1E-05 2.4E-05 7.5E-05 1.7E-02 5.5E-03

1991 9.1E-06 2.8E-05 6.3E-06 8.4E-06 1.7E-05 4.3E-06 2.1E-06 1.1E-05 2.8E-05 6.5E-03 2.5E-03

1992 1.0E-05 3.5E-05 3.3E-06 6.2E-05 6.2E-05 6.1E-06 8.8E-05 3.8E-05 8.8E-05 2.0E-02 8.8E-03

1993 8.7E-06 2.8E-05 7.9E-06 3.7E-06 2.0E-05 4.0E-06 1.9E-06 1.1E-05 2.8E-05 6.5E-03 2.4E-03

1994 4.00E-06 4.70E-05 5.10E-06 2.20E-06 2.60E-06 3.70E-05 4.10E-06 2.90E-06 1.06E-05 2.81E-05 6.74E-03 2.55E-03

1995 3.7E-06 1.4E-06 2.1E-05 9.5E-07 7.0E-06 4.3E-06 1.0E-05 2.6E-06 1.4E-06 5.8E-06 2.1E-05 4.9E-03 1.3E-03

1996 3.1E-06 4.1E-06 1.6E-05 5.6E-05 2.6E-05 2.0E-05 1.4E-05 1.1E-06 1.1E-06 1.6E-05 5.6E-05 1.3E-02 3.6E-03

1997 2.2E-06 3.8E-06 7.5E-06 5.6E-05 2.9E-05 2.3E-05 2.0E-05 1.3E-06 1.3E-06 1.6E-05 5.6E-05 1.3E-02 3.7E-03

1998 3.4E-06 1.7E-05 1.8E-04 2.9E-05 5.5E-05 1.4E-05 4.6E-04 1.1E-04 4.6E-04 1.1E-01 2.5E-02

1999 2.0E-05 1.8E-05 4.2E-06 4.1E-05 2.8E-05 2.7E-05 7.5E-04 1.3E-04 7.5E-04 1.7E-01 2.9E-02

2000 4.1E-06 1.5E-05 5.1E-05 5.4E-05 1.9E-05 9.7E-04 1.9E-04 9.7E-04 2.2E-01 4.3E-02

2001 4.2E-06 3.6E-06 6.2E-06 4.2E-05 2.5E-05 3.8E-06 2.3E-06 1.3E-06 1.1E-03 7.1E-07 6.1E-06 1.1E-04 1.1E-03 2.5E-01 2.5E-02

a. Sources: Environmental Reports list in the References section.

b. Value not provided in annual reports - assumed same as for 1978.
Document No. ORAUT-TKBS-0008-4 Revision No. 01 Effective Date: 05/27/2008 Page 18 of 77

In addition to the annual area tritium concentrations, Table 4-1 provides NTS site average and 

maximum concentrations, which represent the arithmetic averages of the concentrations of all the 

areas and the maximum of all the areas, respectively. For purposes of dose reconstruction, these 

average and maximum site concentrations have been converted to annual organ dose using the dose 

conversion factor for inhalation of 1.80 × 10-11 Sv/Bq (ICRP 1998) and by assuming submersion in the 

respective concentrations for 2,000 hr/yr and an inhalation rate of 2,400 m3

/yr. 

In addition to the inhalation dose, a factor of 1.5 was included in the organ dose calculation to account 

for absorption through the skin.

The uncertainty that is associated with the values in Table 4-1 was not provided in the annual 

environmental report until 2001. Based on the values that were provided for that year, the standard 

deviation would be typically less than 10% of the measured values. Because sampling methods were 

changed over time and the unavailability of uncertainty estimates for previous years, the favorable to 

claimant assumption is made that the 50th-percentile expected intakes are those in Table 4-1 and the 

95th-percentile values are twice the Table 4-1 values.

For purposes of dose reconstruction, the geometric standard deviation (GSD) of the values in 

Table 4-2 would be estimated by using the following expression for a lognormal distribution:

)()( 1.645

1

50th percentile

95th percentile GSD = (4-1)

If the assumption is made that the 95th-percentile values are twice the Table 4-1 values, then the 

GSD can be shown using Equation 4-1 to be 1.52.

If the favorable to claimant assumption is used to estimate the 95th-percentile dose (i.e., add 100%) 

from the maximum annual doses in Table 4-1, the resultant annual organ doses would still be less 

than 2 mrem/yr. 

4.2.1.2.2 Annual Intakes of Plutonium

Routine isotopic atmospheric measurements of plutonium at NTS began in 1971 with samplers in 

15 locations across the site (REECO 1978a). Six additional sampling stations were added in 1978 

(REECO 1979). Equipment at fixed locations continuously sampled the ambient air to monitor for 

radioactive materials. These locations were chosen to provide representative samples from the 

populated areas on the site as well as to monitor resuspension of low-fired plutonium that was spread 

by safety experiments before 1960 in Areas 2, 3, 4, 7, 9, and 10. Access, worker population, 

geographical coverage, presence of radioactivity, and availability of electric power were 

considerations in the site selection for air samplers (Black and Townsend 1997).

In 1988, efforts to monitor radioactive air emissions at NTS were increased as a result of the 

requirements of DOE Order 5400.1. Known and potential effluent sources throughout NTS were 

assessed for their potential to contribute to public dose (Black and Townsend 1997).

The 239Pu concentrations in Table 4-2 for 1989 through 2001 represent the average of the maximum 

concentrations for a given area in a given year. In cases where maximum values were not provided 

(i.e., 1971 through 1988), the average of the average concentrations was reported. In addition to the 

annual area concentrations, Table 4-2 provides NTS site average and maximum concentrations, 

which represent the arithmetic averages of the concentration of all the areas and the maximum of the 

area maximum or area averages of all the areas, respectively. Potential intakes associated with these 

concentrations can be calculated under the assumption that an unmonitored worker was 
Document No. ORAUT-TKBS-0008-4 Revision No. 01 Effective Date: 05/27/2008 Page 19 of 77

occupationally exposed for 2,000 hr/yr and a breathing rate of 2,400 m3/yr. Table 4-3 presents these 

calculated intakes.

Table 4-2. Atmospheric concentrations of 239Pu for sampled areas (pCi/m3

).a 

Year

Area

1 2 3 4 5 6 7 9 10 11 12

1971 3.7E-04 6.6E-04 1.7E-04 1.2E-04 1.7E-04 7.2E-04 2.1E-04 9.1E-05 1.6E-04

1972 1.5E-04 2.2E-04 3.7E-04 1.4E-04 1.3E-03 4.3E-03 2.9E-04 2.4E-04 7.9E-04

1973 8.7E-05 2.2E-04 2.1E-04 8.7E-05 1.2E-04 8.6E-04 4.4E-05 2.4E-04 4.3E-05

1974 8.0E-05 6.8E-05 1.3E-04 5.0E-05 7.9E-05 2.1E-04 5.7E-05 8.0E-05 5.8E-05

1975 2.9E-05 3.8E-05 1.5E-04 4.3E-05 6.3E-05 1.7E-04 5.7E-05 4.9E-05 3.3E-05

1976 2.9E-04 1.7E-04 1.0E-03 1.5E-04 4.4E-04 3.2E-03 2.1E-04 2.4E-04 2.8E-04

1977 2.7E-05 6.5E-05 2.6E-04 3.4E-05 5.5E-05 2.5E-04 4.6E-05 1.2E-04 2.1E-05

1978 4.7E-05 1.3E-04 2.8E-04 6.1E-05 9.9E-05 5.4E-04 2.9E-04 1.1E-04 6.2E-05

1979 7.4E-05 4.7E-04 1.7E-06 2.1E-05 4.0E-05 4.9E-05 5.2E-04 2.9E-04 3.7E-05 2.4E-05

1980 2.3E-05 2.6E-04 1.1E-06 2.9E-05 4.8E-05 4.5E-05 4.5E-04 4.4E-05 1.9E-05

1981 2.6E-05 6.4E-05 1.4E-04 2.5E-05 2.9E-05 3.3E-05 3.2E-04 4.6E-05 2.0E-05

1982 6.1E-05 4.0E-05 6.2E-05 2.0E-05 3.5E-05 7.4E-05 2.2E-04 5.3E-05 2.7E-05

1983 3.1E-05 6.5E-05 9.2E-05 2.3E-05 3.2E-05 3.3E-05 2.1E-04 1.8E-04 2.3E-05

1984 1.9E-05 1.0E-04 2.3E-04 3.8E-05 2.0E-05 3.8E-05 1.0E-03 5.6E-05 1.7E-05

1985 1.5E-05 5.8E-05 2.1E-04 3.1E-05 2.7E-05 2.7E-05 1.3E-03 5.2E-05 2.3E-05

1986 1.5E-04 6.2E-05 5.5E-04 2.8E-05 1.9E-04 4.8E-05 2.8E-04 2.5E-05 3.2E-05

1987 5.5E-05 2.3E-05 2.8E-03 1.7E-05 2.5E-05 1.6E-05 1.1E-04 3.2E-05 1.6E-05

1988 9.2E-05 2.8E-05 2.1E-04 2.8E-05 2.2E-05 1.5E-05 5.1E-05 1.2E-05 1.6E-05

1989 8.6E-04 5.8E-05 5.4E-04 5.0E-05 8.0E-05 2.7E-04 3.5E-04 5.4E-04 6.8E-04

1990 9.9E-05 3.2E-05 3.5E-04 3.1E-05 3.7E-05 4.1E-05 5.0E-04 4.3E-05 1.1E-04 1.0E-05

1991 3.1E-03 2.1E-05 3.8E-04 3.5E-05 7.2E-05 3.2E-05 3.5E-04 5.4E-07 2.2E-04 1.4E-05

1992 1.5E-04 1.1E-04 2.3E-03 8.0E-05 7.3E-05 3.9E-04 8.8E-04 2.4E-04 8.8E-05 3.8E-05

1993 2.0E-04 4.9E-05 4.0E-04 8.0E-05 9.0E-05 7.8E-05 2.7E-03 1.7E-04 4.2E-04 8.4E-04

1994

1995 4.1E-05 1.1E-05 2.4E-04 1.5E-05 5.0E-05 1.2E-05 3.9E-04 3.8E-05 4.0E-05 1.1E-05

1996 7.5E-04 3.0E-04 1.6E-05 4.6E-05 4.5E-04 6.1E-04 5.8E-05 1.9E-05 3.7E-06

1997 1.1E-05 2.6E-05 1.1E-05 1.5E-06

1998 4.7E-04 6.7E-05 9.5E-05 1.1E-05 2.7E-05 7.4E-04 9.5E-06

1999 5.9E-05

2000 8.5E-05 2.7E-04 5.9E-05 2.1E-05 4.1E-04 3.4E-05 2.8E-03

2001 4.5E-04 1.2E-04 1.2E-04 1.9E-04 6.8E-05 9.6E-05 1.7E-05 5.0E-04 8.5E-06

Year

Area Site 

average

Site 

15 16 18 19 20 23 25 27 28 maximum

1971 1.9E-04 8.0E-05 2.3E-04 9.4E-05 8.4E-05 3.9E-05 2.26E-04 7.21E-04

1972 1.7E-04 1.5E-04 2.7E-03 3.5E-04 1.7E-04 5.1E-05 7.63E-04 4.29E-03

1973 4.4E-05 3.8E-05 7.2E-05 5.6E-05 3.7E-05 3.3E-05 1.46E-04 8.59E-04

1974 6.0E-05 6.6E-05 8.5E-05 8.6E-05 7.2E-05 1.2E-04 8.64E-05 2.11E-04

1975 4.3E-05 3.4E-05 4.4E-05 5.7E-05 6.8E-05 7.0E-05 6.27E-05 1.68E-04

1976 1.6E-04 9.9E-05 5.7E-04 2.0E-04 1.4E-04 6.3E-05 4.82E-04 3.18E-03

1977 2.6E-05 2.3E-05 2.7E-05 3.3E-05 3.1E-05 2.3E-05 6.93E-05 2.55E-04

1978 5.1E-05 6.2E-05 5.8E-05 7.0E-05 9.4E-05 3.9E-05 1.33E-04 5.42E-04

1979 5.2E-05 2.1E-05 2.0E-05 2.2E-05 1.8E-05 1.6E-05 2.3E-05 9.98E-05 5.24E-04

1980 6.4E-05 9.5E-05 3.5E-05 3.1E-05 2.1E-05 1.6E-05 7.88E-05 4.52E-04

1981 9.3E-05 2.6E-05 2.3E-05 2.9E-05 2.4E-05 2.5E-05 1.6E-05 5.88E-05 3.22E-04

1982 4.2E-05 1.7E-05 2.2E-05 2.1E-05 1.7E-05 5.04E-05 2.15E-04

1983 2.2E-05 1.5E-05 3.0E-05 2.8E-05 3.0E-05 5.85E-05 2.14E-04

1984 3.2E-05 2.8E-05 2.5E-05 2.9E-05 3.2E-05 4.8E-05 1.15E-04 1.02E-03

1985 4.0E-05 2.1E-05 3.5E-05 2.2E-05 2.6E-05 3.1E-05 1.30E-04 1.33E-03

1986 3.9E-05 2.8E-05 2.2E-05 6.6E-05 2.1E-05 2.3E-05 1.03E-04 5.45E-04

1987 1.8E-05 1.4E-05 1.7E-05 1.6E-05 2.3E-05 2.13E-04 2.81E-03

1988 2.4E-05 1.4E-05 3.5E-05 1.5E-05 1.5E-05 4.15E-05 2.14E-04

1989 2.3E-05 1.1E-05 9.0E-06 1.2E-04 9.5E-06 4.8E-06 2.40E-04 8.55E-04

1990 8.2E-05 5.8E-06 3.9E-06 8.5E-05 1.1E-05 3.5E-05 9.17E-05 4.98E-04

1991 1.2E-04 2.9E-05 8.8E-06 1.3E-05 2.6E-05 7.7E-06 2.77E-04 3.10E-03

1992 1.1E-03 1.9E-05 8.9E-06 1.3E-05 1.8E-05 1.6E-04 3.52E-04 2.26E-03

1993 3.3E-04 1.3E-05 7.7E-06 6.3E-06 3.4E-06 1.1E-05 3.37E-04 2.70E-03

1994

1995 1.6E-04 1.4E-05 1.0E-06 7.4E-06 4.5E-06 6.92E-05 3.90E-04

1996 2.4E-04 1.8E-06 3.5E-06 4.5E-06 6.3E-06 4.2E-06 2.6E-06 1.57E-04 7.50E-04

1997 2.9E-05 1.9E-07 2.0E-05 2.0E-05 6.0E-07 1.0E-06 1.6E-06 1.11E-05 2.90E-05

1998 6.3E-05 3.1E-05 3.3E-05 1.2E-05 6.3E-06 1.30E-04 7.35E-04

1999 1.1E-05 1.1E-05 2.0E-05 6.2E-06 2.14E-05 5.90E-05
Document No. ORAUT-TKBS-0008-4 Revision No. 01 Effective Date: 05/27/2008 Page 20 of 77

Year

Area

1 2 3 4 5 6 7 9 10 11 12

2000 2.6E-04 1.0E-05 6.6E-06 1.1E-05 3.63E-04 2.83E-03

2001 1.4E-05 9.6E-06 1.7E-05 7.6E-06 2.0E-06 4.4E-06 1.02E-04 5.04E-04

a. Sources: Environmental Reports list in the References section.

Table 4-3. Annual inhalation intakes from 239Pu for sampled areas (Bq) (from Attachment A).

Year

Area

1 2 3 4 5 6 7 9 10 11 12

1971 0.0329 0.0587 0.0147 0.0105 0.0153 0.0640 0.0185 0.0081 0.0145

1972 0.0131 0.0199 0.0327 0.0124 0.1190 0.3810 0.0258 0.0216 0.0698

1973 0.0077 0.0198 0.0183 0.0078 0.0107 0.0763 0.0039 0.0216 0.0038

1974 0.0071 0.0060 0.0112 0.0044 0.0070 0.0187 0.0051 0.0071 0.0052

1975 0.0026 0.0034 0.0130 0.0038 0.0056 0.0149 0.0051 0.0044 0.0029

1976 0.0255 0.0153 0.0924 0.0133 0.0392 0.2824 0.0189 0.0210 0.0248

1977 0.0024 0.0058 0.0226 0.0030 0.0049 0.0225 0.0041 0.0107 0.0019

1978 0.0042 0.0118 0.0250 0.0054 0.0088 0.0481 0.0258 0.0095 0.0055

1979 0.0066 0.0413 0.0001 0.0018 0.0035 0.0044 0.0465 0.0258 0.0033 0.0021

1980 0.0020 0.0230 0.0001 0.0026 0.0043 0.0040 0.0401 0.0039 0.0017

1981 0.0023 0.0056 0.0124 0.0022 0.0026 0.0029 0.0286 0.0041 0.0018

1982 0.0054 0.0036 0.0055 0.0018 0.0031 0.0066 0.0191 0.0047 0.0024

1983 0.0028 0.0057 0.0082 0.0020 0.0029 0.0029 0.0190 0.0162 0.0020

1984 0.0017 0.0088 0.0202 0.0034 0.0018 0.0034 0.0906 0.0050 0.0015

1985 0.0013 0.0051 0.0189 0.0027 0.0024 0.0024 0.1181 0.0046 0.0020

1986 0.0131 0.0055 0.0484 0.0025 0.0165 0.0043 0.0249 0.0022 0.0028

1987 0.0048 0.0020 0.2495 0.0015 0.0022 0.0014 0.0098 0.0028 0.0014

1988 0.0082 0.0025 0.0190 0.0025 0.0020 0.0013 0.0045 0.0011 0.0014

1989 0.0759 0.0052 0.0478 0.0045 0.0071 0.0240 0.0311 0.0480 0.0604

1990 0.0088 0.0029 0.0308 0.0027 0.0033 0.0036 0.0442 0.0038 0.0095 0.0009

1991 0.2753 0.0018 0.0340 0.0031 0.0064 0.0029 0.0311 0.0000 0.0199 0.0013

1992 0.0133 0.0098 0.2007 0.0071 0.0065 0.0346 0.0781 0.0213 0.0078 0.0034

1993 0.0180 0.0044 0.0352 0.0071 0.0079 0.0069 0.2398 0.0151 0.0373 0.0746

1994

1995 0.0036 0.0010 0.0216 0.0014 0.0045 0.0011 0.0346 0.0033 0.0036 0.0010

1996 0.0666 0.0000 0.0268 0.0015 0.0041 0.0400 0.0542 0.0051 0.0017 0.0003

1997 0.0010 0.0023 0.0010 0.0001

1998 0.0416 0.0060 0.0084 0.0010 0.0024 0.0653 0.0008

1999

2000 0.0075 0.0239 0.0052 0.0018 0.0366 0.0030 0.2509

2001 0.0404 0.0011 0.0103 0.0166 0.0061 0.0086 0.0015 0.0447 0.0008

Year

Area Site 

average

Site 

15 16 18 19 20 23 25 27 28 maximum

1971 0.0171 0.0071 0.0201 0.0083 0.0075 0.0035 0.0200 0.0640

1972 0.0150 0.0129 0.2424 0.0309 0.0147 0.0045 0.0677 0.3810

1973 0.0039 0.0034 0.0063 0.0050 0.0033 0.0029 0.0130 0.0763

1974 0.0053 0.0059 0.0075 0.0076 0.0064 0.0104 0.0077 0.0187

1975 0.0038 0.0030 0.0039 0.0051 0.0060 0.0062 0.0056 0.0149

1976 0.0139 0.0088 0.0504 0.0178 0.0122 0.0056 0.0428 0.2824

1977 0.0023 0.0020 0.0024 0.0029 0.0028 0.0020 0.0062 0.0226

1978 0.0045 0.0055 0.0052 0.0062 0.0083 0.0035 0.0118 0.0481

1979 0.0046 0.0019 0.0017 0.0020 0.0016 0.0014 0.0020 0.0089 0.0465

1980 0.0057 0.0085 0.0031 0.0027 0.0019 0.0014 0.0070 0.0401

1981 0.0083 0.0023 0.0020 0.0025 0.0022 0.0022 0.0014 0.0052 0.0286

1982 0.0037 0.0015 0.0019 0.0019 0.0015 0.0045 0.0191

1983 0.0020 0.0013 0.0026 0.0025 0.0027 0.0052 0.0190

1984 0.0028 0.0025 0.0022 0.0025 0.0028 0.0043 0.0102 0.0906

1985 0.0036 0.0019 0.0031 0.0020 0.0023 0.0028 0.0115 0.1181

1986 0.0035 0.0025 0.0020 0.0058 0.0018 0.0020 0.0092 0.0484

1987 0.0016 0.0012 0.0015 0.0014 0.0020 0.0000 0.0189 0.2495

1988 0.0022 0.0012 0.0031 0.0013 0.0013 0.0037 0.0190

1989 0.0021 0.0010 0.0008 0.0104 0.0008 0.0004 0.0213 0.0759

1990 0.0073 0.0005 0.0003 0.0075 0.0010 0.0031 0.0081 0.0442

1991 0.0109 0.0026 0.0008 0.0012 0.0023 0.0007 0.0246 0.2753

1992 0.0977 0.0017 0.0008 0.0011 0.0016 0.0142 0.0312 0.2007

1993 0.0293 0.0012 0.0007 0.0006 0.0003 0.0010 0.0300 0.2398

1994 0.0000 0.0000

1995 0.0142 0.0012 0.0001 0.0007 0.0004 0.0061 0.0346

1996 0.0213 0.0002 0.0003 0.0000 0.0004 0.0006 0.0004 0.0002 0.0124 0.0666
Document No. ORAUT-TKBS-0008-4 Revision No. 01 Effective Date: 05/27/2008 Page 21 of 77

1997 0.0026 0.0000 0.0018 0.0018 0.0001 0.0001 0.0001 0.0010 0.0026

1998 0.0056 0.0027 0.0029 0.0010 0.0006 0.0115 0.0653

1999 0.0010 0.0010 0.0018 0.0006 0.0011 0.0018

2000 0.0229 0.0009 0.0006 0.0010 0.0322 0.2509

2001 0.0012 0.0008 0.0015 0.0007 0.0002 0.0004 0.0090 0.0447

It should be noted that some covered employees remained on the site continuously for weeks at a 

time. However, because most of the nonworking hours were spent indoors where ambient air 

particulate loadings would be much less than the outdoor loadings and because of the conservative 

assumptions that were used to estimate the values in Table 4-3, adjustment of the tabular data is not 

required to ensure intakes are not underestimated for these individuals. In addition, employees who 

lived on the site during their work week would have been housed in Area 12 or Area 23 (Mercury). 

For most years, the values in Table 4-2 for these locations are less than the site average values. For 

purposes of dose reconstruction, the inhalation intake values in Table 4-3 can be assumed to be 

bounding; therefore, the organ dose that results from these intakes should be applied with a constant 

distribution. 

It is assumed that plutonium could be any of absorption Type S, Super S, or M dependent on which 

type delivers the maximum organ dose. Because these doses are based on air monitoring results, 

evaluation for Super S (ORAUT 2007a) is required for lung and thoracic cancers if there were no 

bioassay results for the employee.

4.2.1.2.3 Annual Intakes of Other Radionuclides

Extensive studies were performed in the 1980s to quantify residual contamination at NTS (McArthur 

1983, 1985, 1987, 1988, 1989, 1991). Table 4-4 lists the results of these studies (McArthur 1991). 

Table 4-5 lists the total areal depositions based on the inventory values in Table 4-4 divided by the 

respective areal size. It should be noted that the results in Table 4-5 are representative of areas of 

the NTS that have been shown to contain measurable levels of contamination. These areas actually 

represent only about one-third of the total area within the boundaries of NTS.

Table 4-4. Inventory of contaminated soil (Ci). 

Area Area (mi2) Am-241 Pu-238 Pu-239,240 Co-60 Cs 137 Sr-90 Eu-152 Eu-154 Eu-155

1 26.5 4.2 6.5 24 1.1 8.8 15 15 0.1 0.5

2 19.7 2.9 8.6 22 1.2 24 46 14 0.4

3 32.3 4.6 3.1 37 1 12 33 18 0.1 0.5

4 16 6.6 13 40 1.6 12 13 9.1 0.2

5 2.9 0.6 0.1 4.8 0.6 0.4 0.9 10 0.2 0

6 32.3 1.7 3.3 8.4 0.2 2.8 3.5 0

7 19.3 2.2 0.6 16 1 5.2 9.2 22 0.2 0.3

8 13.9 17 8 110 5.7 42 25 4.4 0.6

9 20 4.2 2.2 89 0.7 8.7 13 23 0.2 0.3

10 20 19 19 110 9.7 84 55 2.2 0.3 5

11 4 3.3 0.5 29 0 0.5 0.3

12 39.6 5.7 8.5 39 1.2 20 17

15 35.3 8 7.8 63 0.3 19 22

16 14.3 0.7 1.5 3.7 0.1 2.9 3.7

17 31.4 2.8 4.5 18 1.0 15 19

18 27.3 19 5.6 100 0.7 10 17 1.1 0.1 0.8

19 148.3 21 32 140 1.1 36 31

20 6.2 23 30 41 7.9 5.5 4.3 13 1.6 4.8

25 0.9 0.2 0.1 0.4

26 0.2

30 0.03 3.2 4.5 14 0.8 1.5 1.3 0.7 0.1 0.2

Source: McArthur (1991).
Document No. ORAUT-TKBS-0008-4 Revision No. 01 Effective Date: 05/27/2008 Page 22 of 77

In addition, because the data in Table 4-4 is representative of the soil contamination in 1991, the 

values in Table 4-5 were decay-corrected to the beginning of 1963 – the first year after the cessation 

of atmospheric testing. 

Table 4-5. Radionuclide areal soil deposition decay-corrected to 1963 (Bq/m2

) 

Area Am-241 Pu-238 Pu-239,240 Co-60 Cs 137 Sr-90 Eu-152 Eu-154 Eu-155

1 2.37E+03 4.37E+03 1.29E+04 2.35E+04 9.06E+03 1.57E+04 3.47E+04 4.89E+02 1.35E+04

2 2.20E+03 7.78E+03 1.60E+04 3.45E+04 3.32E+04 6.49E+04 4.35E+04 0.00E+00 1.45E+04

3 2.13E+03 1.71E+03 1.64E+04 1.76E+04 1.01E+04 2.84E+04 3.41E+04 4.01E+02 1.11E+04

4 6.16E+03 1.45E+04 3.57E+04 5.67E+04 2.05E+04 2.26E+04 3.48E+04 8.93E+03

5 3.09E+03 6.15E+02 2.37E+04 1.17E+05 3.76E+03 8.63E+03 2.11E+05 8.94E+03 0.00E+00

6 7.86E+02 1.82E+03 3.72E+03 3.51E+03 2.36E+03 3.01E+03

7 1.70E+03 5.54E+02 1.19E+04 2.94E+04 7.35E+03 1.33E+04 6.98E+04 1.34E+03 1.11E+04

8 1.83E+04 1.03E+04 1.13E+05 2.33E+05 8.24E+04 5.00E+04 1.94E+04 3.08E+04

9 3.14E+03 1.96E+03 6.36E+04 1.98E+04 1.19E+04 1.81E+04 7.04E+04 1.30E+03 1.07E+04

10 1.42E+04 1.69E+04 7.86E+04 2.75E+05 1.15E+05 7.65E+04 6.74E+03 1.94E+03 1.79E+05

11 1.23E+04 2.23E+03 1.04E+05 0.00E+00 3.41E+03 2.09E+03

12 2.15E+03 3.83E+03 1.41E+04 1.72E+04 1.38E+04 1.19E+04

15 3.39E+03 3.94E+03 2.55E+04 4.82E+03 1.47E+04 1.73E+04

16 7.31E+02 1.87E+03 3.70E+03 3.97E+03 5.53E+03 7.20E+03

17 1.33E+03 2.55E+03 8.20E+03 1.81E+04 1.30E+04 1.68E+04

18 1.04E+04 3.66E+03 5.24E+04 1.45E+04 9.99E+03 1.73E+04 2.47E+03 4.75E+02 2.09E+04

19 2.12E+03 3.85E+03 1.35E+04 4.21E+03 6.62E+03 5.81E+03

20 5.54E+04 8.62E+04 9.45E+04 7.23E+05 2.42E+04 1.93E+04 1.28E+05 3.34E+04 5.53E+05

25 6.06E+03 3.09E+03 2.72E+04

26

30 1.59E+05 2.67E+05 6.67E+05 1.51E+06 1.36E+05 1.21E+05 1.43E+05 4.32E+04 4.76E+05

4.2.1.2.4 Scaling Factors for Inhalation Intakes

Because the air sampling program did not provide isotopic analyses for all of the radionuclides 

identified in the NTS soils, scaling factors are developed to estimate potential intakes for these 

radionuclides based on the relative abundances of these radionuclides compared to 239Pu when the 

soil contamination data (McArthur 1991) has been decay corrected to 1963. These area-specific 

ratios are provided in Table 4-6.

Table 4-6. Abundance of radionuclides in NTS soils relative to 239Pu decay-corrected to 1963. 

Area Am-241 Pu-238 Pu-239,240 Co-60 Cs 137 Sr-90 Eu-152 Eu-154 Eu-155

1 0.183 0.338 1.000 1.818 0.700 1.216 2.677 0.038 1.041

2 0.138 0.487 1.000 2.164 2.081 4.067 2.726 0.908

3 0.130 0.104 1.000 1.072 0.619 1.735 2.084 0.024 0.675

4 0.172 0.405 1.000 1.587 0.572 0.632 0.975 0.250

5 0.131 0.026 1.000 4.958 0.159 0.365 8.925 0.378

6 0.212 0.490 1.000 0.944 0.636 0.810

7 0.144 0.047 1.000 2.479 0.620 1.118 5.890 0.113 0.937

8 0.162 0.091 1.000 2.055 0.728 0.442 0.171 0.273

9 0.049 0.031 1.000 0.312 0.187 0.284 1.107 0.020 0.168

10 0.181 0.215 1.000 3.498 1.457 0.973 0.086 0.025 2.271

11 0.119 0.021 1.000 0.033 0.020

12 0.153 0.272 1.000 1.220 0.978 0.848

15 0.133 0.154 1.000 0.189 0.575 0.679

16 0.198 0.505 1.000 1.072 1.495 1.945

17 0.163 0.312 1.000 2.204 1.590 2.053

18 0.199 0.070 1.000 0.278 0.191 0.331 0.047 0.009 0.400

19 0.157 0.285 1.000 0.312 0.491 0.431

20 0.586 0.912 1.000 7.643 0.256 0.204 1.358 0.354 5.849

25

26

30 0.229 0.321 1.000 0.057 0.107 0.093 0.050 0.007 0.014

Maximum scaling 

factor

0.586 0.912 1.000 7.64 2.08 4.07 8.93 0.378 5.85

Scaled maximum 

intake (Bq/yr)

0.223 0.347 0.381 2.91 0.792 1.55 3.40 0.144 2.23
Document No. ORAUT-TKBS-0008-4 Revision No. 01 Effective Date: 05/27/2008 Page 23 of 77

The scaling factors in Table 4-6 are used in conjunction with the 239Pu intakes in Table 4-3 to 

determine the potential environmental intakes of all the radionuclides important to dose (in Table 4-6) 

that have been identified as persistent in NTS soils. For purposes of dose reconstruction, maximum 

intakes of these radionuclides are calculated by selecting the maximum annual intake of plutonium 

(i.e., 0.381 Bq/yr derived for Area 9 in 1972) and multiplying this value by the maximum scaling factor 

for each of the radionuclides in Table 4-6. 

4.2.1.2.5 Correction for Resuspension for Early Times after Atmospheric Tests

Anspaugh (2002) stated that, based on empirical observations, concentration of resuspended 

radionuclides in air have been noted to display a strong time dependence at early times after 

deposition and that this pathway could be important for reoccupation of contaminated property. He 

goes on to state that there has not been universal agreement that resuspension is an important 

pathway but that it is now generally accepted that there are a few instances where the pathway could 

be the dominant one. Many observations have shown that the rate of resuspension decreases rapidly 

with time and, for accident situations, resuspension is only of importance (in comparison with the 

inhalation exposure from the initial cloud passage) over short periods. For this reason, Anspaugh 

(2002) states that the resuspension factor model has been widely used to predict the concentration of 

resuspended radionuclides at early times after initial deposition while the mass loading model (which 

uses measurements of dust loading in air and soil contamination data to predict air concentration of 

radionuclides) has generally been preferred for times long after deposition. However, Anspaugh also 

states that it is always preferable to rely on actual measurements that are performed over long 

periods (such as those in Section 4.2.1.2.2).

Anspaugh (2002) presents several resuspension models that have been proposed but concludes that 

they can be over- or under-predictive at various times after deposition when compared to empirical 

observations. However, with expanded data sets from the work of Hicks (1981e) and others in the 

1980s, Anspaugh has proposed a resuspension model that more accurately describes the observed 

results over the entire time span of the expanded data set for the NTS:

[ ] 07.05 003.09 9 1 10 106 1010 −− −− − ± = ×+ ×+ t t

f eS e m-1 (4-2)

A graphical depiction of Equation 4-2 is provided in Figure 4-2 from time t equal zero to 1,000 d after 

detonation. As shown in Figure 4-2, the resuspension factor Sf ranges from about 10-5 at times early 

after deposition, falls rapidly during the first 100 or so days to a value of about 10-8, and then 

approaches a value of 10-9 after a few years. The factor of 10 at the end of the equation is a 

statement of uncertainty in the model. 

If it is true that the mass loading approach is more predictive at time long after initial deposition and 

that the resuspension proposed by Anspaugh is predictive of the observed results over the expanded 

dataset (including those that were developed in the 1980s), then the factor of 10-9 could be taken to 

be the resuspension factor that would be predictive of the mass loading process that is thought to be 

more important during the times when the air monitoring data are available (i.e., 1971 through 2001, 

see Section 4.2.1.2.2). 

When Anspaugh’s proposed resuspension model (i.e., Equation 4-2) is integrated over 180 to 

545 days and compared to the result of the integral of the constant 10-9 over the same period, a factor 

is developed that can be used to correct the intakes derived from air sampling data (i.e., times long 

after initial deposition) for the early resuspension phenomenon that has been observed at the NTS. 
Document No. ORAUT-TKBS-0008-4 Revision No. 01 Effective Date: 05/27/2008 Page 24 of 77

Resuspension Factor

1.3 E-09 m-1

1.E-09

1.E-08

1.E-07

1.E-06

1.E-05

0 100 200 300 400 500 600 700 800 900 1000

Days

meter-1

Figure 4-2. Resuspension factor as a function of time after initial deposition 

(from Attachment A). 

The period of integration was selected to begin at day 180 because the last atmospheric tests at the 

NTS were in July of 1962 and therefore about 180 d had passed before the beginning of 1963 (the 

starting date for reconstruction of internal dose at the NTS in the absence of bioassay results for the 

employee). For the Anspaugh (2002) model, the early resuspension correction factor has been 

determined to be 3.12. In a similar manner, correction factors for 1964 and 1965 were determined to 

be 1.72 and 1.24, respectively. Thus, for dose reconstruction purposes, the scaled maximum 

inhalation intakes in Table 4-6 should be multiplied by these factors to account for early resuspension 

from 1963 through 1965. If necessary, these intakes can be prorated for time less than a year for a 

best estimate if the worker was on the site for only a fraction of the year.

4.2.1.2.6 Correction for Inhalation Dose from Short-Lived Fission and Activation Products

Correction factors were developed to account for inhalation intakes of short-lived fission and activation 

products based on organ-specific dose from the 90Sr intakes in Table 4-7 (from Attachment A). These 

corrections are shown in Table 4-8.

Table 4-7. Scaled inhalation intakes corrected for early resuspension (Bq/yr). 

Year of intake Am-241 Pu-238 Pu-239, 240 Co-60 Cs 137 Sr-90 Eu-152 Eu-154 Eu-155

1963 0.70 1.08 1.19 9.08 2.47 4.84 10.61 0.45 6.96

1964 0.38 0.59 0.65 4.98 1.35 2.65 5.81 0.25 3.81

1965 0.28 0.43 0.47 3.61 0.98 1.92 4.22 0.18 2.77

All subsequent years 0.223 0.347 0.381 2.91 0.792 1.55 3.40 0.144 2.23

The Integrated Modules for Bioassay Analysis (IMBA) computer program was used to determine the 

organ doses resulting from the scaled intakes of 90Sr in Table 4-7 for a period of 10 years (i.e., 

6.20 Bq for 1963 and 1.68 Bq/yr from 1964 through 1972). These doses were then multiplied by the 

correction factors in Table 4-8 to determine the additional dose that should be added to account for 

potential dose from inhalation of short-lived fission and activation products. The organ-specific fission 

and activation doses greater than 0.001 rem are provided in Table 4-9. Other organs are not shown 

in Table 4-9 because all were less than 0.001 rem. These doses should be added to the input for the 
Document No. ORAUT-TKBS-0008-4 Revision No. 01 Effective Date: 05/27/2008 Page 25 of 77

Table 4-8. Organ-specific inhalation dose fission and activation product correction factors. Organ Fission and activation product correction factor

Year 1963 1964 1965 1966 1967 1968 1969 1970 1971 1972

Skin Adrenals

Thymus SI 

Spleen Skin 

Muscle Uterus

Pancreas Kidneys

Breast Testes

Esophagus Ovaries

Brain Stomach

Thyroid Gall Bladder

730 364 242 182 145 121 104 90.7 80.6 72.5

ULI 458 179 99.2 64.0 45.0 33.4 25.9 20.6 16.8 14.0

Urinary Bladder 335 149 91.3 63.6 47.5 37.2 30.1 24.9 21.0 18.0

Lungs 34,900 14,200 7,960 5,150 3,630 2,700 2,100 1,660 1,360 1,130

ET ET1 ET2 LN(TH)

LN(ET)

1,570 827 598 492 438 412 412 412 412 412

LLI 420 142 70.8 42.4 28.2 20.1 15.1 11.7 9.4 7.6

Colon 390 148 79.4 50.0 34.5 25.2 19.3 15.2 12.3 10.2

Liver 9,260 4,620 1,540 1,190 988 858 769 706 661 629

Red Bone Marrow 37.9 18.2 12.8 10.4 10.4 10.4 10.4 10.4 10.4 10.4

Bone Surfaces 78.5 40.1 28.1 22.4 22.4 22.4 22.4 22.4 22.4 22.4

Table 4-9. Inhalation dose from short-lived fission and activation products (rem). 

Year 1963 1964 1965 1966 1967 1968 1969 1970 1971 1972

Liver 0.002

Bone Surface 0.001 0.002 0.002 0.002 0.002 0.002 0.002 0.002 0.002 0.002

LLI 0.001

ET 0.002

Lung 0.008 0.003 0.002 0.001

ET1 1.635 0.474 0.248 0.164 0.146 0.138 0.138 0.138 0.138 0.138

Interactive RadioEpidemiological Program (IREP) as a constant as 100% 30- to 250-keV photons. 

These doses are input as constants because they are based on bounding intakes (i.e. maximum) and 

they are entered as 30- to 250-keV photons as an assumption favorable to claimants.

4.2.2

To account for potential intakes from inadvertent ingestion of contaminated soil, the area-specific 

radionuclide soil deposition data in Table 4-5 were converted to volumetric data (i.e., Bq/g) by 

assuming a radionuclide relaxation depth of 2.3 cm (DOE 2003) and a soil density of 1.5 g/cm3 (DOE 

2003). The area-specific radionuclide soil concentrations are presented in Table 4-10.

Annual Intakes from Ingestion

If the assumption is made that the workers ingested 100 mg of soil each day (EPA [1989] 

recommends a value of 50 mg/d) and the assumption is made that full-time employment was 250 d/yr, 

annual ingestion can be calculated in a manner favorable to claimants. The area-specific annual 

ingestion rates are in provided in Table 4-11. 

It should be noted that for most radionuclides, Area 30 provided the highest areal deposition and 

resultant intakes. Area 30 is relatively small (150 km2

), inaccessible, and is located on the Western 

edge of the NTS. It has rugged terrain and includes the northern reaches of Fortymile Canyon. In 

1968, it was the site of Project BUGGY, the first nuclear row-charge experiment in the PLOWSHARE 

Program. As a result of the test, a trench 255 m long, 77 m wide, and 206 m deep was created. The 

test resulted in large quantities of vitrified glass. Because of the bias that is introduced when Area 30 
Document No. ORAUT-TKBS-0008-4 Revision No. 01 Effective Date: 05/27/2008 Page 26 of 77

Table 4-10. Radionuclide soil concentration by area (Bq/g). 

Area 241Am 238Pu 239, 240Pu 60Co 137Cs 90Sr 152Eu 154Eu 155Eu

1 0.069 0.127 0.375 0.682 0.263 0.456 1.005 0.014 0.391

2 0.064 0.226 0.463 1.001 0.963 1.882 1.262 0.420

3 0.062 0.050 0.475 0.509 0.294 0.824 0.989 0.012 0.321

4 0.179 0.420 1.036 1.644 0.593 0.655 1.010 0.259

5 0.090 0.018 0.686 3.401 0.109 0.250 6.122 0.259

6 0.023 0.053 0.108 0.102 0.069 0.087 0.000 0.00

7 0.049 0.016 0.344 0.852 0.213 0.384 2.024 0.039 0.322

8 0.530 0.297 3.280 6.741 2.389 1.450 0.562 0.894

9 0.091 0.057 1.844 0.575 0.344 0.524 2.042 0.038 0.311

10 0.411 0.491 2.279 7.973 3.321 2.217 0.195 0.056 5.177

11 0.357 0.065 3.005 0.099 0.060

12 0.062 0.111 0.408 0.498 0.399 0.346

15 0.098 0.114 0.740 0.140 0.426 0.502

16 0.021 0.054 0.107 0.115 0.160 0.209

17 0.039 0.074 0.238 0.524 0.378 0.488

18 0.301 0.106 1.518 0.421 0.290 0.502 0.072 0.014 0.607

19 0.061 0.111 0.391 0.122 0.192 0.169

20 1.607 2.500 2.741 20.946 0.701 0.559 3.723 0.969 16.031

25 0.176 0.090 0.789

26

30 4.620 7.749 19.340 43.835 3.953 3.493 4.143 1.252 13.804

Table 4-11. Area-specific and maximum annual ingestion rates (Bq/yr). 

Area 241Am 238Pu 239, 240Pu 60Co 137Cs 90Sr 152Eu 154Eu 155Eu

1 1.72 3.17 9.38 17.06 6.56 11.41 25.12 0.35 9.77

2 1.59 5.64 11.57 25.03 24.08 47.06 31.54 10.51

3 1.54 1.24 11.87 12.72 7.34 20.59 24.73 0.29 8.01

4 4.47 10.49 25.90 41.10 14.83 16.37 25.24 6.47

5 2.24 0.45 17.15 85.03 2.73 6.25 153.05 6.48 0.00

6 0.57 1.32 2.69 2.54 1.71 2.18

7 1.23 0.40 8.59 21.29 5.33 9.61 50.59 0.97 8.05

8 13.24 7.43 81.99 168.52 59.73 36.25 14.05 0.00 22.34

9 2.27 1.42 46.11 14.38 8.60 13.10 51.04 0.94 7.76

10 10.29 12.27 56.98 199.31 83.02 55.42 4.88 1.41 129.41

11 8.93 1.61 75.12 0.00 2.47 1.51

12 1.56 2.77 10.20 12.45 9.98 8.65

15 2.45 2.85 18.49 3.49 10.64 12.56

16 0.53 1.35 2.68 2.87 4.01 5.21

17 0.97 1.85 5.94 13.09 9.44 12.19

18 7.54 2.65 37.95 10.54 7.24 12.55 1.79 0.34 15.17

19 1.53 2.79 9.78 3.05 4.80 4.21

20 40.17 62.49 68.51 523.64 17.54 13.98 93.06 24.23 400.77

25 4.39 2.24 19.73

26

30 115.50 193.72 483.50 1,095.89 98.83 87.33 103.56 31.30 345.10

Maxa 40.17 62.49 81.99 523.64 83.02 55.42 153.05 24.23 400.77

a. Maximum value with Area 30 excluded.

is included, the maximum annual intakes in Table 4-11 have been provided without the Area 30 areal 

concentrations.

Similar to what was done for inhalation intakes, a method was developed to adjust ingestion doses for 

potential dose resulting from short-lived fission and activation products that are no longer persistent in 

the NTS soils in measurable amounts. The organ-specific fission and activation product correction 
Document No. ORAUT-TKBS-0008-4 Revision No. 01 Effective Date: 05/27/2008 Page 27 of 77

factors were developed based on the relative contribution of 90Sr to the total ingestion dose. These 

correction factors are presented in Table 4-12. 

Table 4-12. Organ-specific ingestion fission and activation correction factors (from Attachment A). 

Organ Correction factor

Year 1963 1964 1965 1966 1967 1968 1969 1970 1971 1972

Adrenals Breast 

Brain Skin Bladder

Stomach Kidneys Muscle 

Pancreas Brain Esophagus 

SI Liver Ovaries ET ET1

ET2 LM(ET) LN(TH) Lungs 

Skin Spleen Testes 

Thymus Thyroid Uterus

Gall Bladder

416 219 155 123 106 95.0 88.1 84.0 82.0 81.8

ULI 514 184 95.2 58.2 39.3 18.6 16.3 14.6 13.4 12.4

Bone Surface / RBM 3.8 3.1 2.7 2.5 2.3 2.3 2.3 2.3 2.3 2.3

LLI / Colon 417 208 138 25.6 20.4 17.4 14.6 12.8 11.3 10.2

To simplify the application of organ-specific ingestion dose from short-lived fission and activation 

products, the IMBA computer program was used to determine the organ-specific annual doses for the 

90Sr intake of 55.42 Bq/yr for 1963 through 1972. These doses are shown in Table 4-13. The doses 

in Table 4-13 are then multiplied by the Table 4-12 organ-specific correction factors to provide the 

annual doses resulting from ingestion of short-lived fission and activation products. The annual 

organ-specific doses that were determined to be greater than 0.001 rem are provided in Table 4-14.

Table 4-13. Organ-specific annual ingestion doses (rem) for the 90Sr intake of 55.42 Bq/yr. Organ 1963 1964 1965 1966 1967 1968 1969 1970 1971 1972

Adrenals 2.11E-06 2.47E-06 2.64E-06 2.77E-06 2.89E-06 3.00E-06 3.08E-06 3.16E-06 3.22E-06 3.28E-06

Urinary 

bladder

5.68E-06 6.42E-06 6.67E-06 6.87E-06 7.04E-06 7.18E-06 7.31E-06 7.41E-06 7.50E-06 7.58E-06

Bone surface 1.58E-04 3.74E-04 5.47E-04 6.98E-04 8.32E-04 9.50E-04 1.06E-03 1.15E-03 1.23E-03 1.31E-03

Brain 2.11E-06 2.47E-06 2.64E-06 2.77E-06 2.89E-06 3.00E-06 3.08E-06 3.16E-06 3.22E-06 3.28E-06

Breast 2.11E-06 2.47E-06 2.64E-06 2.77E-06 2.89E-06 3.00E-06 3.08E-06 3.16E-06 3.22E-06 3.28E-06

Esophagus 2.11E-06 2.47E-06 2.64E-06 2.77E-06 2.89E-06 3.00E-06 3.08E-06 3.16E-06 3.22E-06 3.28E-06

Stomach 3.42E-06 3.78E-06 3.95E-06 4.09E-06 4.21E-06 4.31E-06 4.40E-06 4.47E-06 4.54E-06 4.59E-06

Small intestine 4.71E-06 5.07E-06 5.24E-06 5.38E-06 5.50E-06 5.60E-06 5.69E-06 5.76E-06 5.83E-06 5.88E-06

Upper large 

intestine

3.42E-05 3.43E-05 3.43E-05 3.43E-05 3.43E-05 3.43E-05 3.43E-05 3.43E-05 3.43E-05 3.43E-05

Lower large 

intestine

1.45E-04 1.45E-04 1.45E-04 1.45E-04 1.45E-04 1.45E-04 1.45E-04 1.45E-04 1.45E-04 1.45E-04

Colon 4.36E-03 2.26E-03 1.51E-03 1.14E-03 9.16E-04 7.71E-04 6.61E-04 5.84E-04 5.21E-04 4.71E-04

Kidneys 2.11E-06 2.47E-06 2.64E-06 2.77E-06 2.89E-06 3.00E-06 3.08E-06 3.16E-06 3.22E-06 3.28E-06

Liver 2.11E-06 2.47E-06 2.64E-06 2.77E-06 2.89E-06 3.00E-06 3.08E-06 3.16E-06 3.22E-06 3.28E-06

Muscle 2.11E-06 2.47E-06 2.64E-06 2.77E-06 2.89E-06 3.00E-06 3.08E-06 3.16E-06 3.22E-06 3.28E-06

Ovaries 2.11E-06 2.47E-06 2.64E-06 2.77E-06 2.89E-06 3.00E-06 3.08E-06 3.16E-06 3.22E-06 3.28E-06

Pancreas 2.11E-06 2.47E-06 2.64E-06 2.77E-06 2.89E-06 3.00E-06 3.08E-06 3.16E-06 3.22E-06 3.28E-06

Red marrow 1.08E-04 2.66E-04 3.87E-04 4.86E-04 5.69E-04 6.37E-04 6.94E-04 7.42E-04 7.81E-04 8.14E-04

Extrathoracic 

airways

2.11E-06 2.47E-06 2.64E-06 2.77E-06 2.89E-06 3.00E-06 3.08E-06 3.16E-06 3.22E-06 3.28E-06

Lungs 2.11E-06 2.47E-06 2.64E-06 2.77E-06 2.89E-06 3.00E-06 3.08E-06 3.16E-06 3.22E-06 3.28E-06

Skin 2.11E-06 2.47E-06 2.64E-06 2.77E-06 2.89E-06 3.00E-06 3.08E-06 3.16E-06 3.22E-06 3.28E-06

Spleen 2.11E-06 2.47E-06 2.64E-06 2.77E-06 2.89E-06 3.00E-06 3.08E-06 3.16E-06 3.22E-06 3.28E-06

Testes 2.11E-06 2.47E-06 2.64E-06 2.77E-06 2.89E-06 3.00E-06 3.08E-06 3.16E-06 3.22E-06 3.28E-06

Thymus 2.11E-06 2.47E-06 2.64E-06 2.77E-06 2.89E-06 3.00E-06 3.08E-06 3.16E-06 3.22E-06 3.28E-06

Thyroid 2.11E-06 2.47E-06 2.64E-06 2.77E-06 2.89E-06 3.00E-06 3.08E-06 3.16E-06 3.22E-06 3.28E-06

Uterus 2.11E-06 2.47E-06 2.64E-06 2.77E-06 2.89E-06 3.00E-06 3.08E-06 3.16E-06 3.22E-06 3.28E-06
Document No. ORAUT-TKBS-0008-4 Revision No. 01 Effective Date: 05/27/2008 Page 28 of 77

Table 4-14. Organ-specific doses (rem) from ingestion of fission and activation products. 

Year 1963 1964 1965 1966 1967 1968 1969 1970 1971 1972

Bladder 0.002 0.002 0.001

Bone surface 0.001 0.002 0.002 0.002 0.003 0.003 0.003 0.003 0.003

Stomach 0.001

SI 0.002 0.001

ULI 0.015 0.006 0.003 0.002 0.001

LLI 0.048 0.025 0.017 0.003 0.003 0.002 0.002 0.002 0.001 0.001

Colon 0.028 0.015 0.010 0.002 0.002 0.001 0.001 0.001

4.3 OCCUPATIONAL EXTERNAL DOSE

Workers incur external doses from the ambient radiation levels and from noble gases. Ambient 

radiation measurements were reported in NTS annual environmental reports (see References section) 

starting in 1967; however, no ambient radiation data were provided in the reports for 1968 to 1976. 

4.3.1

Before 1967, ambient radiation levels that were unaffected by weapons testing were not reported in 

the annual environmental reports. Although there were many radiation measurements completed 

between 1951 and 1967, most of these were to characterize the effects of weapons tests and were 

therefore not appropriate for use in estimating external environmental dose for unmonitored 

employees.

Ambient Radiation

In 1967, ambient radiation levels were measured using Victoreen Model 239 indirect reading 

ionization chambers with effective ranges of 0 to 10 mR/hr. Five of these chambers were in small, 

semiprotective enclosures in NTS living areas (REECO 1968). As a backup for the ionization 

chambers, standard NTS film dosimeters were included at each sample location. However, all results 

from these film dosimeters for 1967 were either zero or lost due to light or heat damage. Results from 

environmental NTS film dosimeters were subsequently discarded because of the likelihood of heat 

damage.

The ionization chambers were collected on a weekly basis and read on a Victoreen Minometer II 

reader. Corrections were made for background and for nonradiation induced drift. Two sets of 

chambers were used, one set at the sample locations for measurements and another set stored fully 

charged in the laboratory. Each week these sets were exchanged, the fresh set being recharged and 

a record kept of the amount of "drift" while stored in the laboratory. A specially designed shock-proof 

box for transportation minimized accidental discharge from mechanical shock.

Readings from the five chambers in each location were averaged to obtain a mean value for each 

location each week. Readings significantly higher than others at a particular location were not used in 

compiling the data because the abnormal readings were most likely the result of shock or other 

malfunction and were not representative measurements. Because of the method of background 

subtraction that is discussed above, the data in the 1967 annual report were assumed to be due to 

manmade radiation in excess of natural background.

No results were reported between 1968 and 1976 in the NTS annual environmental reports, however, 

ambient radiation reporting was reestablished in 1977 using thermoluminescent dosimeters (TLDs) 

(REECO 1978a). The dosimeters used were CaF2:Dy (TLD-200) 0.25- by 0.25- by 0.035-in. chips 

from Harshaw Chemical Company. A badge that consisted of at least two chips shielded by 0.047 in. 

of cadmium (1030 mg/cm2

) inside a 0.050-in. black plastic (140 mg/cm2) holder was placed about 1 m 

above the ground at each of 10 locations that coincided with air sampling stations. These sites were 

selected because of their proximity to workers. During that year, the natural background ambient 

radiation level was established at 0.26 mrem/d or about 95 mrem/yr. 
Document No. ORAUT-TKBS-0008-4 Revision No. 01 Effective Date: 05/27/2008 Page 29 of 77

In 1987, the Harshaw dosimeters were replaced with Panasonic Model UD-814 TLDs (Gonzalez 

1988). These specifically designed environmental dosimeters contain three identical CaSO4:Tm 

elements and one Li2B4O7:Cu element. The lithium element is shielded with 14 mg/cm2 of material to 

monitor beta particles in the environment. The three calcium elements are encapsulated in 

1,000 mg/cm2 of plastic and lead to monitor ambient gamma levels. The UD-814 TLDs have 

remained in use to the present time.

In subsequent years, the number of sampling locations increased to over 150 and eventually covered 

all populated areas at NTS. Table 4-15 lists average ambient background radiation levels. 

Table 4-16 lists the annual measured ambient radiation including background and site average and 

maximum values. For purposes of dose reconstruction, because the values in Table 4-16 represent 

continuous exposure for an entire year, these values should be adjusted for occupational exposure 

(i.e., 2,000 hr/yr) and added to the dose of record for unmonitored employees. If the area in which the 

employee worked is known, the average value for that area should be used. If the area in which the 

employee worked is not known, the site maximum value should be used to be favorable to claimants.

Table 4-15. Ambient background radiations levels by year.

a

Year mrem/yr- Year mrem/yr- Year mrem/yr1964 (b) 1977 95 1990 110

1965 (b) 1978 95 1991 112

1966 (b) 1979 95 1992 109

1967 (b) 1980 99 1993 131

1968 (b) 1981 110 1994 93

1969 (b) 1982 100 1995 94

1970 (b) 1983 100 1996 91

1971 (b) 1984 100 1997 95

1972 (b) 1985 100 1998 88

1973 (b) 1986 72 1999 91

1974 (b) 1987 102 2000 115

1975 (b) 1988 131 2001 114

1976 (b) 1989 106

a. Sources: Environmental Reports list in the References section.

b. Not reported.

It should also be noted that some covered employees remained onsite continuously for weeks at a 

time. However, because most of the non-working hours were spent indoors where elevated ambient 

radiation would be reduced by shielding from the building and because of the conservative 

assumptions used to estimate the values in Table 4-16, adjustment of the tabular data is not required 

to ensure ambient exposures are not underestimated for these individuals.

It is likely that the elevated ambient radiation peaked in the early 1960s near the end of aboveground 

testing. Therefore, it would be reasonable and probably favorable to claimants to assign maximum 

annual environmental doses for 1963 through 1966 equal to the maximum value reported for 1967 

(i.e., 318 mrem/yr). It should be noted that for 1967, the maximum value is about 50-percent higher 

than the average value for all areas where measurements were reported. Following this logic, 

maximum doses have been added to Table 4-16 (shown in italics) for 1969 through 1977 for which 

data were not readily available. These values were based on the maximum value for 1978 and 

should be used to adjust environmental dose for unmonitored employees for those years.

Table 4-16. Ambient radiation by area (mrem/yr).

a,b

Area

Year 1 2 3 4 5 6 7 8 9 10 11 12 13 14

1967 318 194 205

1968 285 183 175 190
Document No. ORAUT-TKBS-0008-4 Revision No. 01 Effective Date: 05/27/2008 Page 30 of 77

1969

1970

1971

1972

1973

1974

1975

1976

1977

1978 130 167 200 138 110 110 120 150 320 140 150

1979 119 147 190 140 133 104 125 115 140 168 155 191

1980 123 156 199 155 133 106 130 135 145 192 165 196

1981 120 155 218 150 166 113 125 150 180 160 187

1982 126 149 188 145 141 108 125 140 209 140 189

1983 115 141 181 143 144 103 95 135 201 140 174

1984 97 149 162 116 111 86 327 125 102 166 112 155

1985 141 78 101 107 347 94 114 225 180

1986 93 126 150 92 92 77 318 94 96 120 107 106

1987 146 189 184 164 191 126 120 149 181 133 168

1988 155 209 207 164 177 134 150 179 203 158 194

1989 140 217 205 132 141 106 126 151 179 153 139

1990 134 167 187 142 142 109 122 148 170 132 140

1991 128 168 194 146 165 116 132 124 132 150 130 190

1992 123 167 173 141 150 104 145 128 136 157 158 170

1993 147 178 200 158 198 134 165 153 156 172 184 201

1994 62 62 89 84 70 102 97

1995 102 136 150 121 212 87 176 102 104 179 124 130

1996 102 131 151 110 225 105 164 98 99 176 124 135

1997 131 86 88 81

1998 106 132 137 110 134 100 201 127 101 168 121 148

1999 108 133 135 122 110 105 191 129 99 160 122 152

2000 127 155 161 129 134 122 217 147 190 187 144 176

2001 121 299 151 120 136 107 214 143 115 180 140 163

Area

Year 15 16 17 18 19 20 21 22 23 24 25 26 27 28 29 30 Average Maximum

1967 142 175 207 318

1968 186 131 172 189 285

1969 320c

1970 320c

1971 320c

1972 320c

1973 320c

1974 320c

1975 320c

1976 320c

1977 320c

1978 129 173 185 150 162 140 120 155 320

1979 121 150 145 155 160 115 117 152 130 120 125 140 191

1980 134 165 163 185 207 70 98 163 135 130 205 152 207

1981 138 155 168 174 163 75 126 175 150 140 152 218

1982 134 150 173 176 200 75 95 150 135 135 140 146 209

1983 124 135 151 158 206 125 163 149 140 185 148 206

1984 108 129 124 132 157 58 154 129 173 137 327

1985 115 123 131 127 103 153 143 347

1986 81 124 134 160 55 71 106 89 135 115 318
Document No. ORAUT-TKBS-0008-4 Revision No. 01 Effective Date: 05/27/2008 Page 31 of 77

Area

Year 15 16 17 18 19 20 21 22 23 24 25 26 27 28 29 30 Average Maximum

1987 135 151 174 246 204 68 96 129 139 155 246

1988 155 172 212 223 211 84 123 163 172 223

1989 162 160 168 178 189 78 116 146 118 150 217

1990 124 164 174 178 192 83 177 139 146 148 192

1991 122 165 175 173 189 81 79 143 154 110 155 144 194

1992 134 152 168 189 185 79 75 132 143 179 145 189

1993 145 177 195 192 213 91 96 158 169 195 167 213

1994 143 141 131 58 61 115 121 140 98 143

1995 109 129 137 149 153 58 70 117 121 127 212

1996 103 137 144 153 150 55 56 114 124 165 128 225

1997 102 150 153 83 63 127 135 170 114 170

1998 99 135 151 156 76 58 113 123 165 127 201

1999 108 137 153 151 75 57 111 126 170 126 191

2000 117 160 185 182 92 81 139 158 205 153 217

2001 112 167 159 131 90 74 135 145 193 147 299

a. Sources: Environmental Reports list in the References section.

b. Blank cells indicate data not available.

c. Ambient radiation not provided for these years - assumed equal to value for 1978.

Although not explicitly stated in the NTS annual environmental reports, the overall uncertainty of 

measurements from modern TLDs (e.g., Panasonic UD-814) has been determined to be less than 

20% (ORAUT 2007b). However, for earlier measurement methods, the overall uncertainty could have 

been greater. Therefore, for purposes of dose reconstructions that are favorable to claimants, the 

assumption is made that the 95th-percentile values of the expected values (50th percentile) in 

Tables 4-15 and 4-16 are enveloped by a factor of ±2 of the overall error of 50% (i.e., ±100%). 

Equation 4-1 (Section 4.2.1.2.1) would then estimate the GSD for the expected values at 1.52.

4.3.2

Ground seepage can increase when changes in ambient pressure pump small amounts of noble 

gases (primarily 85Kr and 133Xe) up through the overburden and into the atmosphere from the cavity 

that is created by a nuclear test. This process, sometimes referred to as atmospheric pumping, 

creates a diffuse source of radiological effluents.

Releases of Noble Gases

In 1982, REECo assumed responsibility for six noble gas sampling stations EPA previously ran and 

replaced them with new samplers. These sampling units are housed in a metal tool box with three 

metal air bottles attached with quick disconnect hoses. A vacuum is maintained on the first bottle, 

which causes a steady flow of air to be collected in the other two bottles. The flow rate is 

approximately 0.5 cm3

/min. The two collection bottles are exchanged weekly and yield a sample 

volume of about 3 × 105

 cm3 (Scoggins 1983).

The noble gases are separated and collected from the atmospheric sample by a series of 

cryogenic-gas chromatographic techniques. Water and carbon dioxide are removed at room 

temperature and the krypton and xenon are collected on charcoal at liquid nitrogen temperatures. 

These gases are transferred to a molecular sieve where they are separated from any remaining gases 

and each other. The krypton and xenon are transferred to separate scintillation vials and counted on 

a liquid scintillation counter. The lower limits of detection for krypton and xenon are 4 × 10-6 and 

10 × 10-6 pCi/m3

, respectively (Scoggins 1983).

Krypton-85

The original six samplers were in Areas 1, 5, 12, 15, 23, and 25. These six samplers showed no 

statistical difference in atmospheric concentrations of 85Kr during 1982 and 1983. The average 
Document No. ORAUT-TKBS-0008-4 Revision No. 01 Effective Date: 05/27/2008 Page 32 of 77

concentration of all stations for these years was 24.6 pCi/m3 (0.91 Bq/m3), which was established as 

NTS background. This background measurement shows good agreement with the global background 

of 27 pCi/m3 (1.0 Bq/m3

) (WMO 2002).

In 1983, REECo measured a statistically significant elevation in 85Kr concentrations in Area 20 of 

4.0 pCi/m3 (0.15 Bq/m3

) above background, which was subsequently related to seeps from nuclear 

cavities on Pahute Mesa (Scoggins 1984). These elevated concentrations continued until after the 

cessation of nuclear testing in 1992. Table 4-17 lists the average background and the Area 20 

elevated concentrations from 1983 through 1992.

Table 4-17. 85Kr average background and the Area 20 net concentrations and annual organ doses 

from 1983 through 1992.

a

Year

NTS 

background 

(pCi/m3)

Area 20 

Camp 

(pCi/m3)

Net 

elevated 

(pCi/m3)

Sourceb elevated 

concentration 

(pCi/m3)

Sourceb elevated 

concentration 

(Bq/m3)

Other organc

dosed at sourceb

(mrem/yr)

Skin dosee

at sourceb

(mrem/yr)

1983 27 46 19 502 18.6 0.0029 0.18

1984 25 29 4 106 3.9 0.0006 0.04

1985 28 39 11 290 10.8 0.0017 0.10

1986 30 58 28 739 27.4 0.0043 0.26

1987 26 39 13 343 12.7 0.0020 0.12

1988 25 29 4 106 3.9 0.0006 0.04

1989 23 27 4 106 3.9 0.0006 0.04

1990 29 37 8 211 7.8 0.0012 0.07

1991 24 32 8 211 7.8 0.0012 0.07

1992 26 30 4 106 3.9 0.0006 0.04

1993 27 28 1 26 1.0 0.0002 0.01

a. Sources: Environmental Reports list in the References section.

b. Source receptor location assumed to be 500 m from U-20a emplacement hole.

c. Includes gonads, breast, lung, red marrow, bone surfaces, thyroid, remainder, and effective whole body.

d. Assumes 2,000 hr/yr exposure and dose conversion factor of 2.2 × 10-16 Sv/Bq/s/m3 (Eckerman and Ryman 1993).

e. Assumes 2,000 hr/yr exposure and dose conversion factor of 1.32 × 10-14 Sv/Bq/s/m3 (Eckerman and Ryman 1993).

The source of the elevated 85Kr concentrations was attributed to the emplacement hole nearest to the 

Area 20 sampler (U-20a), which is about 3,660 m (12,000 ft) to the south. To determine whether 

these elevated concentrations could be of any consequence for unmonitored workers, upper-bound 

concentrations of 85Kr within 500 m of U-20a were estimated using Area 20 meteorological data 

station files and the CAP88-PC (EPA 1998) computer program. The program was used to develop 

atmospheric dispersion factors (X/Q) for various distances from the source of emissions. The ratio of 

the X/Q value at 500 m to that at 3,660 m was determined to be about 27, which means that the 85Kr 

concentration 500 m from the source could, on average, be 27 times greater than that measured at 

the sampler location. The potential dose resulting from submersion in the plume of this noble gas is 

considered to be entirely external because the gas is not readily assimilated by the body. Therefore, 

these source concentrations were converted to potential organ doses using submersion dose 

conversion factors from Federal Guidance Report No. 12 (Eckerman and Ryman 1993). 

As Table 4-17 shows, the upper-bound annual organ doses for all years are less than 1 mrem. For 

purposes of reconstructing environmental doses for unmonitored employees, these doses are 

inconsequential for determination of probability of causation and therefore should be ignored. In 

addition, if a monitored employee was exposed to the calculated upper-bound 85Kr concentrations or 

significantly higher concentrations during work activities, these exposures would have been monitored 

by personal dosimeters and would, therefore, already be present in the individual dosimetry record.

Uncertainty estimates provided for a number of the 85Kr measurements indicate, generally, a standard 

deviation of less than 20% of the measured value. Even at the 2-sigma level of ±40 percent, the 

annual organ doses in Table 4-17 are less than 1 mrem and therefore inconsequential for employee 

dose reconstruction.
Document No. ORAUT-TKBS-0008-4 Revision No. 01 Effective Date: 05/27/2008 Page 33 of 77

Xenon-133

Although the vast majority of the 133Xe measurements were below the lower limit of detection (i.e., 

10 × 10-6 pCi/m3

), elevated atmospheric concentrations of 133Xe were occasionally measured in 

Area 20. Like 85Kr, these elevated concentrations were attributed to seepage from the underground 

test cavities at the Pahute and Rainier Mesa tests (Fauver 1985). However, 133Xe has a short half-life 

(5.25 days versus 10.7 years for 85Kr). Therefore, these elevated concentrations were short lived. In 

addition, these elevated 133Xe concentrations were always less than the derived air concentrations 

(typically less than 3%) and would result in annual organ doses of less than 1 mrem. Therefore, 

potential doses to unmonitored employees from elevated concentrations of 133Xe are inconsequential 

for dose reconstruction.

4.4 INTERNAL INHALATION AND INGESTION DOSE TO UNDERGROUND WORKERS 

In the early 1980s, the Environmental Sciences Department of REECo recognized that the buildup of 

radon and radon daughter concentrations (RDCs) could pose a potential health problem in tunnels on 

Rainer Mesa and at other locations at the Nevada Test Site. In 1984, to determine the concentrations 

of the RDCs and the effect of environmental conditions on the buildup of these concentrations, 

REECo conducted radon measurement surveys in G-, T-, and N-Tunnels (Favor 1987). This section 

discusses the results of these surveys and those conducted in 1991 and 1992 (Lyons 1992a,b).

4.4.1

Area 12, which is in the Nuclear or High Explosive Test Zone, occupies 104 km2 (40 mi2

) at the 

northern boundary of NTS, known as Rainier Mesa. No atmospheric nuclear tests have occurred at 

this location; however, Area 12 was the site of the U.S. Atomic Energy Commission’s first fully 

contained underground nuclear detonation, named the Rainier test, on September 19, 1957, in a 

horizontal tunnel about 487 m (1,600 ft) into the mesa and 274 m (900 ft) beneath the top of the mesa 

(ORAUT 2007c). In the past several decades, a number of tunnels have been mined into Rainier 

Mesa, in which most of the U.S. Department of Defense (DOD) horizontal line-of-sight exposure 

experiments have occurred. The N-, P-, and T-tunnel complexes, in particular, were developed 

extensively during the 1970s and 1980s. The tunnel experiments usually involved complex 

construction of large-diameter (up to 9 m [27 ft]), line-of-sight pipes and special closure mechanisms, 

blast and gas seal doors, stemming plugs, and the like. The G-tunnel complex was originally 

established for nuclear testing purposes but, since 1971, has been used only as an underground 

research facility (ORAUT 2007c). 

Underground Activities

In addition to its use for nuclear testing purposes, N-tunnel was the location of a Nonproliferation 

Experiment involving 1.3 million kg (2.9 million lb) of conventional explosives on September 22, 1993. 

DOD operates a high-explosives research and development tunnel in Area 12. This reusable test bed 

supports programs involving the detonation of conventional or prototype high explosives and 

munitions (ORAUT 2007c). 

The U1a Complex in Area 1 is an underground laboratory of horizontal tunnels about 0.5 mi in length 

at the base of a vertical shaft about 960 ft beneath the surface. The vertical shaft is equipped with a 

mechanical hoist for personnel and equipment access. Another vertical shaft about 1,000 ft away 

provides cross ventilation, instrumentation, utility access, and emergency egress. The shaft was 

excavated in the 1960s, and a nuclear test was conducted in a horizontal tunnel mined from its base 

in 1990 (ORAUT 2007c).

4.4.2

Radon-222, with a radioactive half-life of 3.8 days, occurs in the 238U decay chain; 220Rn, with a 

radioactive half-life of 54.5 s, occurs in the 232Th chain. The decay daughters associated with 222Rn 

Radon Measurements
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include 218Po, 214Pb, 214Bi, and 214Po, and those associated with 220Rn include 216Po, 212Pb, 212Bi, and 

208Tl. Because radon is a noble gas and thus chemically inert, it migrates easily from the tunnel rock 

and soil, which contains naturally occurring trace quantities of uranium and thorium.

Radon progeny measurements in N- and T-Tunnels were made in drifts mined in a rock formation 

known as Tunnel Bed Non-Welded Ash Fall Tuff. This rock is unconsolidated and only slightly 

fractured. The drift that was sampled in the inclined G-Tunnel was mined in Grouse Canyon Welded 

Ash Fall Tuff, which is extremely fractured. Some factors that affect concentrations of radon and 

radon progeny in air are ventilation rates, barometric pressure, relative humidity, temperature 

inversions, the degree of fracturing in the rock, and the amount of smoke and dust in the air. Factors 

such as barometric pressure and fractures affect the rate at which radon emanates from the rock, 

while others such as dust and ventilation rates affect the accumulation rate of progeny in the air. 

Concentrations fluctuate with changing seasons, weather conditions, and activities in the area being 

monitored (Fauver 1985).

The concentration of radon progeny (the major dose contributors) in air is measured in working levels

(WL). This is the common unit for expressing radon progeny exposure rates. The WL was developed 

for use in uranium mines but is now also used for environmental exposures. Numerically, the WL is 

any combination of short-lived decay products in 1 L of air that results in the emission of 

1.3 × 105 MeV of potential alpha energy. When radon is in complete equilibrium with its short-lived 

decay products, 1 WL equals 100 pCi/L (i.e., 100 pCi/L each of 222Rn and short-lived decay products 

218Po, 214Pb, 214Bi, 214Po) (NCRP 1988, p. 17). For 220Rn and its decay products (216Po, 212Pb, 212Bi, 

and 208Tl), 1 WL is equal to 7.47 pCi/L. The advantage of the WL unit is that it allows comparison of 

different equilibrium levels and different concentrations of radon decay products. The degree of 

equilibrium is a critical factor for estimating inhalation exposure and is of equal importance to the 

radon concentration itself (NCRP 1998, p. 19). The WL unit considers this factor. However, because 

the dose conversion factors for 220Rn progeny (212Pb + 212Po) are one-third the values for the 222Rn 

progeny (ICRP 1981), the dose equivalent WL for 220Rn is only one-third that of a 222Rn WL.

The exposure of tunnel workers can be expressed in units of working-level months (WLM), which is 

an exposure rate of 1 working level (WL) for a working month of 170 hours (NCRP 1988, p. 17). For 

example, an exposure of 1 WLM would result from exposure to a concentration of 1 WL for 1 month or 

0.5 WL for 2 months. 

4.4.3

Measurements were taken in tunnels N, T, and G. The preliminary measurements indicated N- and 

T-Tunnel RDCs of about 0.01 WL under normal ventilation conditions. However, the data 

demonstrated that RDCs could rise to relatively high levels (i.e., 0.24 WL) when ventilation rates were 

significantly lower (Favor 1987, Figure 2, p. 7). The radon daughter concentrations in G-Tunnel were 

an order of magnitude higher than those in N- and T-Tunnels. The average RDC in the rock 

mechanics drift (the worst-case location in G-Tunnel) was 0.13 WL (ranging from 0.07 to 0.23 WL. 

Elevated RDCs in the rock mechanics drift of G-Tunnel seemed to be from a lower ventilation rate in 

conjunction with the more highly fractured nature of the welded tuff in which the incline drift was 

mined. By increasing the ventilation rate, a 60% reduction in RDCs from an average of 0.13 WL to an 

average of 0.05 WL was achieved (Table 4-18). 

Underground Radon Concentrations
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Table 4-18. Results of experiment to determine the effects of ventilation 

conditions on RDCs in the G-Tunnel inclined drift.

a

Location

Ventilation 

conditions

Radon daughter integrated 

sample average (WL)

Rock mechanics drift at 0+52 Alternating 0.19

Continuous 0.05

Average Alternating 0.13

Continuous 0.03

a. Source: Fauvor (1985, Table 3).

Lyons (1992a,b) reported additional radon measurements from 1991 and 1992. Although the 

G-Tunnel complex remained inactive during 1992, radon samples were taken to document potential 

radon WL in a worst-case scenario of complete ventilation failure throughout the complex. The 

maximum WL for radon daughters was 1.4. No samples were taken during 1992 in T-Tunnel, which 

was inactive that year.

Tables 4-19 and 4-20 list the results of extensive radon sampling in N- and P-Tunnel complexes in 

1992 (from Lyons 1992a, Table 3, p. 8; and Lyons 1992b, Table 1, p. 8, respectively). The average 

concentrations in N- and P-Tunnels were 0.021 and 0.009 WL, respectively, and the maximum 

concentrations in N- and P-Tunnel were 0.038 and 0.017 WL, respectively.

Table 4-19. Radon daughter concentrations for N-Tunnel in 1991 and 

1992. 

N-Tunnel location

222Rn WL 220Rn WL

Average Maximum Average Maximum

January–June 1992

Miner's Lunchroom 0.005 0.007 0.01 0.016

Raytheon Alcove 0.003 0.004 0.005 0.008

Slow Alcove 0.005 0.007 0.008 0.009

24 Bypass Drift 0.005 0.006 0.015 0.015

24 LOS Drift at GZ 0.005 0.007 0.012 0.014

22 Bypass 0.006 0.007 0.013 0.015

July–December 1991

Miner's Lunchroom 0.005 0.007 0.011 0.018

Raytheon Alcove 0.009 0.03 0.017 0.06

21 LOS at 2 + 50 0.034 0.059 0.029 0.046

15 Assembly Drift 0.006 0.009 0.015 0.021

Slow Alcove 0.004 0.008 0.01 0.025

23 Fast Alcove 0.009 0.014 0.018 0.03

24 Bypass Drift 0.005 0.007 0.012 0.041

24 LOS Drift 0.005 0.01 0.015 0.036

Average 0.008 0.013 0.014 0.025

4.4.4

Although measurements were periodically performed in the tunnel complexes to ensure adequate 

worker protection, neither the U.S. Department of Energy nor its predecessor agencies attempted to 

quantify or record occupational exposures to radon and its daughters. Therefore, dose reconstructors 

should adjust the dose for lung, ET1, and ET2 cancers of any employee who was a miner or tunnel 

worker to account for radon exposure while working in the tunnel complexes. To quantify the 

exposure, Table 4-21 lists favorable to claimant airborne 222Rn RDCs based on measurement results. 

It should be noted that the values provided in Table 4-21 pertain only to 222Rn and its daughter 

products. The 220Rn WL concentrations and resultant exposures for N- and P-Tunnels must be 

converted to annual alpha organ dose prior to entry into IREP.

Underground Workers’ Exposure to Radon
Document No. ORAUT-TKBS-0008-4 Revision No. 01 Effective Date: 05/27/2008 Page 36 of 77

Table 4-20. Radon daughter concentrations for P-Tunnel in 1991 and 

1992. 

P-Tunnel location

222Rn WL 220Rn WL

Average Maximum Average Maximum

January–June 1992

01 Drift at Access Drift 0.001 0.001 0.002 0.005

01 Fast Alcove 0.001 0.001 0.004 0.006

02 Main Drift at 6 + 00 0.001 0.001 0.002 0.004

04 Reentry 0.001 0.002 0.004 0.005

04 LOS at 12 + 00 0.002 0.002 0.005 0.006

HPD Base Station 0.001 0.001 0.003 0.006

Miner's Lunchroom 0.001 0.001 0.003 0.006

05 Cavity 0.001 0.001 0.003 0.004

July–December 1991

01 Drift at Access Drift 0.003 0.006 0.004 0.007

01 Fast Alcove 0.003 0.005 0.003 0.006

02 Main Drift at 6 + 00 0.01 0.032 0.01 0.046

04 LOS at VP X-Cut 0.008 0.013 0.012 0.015

04 LOS Drift at GZ 0.005 0.006 0.014 0.02

04 LOS Test Ch. 0.003 0.004 0.005 0.006

04 Bypass at RE#1 0.007 0.013 0.008 0.011

IHD Alcove 0.003 0.009 0.004 0.015

LLNL Alcove 0.003 0.006 0.007 0.02

05 Cavity 0.003 0.006 0.006 0.015

Average 0.003 0.006 0.006 0.011

Table 4-21. Annual 222Rn exposures and uncertainties for internal dose reconstruction 

for miners and tunnel workers (Fauver 1985). 

Tunnel complex

RDC concentration 

(WL)

Annual exposure

(WLM)a UncertaintyGSDb

G (before 1984) 0.16 1.92 1.52

G (1984 and later) 0.05 0.60 1.52

N 0.013 0.16 1.52

P 0.006 0.07 1.52

T 0.01 0.12 1.52

Unidentifiedb (before 1984) 0.16 1.92 1.52

Unidentifiedb (1984 and later) 0.05 0.60 1.52

a. Based on 170 hours of exposure per month for 12 months.

b. Use these values when underground work location is not known.

For G-Tunnel workers, annual exposures (assuming a full year of underground activity) are assumed 

to be 1.92 WLM before 1984 and 0.60 WLM during and after 1984. This is because REECo did not 

recognize until 1984 that significant reductions in RDCs could be effected by leaving all ventilation 

fans running overnight. Until 1984, the practice was to shut down alternating fans each night, which 

resulted in higher average RDCs.

Because no radon measurement data were readily available for the underground portions of the 

U1a Complex, the favorable to claimant assumption was made that the RDCs in U1a would be similar 

to those in G-Tunnel (which is mined in highly fractured tuff) [1]. In addition, a reasonable assumption 

was made that the U1a Complex is well ventilated, similar to G-Tunnel. Dose reconstructors should 

use these average, annual exposure values for miners and tunnel workers without identified tunnel 

locations.

For purposes of dose reconstruction, the GSD of the values in Table 4-21 can be estimated using 

Equation 4-1 (Section 4.2.1.2.1). Under the assumption that the 50th-percentile expected annual 
Document No. ORAUT-TKBS-0008-4 Revision No. 01 Effective Date: 05/27/2008 Page 37 of 77

exposures are those in Table 4-21 and that the 95th-percentile values are twice the values in 

Table 4-21, then the GSD for the values in Table 4-21 are 1.52 for all values. 

Gravel Gerties

As discussed in Chapter 2, Section 2.2.5 (ORAUT 2007c), underground bunkers were constructed in 

Area 5 for the purpose of testing containment capabilities involving accidental explosion of HE 

material in the underground bunker. Containment tests were conducted in the Gravel Gerties in 1957 

and again in 1982. These tests resulted in successful designs that were used at the Pantex Plant in 

Texas and for use in the Device Assembly Facility in Area 6. Five Gravel Gerties were constructed in 

Area 6 from 1988 to 1992. However, with the cessation of testing in 1992, these Gravel Gerties have 

not been activated for device assembly.

Because the Area 5 Gravel Gerties were constructed for testing purposes and the Area 6 Gravel 

Gerties were never activated, it is not likely that workers spent significant time inside the structures. 

However, if it can be determined from the dosimetry records or telephone interviews that a claimant 

worked in the Gravel Gerties for any significant time, it should be assumed that they would have been 

exposed to elevated levels of radon and, therefore, this radon exposure should be evaluated and 

included in the dose reconstruction.

Although radon measurements are not available inside the NTS Gravel Gerties, because of the 

similarities of construction, application of the radon concentrations measured inside the Pantex Gravel 

Gerties is assumed to be appropriate. Therefore, for a 12-month period that it can be determined that 

a claimant worked in the NTS Gravel Gerties, a radon exposure of 0.072 WLM should be assigned 

(ORAUT 2007a, Section 5.3.3). This exposure may be prorated as appropriate for employment 

periods of less than a full year. These exposures should be applied as lognormal with a GSD of 3 

(ORAUT 2007a, Section 5.3.3).

Uncertainty Associated with Radon Measurements

Two methods used to measure the concentration of radon and its progeny were the grab sample 

technique and the integration technique. Preliminary measurements were made by grab sampling for 

a general estimate of the concentrations. Lucas cells collected and counted grab samples of radon 

gas. The cells were evacuated in the laboratory and opened at the sampling location for a single 

intake of air. The filled cells were counted at the EPA Las Vegas Laboratory within 24 hours to 

minimize decay. The number of measurements was small due to limited availability of Lucas cells. 

Grab samples of radon daughters were collected and counted according to a technique described by 

Rolle (ANSI 1973), which utilized a single count of an air sample collected on a filter. Samples were

collected on a 2-cm Whatman fiberglass filter at a rate of approximately 5 L min-1 and counted on an 

EDA Instruments Incorporated Radon Detector (Model #RD200). The Rolle method allowed for a 

choice of several different analysis regimes. The regime chosen for tunnel measurements was a 

5-min sampling time, a 6-min decay time, and a 5-min counting time. Using appropriate correction 

factors, the RDC in WL can be evaluated 11 min after sample collection. The values obtained using 

the Rolle method were periodically checked using the Kusnetz (Favor 1987, p.3) method, which also 

utilized a single count of an air filter but allowed for a 40- to 90-min decay time. These measurement 

techniques have been found to have good accuracy and a relative standard deviation of less than 

15% (Favor 1987, p. 3). 

Grab sampling is quick and convenient. However, as mentioned above, radon concentrations can 

fluctuate widely with time and location. To account for this variability, integrating monitoring 

instruments were used. A passive environmental radon monitor (PERM) was used to measure radon 

in pCi L-1. A radon progeny integrating sampling unit (RPISU) measured the RDC in WL.

The PERM is an integrating radon monitor that employs electrostatic collection of radon daughter ions 

and uses a TLD as the radiation dose integration element. The lower limit of detection at the 95% 
Document No. ORAUT-TKBS-0008-4 Revision No. 01 Effective Date: 05/27/2008 Page 38 of 77

confidence level is approximately 0.3 pCi L-1 for a 1-week sample with a relative standard deviation of 

20% (Favor 1987, p. 4). The RPISU is a low-volume air sampler that draws air through a 0.65-μmpore-size Millipore membrane filter. A Teflon disc TLD containing dysprosium-activated calcium 

fluoride (Harshaw TLD200) is positioned close to the collection surface of the filter and serves as the 

dose integrating element. The lower limit of detection at the 95% confidence level is approximately 

0.0001 WL for a 1-week sampling period with a relative standard deviation of 10% (Favor 1987, p. 4). 

The RPISU technique is recommended for the measurement of RDCs by the Technical 

Measurements Center of the U.S. Department of Energy, Division of Remedial Action Projects (Favor 

1987, p. 4).

4.4.5

In addition to exposure to elevated levels of radon, underground workers may have been potentially 

exposed to other sources of airborne contamination. Because the breathing air in the underground 

environment was drawn from the outside environment, the underground workers would have been 

exposed to the same inhalation intakes as the aboveground workers (see Sections 4.2.1.2.2 through 

4.2.1.2.6 above). 

Underground Workers’ Exposure to Other Radionuclides

4.5 INSTRUCTION TO DOSE RECONSTRUCTORS FOR ASSIGNMENT OF 

ENVIRONMENTAL INTAKES 

4.5.1

Starting in April of 1957, all personnel entering the NTS were monitored for external radiation 

exposure (ORAUT 2007b). Because unexposed control badges were processed together with the 

personnel dosimeters and the readings from the control badges were subtracted from the dosimeter 

to obtain a net reading for determining exposure, ambient external dose should not be included in 

dose reconstruction after this time. Before the institution of universal badging in April 1957 for those 

employees who were not monitored, coworker doses should be assigned based on the values in 

ORAUT (2007b, Table 6-11) and no additional ambient dose should be assigned. Based on a review 

of the dosimetry literature of the site, no change in the handling of dosimeters occurred when 

universal badging was introduced; therefore, ambient dose is considered to be appropriately included 

in the coworker doses.

Assigning External Ambient Dose

External ambient doses for each year are available in ORAUT (2006) and can be used for workers 

who have continuous employment but nonexistent or incomplete monitoring records. For example, 

EG&G employees were likely stationed in Las Vegas and received dosimeters when they traveled to 

NTS. If the employee’s location cannot be clearly identified, it might be appropriate to assign ambient 

dose for intervals with no dosimeter exchanges. This would be favorable to the claimant, but the 

assignment of ambient dose is not a requirement. A second example includes REECo employees 

who were processed as new employees through NTS (Mercury) and assigned to work at Tonopah 

Test Range (TTR). These employees have a minimum of one NTS dosimeter. Often no additional 

records are provided for these employees by Sandia National Laboratory, the responsible 

organization for dosimetry at TTR. For these REECo employees assigned at TTR, ambient dose from 

NTS can be assigned as appropriate. 

4.5.2

With the exception of cases that can be worked using the bounding assumptions provided by ORAUTOTIB-0002 (ORAUT 2007d) or ORAUT-OTIB-0018 (ORAUT 2005), environmental inhalation and 

ingestion intakes presented in Tables 4-7 and 4-11, respectively, shall be applied for all cases. In 

addition, for applicable years of employment and affected organs, the annual doses provided in 

Tables 4-9 and 4-14 shall be applied, as a favorable to claimant assumption, as 30-250 keV photons 

Assigning Internal Ambient Dose
Document No. ORAUT-TKBS-0008-4 Revision No. 01 Effective Date: 05/27/2008 Page 39 of 77

to account for dose from inhalation and ingestion of short-lived fission and activation products. These 

intakes and resultant doses shall be entered into the IREP analysis with a constant distribution 

because they are considered to be reasonable overestimates of the actual intakes and doses. 

4.6 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] Rollins, Eugene M. Oak Ridge Associated Universities Team. Division Manager. October 

2007.

The highest radon concentrations were measured in the G-Tunnel and occurred in fractured 

tuff. Therefore, with the lack of measured data in the U1a Complex, an assumption favorable 

to claimants was made that the radon concentrations in the U1a Complex were the same as 

the highest measured concentrations in the G-Tunnel complex.
Document No. ORAUT-TKBS-0008-4 Revision No. 01 Effective Date: 05/27/2008 Page 40 of 77

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Wruble, D. T., and E. M. McDowell, 1990a, Nevada Test Site Annual Site Environmental Report – 

1989 Volume I, DOE/NV/10630-11 Volume I, U.S. Department of Energy, Nevada Operations 

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Wruble, D. T., and E. M. McDowell, 1990b, Nevada Test Site Annual Site Environmental Report – 

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Processes,” Health Physics, volume 82, number 5, pp. 669-679, Lippincott Williams and 

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Tests 1961 Through 1992, DOE/NV-317 (Rev. 1) UC-702, Nevada Operations Office, Las 

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Tunnels, DOE/NV/10327-31, U.S. Department of Energy, Nevada Operations Office, Las 

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Hicks, H. G., 1981a, Results of Calculations of External Gamma Radiation Exposure Rates from 

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Fallout and the Related Radionuclide Compositions, Operation Teapot, 1955, UCRL-53152-Pt. 

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Hicks, H. G., 1981c, Results of Calculations of External Gamma Radiation Exposure Rates from 

Fallout and the Related Radionuclide Compositions, Operations Nougat through Bowline, 

1962–1968, UCRL-53152-Pt. 8, University of California, Lawrence Livermore National 

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Fallout and the Related Radionuclide Compositions, Operation Ranger, 1951, UCRL-53152-

Pt. 1, University of California, Lawrence Livermore National Laboratory, Livermore, California, 

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Hicks, H. G., 1981e, Calculation of the Concentration of any Radionuclide Deposited on the Ground 

by Offsite Fallout from a Nuclear Detonation, UCRL-86177, University of California, Lawrence 

Livermore Laboratory, Livermore, California, June. [SRDB Ref ID: 1824]

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and the Related Radionuclide Compositions of Selected U.S. Pacific Events, UCRL-53505, 

Lawrence Livermore National Laboratory/Nevada Test Site, Livermore, California, February.

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Daughters by Workers, Publication 32, Pergamon Press, Oxford, England.

ICRP (International Commission on Radiological Protection), 1995, “Publication 68: Dose Coefficients 

for Intakes of Radionuclides by Workers,” Annals of the ICRP Volume 24/4, July. 

ICRP (International Commission on Radiological Protection), 1998, Individual Monitoring for Internal 

Exposure of Workers, Publication 78, Pergamon Press, Oxford, England.

Lyons, C. L., 1992a, Operational Air Sampling Report - January 1 through June 30, 1992, 

DOE/NV/10630-42, U.S. Department of Energy, Nevada Operations Office, Las Vegas, 

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Lyons, C. L., 1992b, Operational Air Sampling Report - July through December, 1991, 

DOE/NV/10630-37, U.S. Department of Energy, Nevada Operations Office, Las Vegas, 

Nevada. [SRDB Ref ID: 8583]

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DOE/NV/10162-14, U.S. Department of Energy, Nevada Operations Office, Las Vegas, 

Nevada. [SRDB Ref ID: 8493]

McArthur, S. W., 1985, Radionuclide Inventory and Distribution Program: Report #2, Areas 2 and 4, 

DOE/NV/10162-20, U.S. Department of Energy, Nevada Operations Office, Las Vegas, 

Nevada. [SRDB Ref ID: 7856]

McArthur, S. W., 1987, Radionuclide Inventory and Distribution Program: Report #3, Areas 3, 7, 8, 9, 

and 10, DOE/NV/10384-15, U.S. Department of Energy, Nevada Operations Office, Las 

Vegas, Nevada. [SRDB Ref ID: 7860]

McArthur, S. W., 1988, Radionuclide Inventory and Distribution Program: Report #4, Areas 18 and 

19, DOE/NV/10384-22, U.S. Department of Energy, Nevada Operations Office, Las Vegas, 

Nevada. [SRDB Ref ID: 7854]

McArthur, S. W., 1989, Radionuclide Inventory and Distribution Program: Report #5, Areas 5, 11, 12, 

15, 17, 18, 19, 25, 26, and 30, DOE/NV/10384-26, U.S. Department of Energy, Nevada 

Operations Office, Las Vegas, Nevada. [SRDB Ref ID: 7865]

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7872]

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

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Support, Cincinnati, Ohio, November. [SRDB Ref ID: 35987]

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Document No. ORAUT-TKBS-0008-4 Revision No. 01 Effective Date: 05/27/2008 Page 47 of 77

TABLE OF CONTENTS

SECTION TITLE

A.1 Introduction ............................................................................................................................

PAGE

48

A.2 Air Sampling Data and Intake Estimates ................................................................................49

A.3 Soil Contamination and Decay Corrections ............................................................................52

A.4 Scaling Factors for Inhalation Intake Estimates......................................................................54

A.5 Correction for Resuspension for Early Times After Atmospheric Tests................................... 54

A.6 Corrections for Inhalation Dose From Short-lived Fission and Activation Products................. 57

A.7 Ingestion Pathway..................................................................................................................65

A.8 Instruction to Dose Reconstructors for Assignment of Environmental Intakes ........................ 73

LIST OF TABLES

TABLE TITLE

A-1 Atmospheric concentrations of 239Pu for sampled areas.........................................................

PAGE

50

A-2 Annual inhalation intakes from 239Pu for sampled areas ......................................................... 51

A-3 Inventory of contaminated soil................................................................................................ 53

A-4 Radionuclide areal soil deposition.......................................................................................... 53

A-5 Radionuclide areal soil deposition decay-corrected to 1963................................................... 54

A-6 Abundance of radionuclides in the NTS soils relative to 239Pu decay-corrected to 1963......... 55

A-7 Scaled inhalation intakes corrected for early resuspension in 1963........................................ 56

A-8 Far-field refractory enrichment ratios...................................................................................... 58

A-9 Organ-specific inhalation dose fission and activation product correction factors..................... 61

A-10 Inhalation dose from short-lived fission and activation products ............................................. 65

A-11 Radionuclide soil concentration by area at the NTS ............................................................... 66

A-12 Area-specific and maximum annual ingestion rates................................................................ 66

A-13 Annual organ doses from ingestion of 238,239Pu and 241Am...................................................... 68

A-14 Organ-specific ingestion fission and activation correction factors........................................... 72

A-15 Organ-specific annual ingestion doses for the 90Sr intake of 55.42 Bq/yr ............................... 72

A-16 Organ specific doses from ingestion of fission and activation products .................................. 73

ATTACHMENT A

AMBIENT ENVIRONMENTAL INTAKES AT THE NEVADA TEST SITE

BASED ON AIR S AMPLING AND SOIL CONTAMINATION DATA 

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LIST OF FIGURES

FIGURE TITLE

A-1 Resuspension factor as a function of time after initial deposition............................................

PAGE

56

A-2 Small Boy – 90Sr fraction of total dose lungs........................................................................... 59

A-3 Little Feller I – 90Sr fraction of total dose to lungs ................................................................... 60

A-4 Turk – 90Sr fraction of total dose to lungs ............................................................................... 60

A-5 Strontium-90 fraction of total dose.......................................................................................... 61

A-6 Strontium-90 fraction of total dose.......................................................................................... 62

A-7 Strontium-90 fraction of total dose.......................................................................................... 62

A-8 Strontium-90 fraction of total dose.......................................................................................... 63

A-9 Strontium-90 fraction of total dose to the liver ........................................................................ 63

A-10 Strontium-90 fraction of total dose red marrow....................................................................... 64

A-11 Strontium-90 fraction of total dose to bone surface ................................................................ 64

A-12 Strontium-90 fraction of total ingestion dose .......................................................................... 69

A-13 Strontium-90 fraction of total ingestion dose to red bone marrow........................................... 69

A-14 Strontium-90 fraction of total ingestion dose to red bone marrow........................................... 70

A-15 Strontium-90 fraction of total ingestion dose .......................................................................... 70

A-16 Strontium-90 fraction of ingestion dose to the upper large intestine ....................................... 71

A-17 Strontium-90 fraction of total ingestion dose .......................................................................... 71

A.1 Introduction

With the cessation of atmospheric testing at NTS in 1962, the greatest potential for environmental 

intakes of radioactive material results from the inhalation of radioactive particles that were 

resuspended from NTS soils into the atmosphere and from ingestion of soils previously contaminated 

by atmospheric nuclear weapons tests, reactor tests, and safety tests. The potential inhalation 

intakes can be estimated from air sampling data in the NTS annual environmental reports (see 

References section) coupled with extensive soil contamination data from between 1983 and 1991 

McArthur 1983, 1985, 1987, 1988, 1989, 1991). Because the air monitoring data was limited to gross 

alpha and beta measurements, tritium, and isotopes of plutonium (e.g., 238Pu, 239Pu, and 240Pu), 

inhalation intakes of other relatively long-lived radionuclides that have been identified in the NTS soils 

(e.g., 241Am, 60Co, 137Cs, 90Sr, 152Eu, 154Eu, and 155Eu) are scaled to those of plutonium based on their 

relative abundance in NTS soils. Ingestion intakes can also be estimated by assuming consumption 

of contaminated NTS soils. To ensure that both inhalation and ingestion intakes are not 

underestimated, the relative abundances of the long-lived radionuclides in NTS soils that were 

determined from the 1991 soil contamination data (McArthur 1991) were decay-corrected to 1963. In 

addition, to ensure that intakes and resultant doses are not underestimated, correction factors are 

developed to account for potential exposures to short-lived fission and activation products based on 

test-specific data from Hicks (1981a,b,c,d). In addition, a correction factor is developed for inhalation 

intakes which accounts for the phenomenon of early resuspension (Anspaugh 2002).

This approach has been developed to address the unique character of NTS, which is a large outdoor 

testing facility where potential exposure to radioactive materials was primarily based on residual 

contamination from atmospheric testing. For most of the employees of the primary contractor, 

ATTACHMENT A

AMBIENT ENVIRONMENTAL INTAKES AT THE NEVADA TEST SITE

BASED ON AIR S AMPLING AND SOIL CONTAMINATION DATA 

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occupational internal doses can be equated to the doses they potentially received from the 

resuspension of residual radioactivity from the atmospheric testing (i.e., environmental internal dose). 

For workers who were primarily indoors (cafeterias, administrative facilities, and maintenance shops) 

this dose likely represents an overestimate of internal exposure. The environmental internal dose is 

the occupational internal dose, regardless of the recorded external dose. If an internal exposure was 

suspected, bioassay was performed. Managing radioactive material in the form of devices was 

episodic and limited to a few workers (e.g., radiation safety and industrial hygiene personnel, miners, 

and experimenters). These workers are identified on the rosters that were published before the event, 

and these workers likely have bioassay results included in the DOE records.

Section A.2 presents air sampling data from the NTS annual environmental reports from 1971 through 

2001 and the calculated annual inhalation 239Pu intakes. Section A.3 presents an area-specific 

summary of NTS soil contamination data that was gathered and reduced in the late 1980s (McArthur 

1991). This section also presents soil contamination data that is decay-corrected to 1963. Section 

A.4 describes the methods that were used to develop inhalation intake scaling factors for 

radionuclides identified in NTS soils in comparison with intakes of plutonium that were calculated from 

air sampling data. Section A.5 describes the method that was used to account for higher air 

concentrations due to resuspension for 1963 (the first year after suspension of atmospheric testing). 

Section A.6 describes the methods that were used to develop the correction factors that were used to 

account for organ dose from inhalation of short-lived radionuclides that no longer persist in the NTS 

environment in measurable quantities. Section A.7 presents the methodology that was used to 

calculate ingestion intakes from oral consumption of contaminated NTS soils. In addition, Section A.7 

provides the methodology that was used to account for ingestion dose from short-lived fission and 

activation products. Section A.8 includes step-by-step instructions to the dose reconstructor for 

application of ambient environmental inhalation and ingestion intakes. Attachment B provides a 

discussion that includes the results of 30 years of inhalation and ingestion intakes in Sections A.6 and 

A.7. 

A.2 Air Sampling Data and Intake Estimates 

Routine isotopic atmospheric measurements of plutonium at NTS began in 1971 with samplers in 

15 locations across the site. Six additional sampling stations were added in 1978. Equipment at fixed 

locations continuously sampled the ambient air to monitor for radioactive materials. These locations 

were chosen to provide representative samples from the populated areas on the site as well as to 

monitor resuspension of low-fired plutonium spread by safety experiments before 1960 in Areas 2, 3, 

4, 7, 9, and 10. Access, worker population, geographical coverage, presence of radioactivity, and 

availability of electric power were considerations in the site selection for air samplers (Black and 

Townsend 1997).

In 1988, efforts to monitor radioactive air emissions at NTS were increased as a result of the 

requirements of DOE Order 5400.1. Known and potential effluent sources throughout NTS were 

assessed for their potential to contribute to public dose (Black and Townsend 1997).

The 239Pu concentrations in Table A-1 for 1989 through 2001 represent the average of the maximum 

concentrations for a given area in a given year. In cases where maximum values were not provided 

(i.e., 1971 through 1988), the average of the average concentrations was reported. In addition to the 

annual area concentrations, Table A-1 also provides NTS site average and maximum concentrations, 

which represent the arithmetic averages of the concentration of all the areas and the maximum of the 

ATTACHMENT A

AMBIENT ENVIRONMENTAL INTAKES AT THE NEVADA TEST SITE

BASED ON AIR S AMPLING AND SOIL CONTAMINATION DATA 

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area maximum or area averages of all the areas, respectively. Potential intakes associated with these 

concentrations can be calculated under the assumption that an unmonitored worker was 

occupationally exposed for 2,000 hr/yr with a breathing rate of 2,400 m3

/yr. Table A-2 presents these 

calculated intakes.

Table A-1. Atmospheric concentrations of 239Pu for sampled areas (pCi/m3

).a

Year

Area

1 2 3 4 5 6 7 9 10 11 12

1971 3.7E-04 6.6E-04 1.7E-04 1.2E-04 1.7E-04 7.2E-04 2.1E-04 9.1E-05 1.6E-04

1972 1.5E-04 2.2E-04 3.7E-04 1.4E-04 1.3E-03 4.3E-03 2.9E-04 2.4E-04 7.9E-04

1973 8.7E-05 2.2E-04 2.1E-04 8.7E-05 1.2E-04 8.6E-04 4.4E-05 2.4E-04 4.3E-05

1974 8.0E-05 6.8E-05 1.3E-04 5.0E-05 7.9E-05 2.1E-04 5.7E-05 8.0E-05 5.8E-05

1975 2.9E-05 3.8E-05 1.5E-04 4.3E-05 6.3E-05 1.7E-04 5.7E-05 4.9E-05 3.3E-05

1976 2.9E-04 1.7E-04 1.0E-03 1.5E-04 4.4E-04 3.2E-03 2.1E-04 2.4E-04 2.8E-04

1977 2.7E-05 6.5E-05 2.6E-04 3.4E-05 5.5E-05 2.5E-04 4.6E-05 1.2E-04 2.1E-05

1978 4.7E-05 1.3E-04 2.8E-04 6.1E-05 9.9E-05 5.4E-04 2.9E-04 1.1E-04 6.2E-05

1979 7.4E-05 4.7E-04 1.7E-06 2.1E-05 4.0E-05 4.9E-05 5.2E-04 2.9E-04 3.7E-05 2.4E-05

1980 2.3E-05 2.6E-04 1.1E-06 2.9E-05 4.8E-05 4.5E-05 4.5E-04 4.4E-05 1.9E-05

1981 2.6E-05 6.4E-05 1.4E-04 2.5E-05 2.9E-05 3.3E-05 3.2E-04 4.6E-05 2.0E-05

1982 6.1E-05 4.0E-05 6.2E-05 2.0E-05 3.5E-05 7.4E-05 2.2E-04 5.3E-05 2.7E-05

1983 3.1E-05 6.5E-05 9.2E-05 2.3E-05 3.2E-05 3.3E-05 2.1E-04 1.8E-04 2.3E-05

1984 1.9E-05 1.0E-04 2.3E-04 3.8E-05 2.0E-05 3.8E-05 1.0E-03 5.6E-05 1.7E-05

1985 1.5E-05 5.8E-05 2.1E-04 3.1E-05 2.7E-05 2.7E-05 1.3E-03 5.2E-05 2.3E-05

1986 1.5E-04 6.2E-05 5.5E-04 2.8E-05 1.9E-04 4.8E-05 2.8E-04 2.5E-05 3.2E-05

1987 5.5E-05 2.3E-05 2.8E-03 1.7E-05 2.5E-05 1.6E-05 1.1E-04 3.2E-05 1.6E-05

1988 9.2E-05 2.8E-05 2.1E-04 2.8E-05 2.2E-05 1.5E-05 5.1E-05 1.2E-05 1.6E-05

1989 8.6E-04 5.8E-05 5.4E-04 5.0E-05 8.0E-05 2.7E-04 3.5E-04 5.4E-04 6.8E-04

1990 9.9E-05 3.2E-05 3.5E-04 3.1E-05 3.7E-05 4.1E-05 5.0E-04 4.3E-05 1.1E-04 1.0E-05

1991 3.1E-03 2.1E-05 3.8E-04 3.5E-05 7.2E-05 3.2E-05 3.5E-04 5.4E-07 2.2E-04 1.4E-05

1992 1.5E-04 1.1E-04 2.3E-03 8.0E-05 7.3E-05 3.9E-04 8.8E-04 2.4E-04 8.8E-05 3.8E-05

1993 2.0E-04 4.9E-05 4.0E-04 8.0E-05 9.0E-05 7.8E-05 2.7E-03 1.7E-04 4.2E-04 8.4E-04

1994

1995 4.1E-05 1.1E-05 2.4E-04 1.5E-05 5.0E-05 1.2E-05 3.9E-04 3.8E-05 4.0E-05 1.1E-05

1996 7.5E-04 3.0E-04 1.6E-05 4.6E-05 4.5E-04 6.1E-04 5.8E-05 1.9E-05 3.7E-06

1997 1.1E-05 2.6E-05 1.1E-05 1.5E-06

1998 4.7E-04 6.7E-05 9.5E-05 1.1E-05 2.7E-05 7.4E-04 9.5E-06

1999 5.9E-05

2000 8.5E-05 2.7E-04 5.9E-05 2.1E-05 4.1E-04 3.4E-05 2.8E-03

2001 4.5E-04 1.2E-04 1.2E-04 1.9E-04 6.8E-05 9.6E-05 1.7E-05 5.0E-04 8.5E-06

Year

Area Site 

averageb

Site 

maximumc 15 16 18 19 20 23 25 27 28

1971 1.9E-04 8.0E-05 2.3E-04 9.4E-05 8.4E-05 3.9E-05 2.26E-04 7.21E-04

1972 1.7E-04 1.5E-04 2.7E-03 3.5E-04 1.7E-04 5.1E-05 7.63E-04 4.29E-03

1973 4.4E-05 3.8E-05 7.2E-05 5.6E-05 3.7E-05 3.3E-05 1.46E-04 8.59E-04

1974 6.0E-05 6.6E-05 8.5E-05 8.6E-05 7.2E-05 1.2E-04 8.64E-05 2.11E-04

1975 4.3E-05 3.4E-05 4.4E-05 5.7E-05 6.8E-05 7.0E-05 6.27E-05 1.68E-04

1976 1.6E-04 9.9E-05 5.7E-04 2.0E-04 1.4E-04 6.3E-05 4.82E-04 3.18E-03

1977 2.6E-05 2.3E-05 2.7E-05 3.3E-05 3.1E-05 2.3E-05 6.93E-05 2.55E-04

1978 5.1E-05 6.2E-05 5.8E-05 7.0E-05 9.4E-05 3.9E-05 1.33E-04 5.42E-04

1979 5.2E-05 2.1E-05 2.0E-05 2.2E-05 1.8E-05 1.6E-05 2.3E-05 9.98E-05 5.24E-04

1980 6.4E-05 9.5E-05 3.5E-05 3.1E-05 2.1E-05 1.6E-05 7.88E-05 4.52E-04

1981 9.3E-05 2.6E-05 2.3E-05 2.9E-05 2.4E-05 2.5E-05 1.6E-05 5.88E-05 3.22E-04

1982 4.2E-05 1.7E-05 2.2E-05 2.1E-05 1.7E-05 5.04E-05 2.15E-04

1983 2.2E-05 1.5E-05 3.0E-05 2.8E-05 3.0E-05 5.85E-05 2.14E-04

1984 3.2E-05 2.8E-05 2.5E-05 2.9E-05 3.2E-05 4.8E-05 1.15E-04 1.02E-03

1985 4.0E-05 2.1E-05 3.5E-05 2.2E-05 2.6E-05 3.1E-05 1.30E-04 1.33E-03

1986 3.9E-05 2.8E-05 2.2E-05 6.6E-05 2.1E-05 2.3E-05 1.03E-04 5.45E-04

1987 1.8E-05 1.4E-05 1.7E-05 1.6E-05 2.3E-05 2.13E-04 2.81E-03

1988 2.4E-05 1.4E-05 3.5E-05 1.5E-05 1.5E-05 4.15E-05 2.14E-04

1989 2.3E-05 1.1E-05 9.0E-06 1.2E-04 9.5E-06 4.8E-06 2.40E-04 8.55E-04

1990 8.2E-05 5.8E-06 3.9E-06 8.5E-05 1.1E-05 3.5E-05 9.17E-05 4.98E-04

ATTACHMENT A

AMBIENT ENVIRONMENTAL INTAKES AT THE NEVADA TEST SITE

BASED ON AIR S AMPLING AND SOIL CONTAMINATION DATA 

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Year

Area

1 2 3 4 5 6 7 9 10 11 12

1991 1.2E-04 2.9E-05 8.8E-06 1.3E-05 2.6E-05 7.7E-06 2.77E-04 3.10E-03

1992 1.1E-03 1.9E-05 8.9E-06 1.3E-05 1.8E-05 1.6E-04 3.52E-04 2.26E-03

1993 3.3E-04 1.3E-05 7.7E-06 6.3E-06 3.4E-06 1.1E-05 3.37E-04 2.70E-03

1994

1995 1.6E-04 1.4E-05 1.0E-06 7.4E-06 4.5E-06 6.92E-05 3.90E-04

1996 2.4E-04 1.8E-06 3.5E-06 4.5E-06 6.3E-06 4.2E-06 2.6E-06 1.57E-04 7.50E-04

Year

Area Site 

averageb

Site 

maximumc 15 16 18 19 20 23 25 27 28

1997 2.9E-05 1.9E-07 2.0E-05 2.0E-05 6.0E-07 1.0E-06 1.6E-06 1.11E-05 2.90E-05

1998 6.3E-05 3.1E-05 3.3E-05 1.2E-05 6.3E-06 1.30E-04 7.35E-04

1999 1.1E-05 1.1E-05 2.0E-05 6.2E-06 2.14E-05 5.90E-05

2000 2.6E-04 1.0E-05 6.6E-06 1.1E-05 3.63E-04 2.83E-03

2001 1.4E-05 9.6E-06 1.7E-05 7.6E-06 2.0E-06 4.4E-06 1.02E-04 5.04E-04

a. Source: NTS Environmental Reports in References section.

b. Values represent the arithmetic average of the area average concentrations for years 1971 through 1988 and the arithmetic average of 

the area maximum concentrations for years 1989 through 2001.

c. Values represent the maximum of the average area concentrations for years 1971 through 1988 and the maximum of the maximum 

area concentrations for years 1989 through 2001.

Table A-2. Annual inhalation intakes from 239Pu for sampled areas (Bq). 

Year

Area

1 2 3 4 5 6 7 9 10 11 12

1971 0.0329 0.0587 0.0147 0.0105 0.0153 0.0640 0.0185 0.0081 0.0145

1972 0.0131 0.0199 0.0327 0.0124 0.1190 0.3810 0.0258 0.0216 0.0698

1973 0.0077 0.0198 0.0183 0.0078 0.0107 0.0763 0.0039 0.0216 0.0038

1974 0.0071 0.0060 0.0112 0.0044 0.0070 0.0187 0.0051 0.0071 0.0052

1975 0.0026 0.0034 0.0130 0.0038 0.0056 0.0149 0.0051 0.0044 0.0029

1976 0.0255 0.0153 0.0924 0.0133 0.0392 0.2824 0.0189 0.0210 0.0248

1977 0.0024 0.0058 0.0226 0.0030 0.0049 0.0225 0.0041 0.0107 0.0019

1978 0.0042 0.0118 0.0250 0.0054 0.0088 0.0481 0.0258 0.0095 0.0055

1979 0.0066 0.0413 0.0001 0.0018 0.0035 0.0044 0.0465 0.0258 0.0033 0.0021

1980 0.0020 0.0230 0.0001 0.0026 0.0043 0.0040 0.0401 0.0039 0.0017

1981 0.0023 0.0056 0.0124 0.0022 0.0026 0.0029 0.0286 0.0041 0.0018

1982 0.0054 0.0036 0.0055 0.0018 0.0031 0.0066 0.0191 0.0047 0.0024

1983 0.0028 0.0057 0.0082 0.0020 0.0029 0.0029 0.0190 0.0162 0.0020

1984 0.0017 0.0088 0.0202 0.0034 0.0018 0.0034 0.0906 0.0050 0.0015

1985 0.0013 0.0051 0.0189 0.0027 0.0024 0.0024 0.1181 0.0046 0.0020

1986 0.0131 0.0055 0.0484 0.0025 0.0165 0.0043 0.0249 0.0022 0.0028

1987 0.0048 0.0020 0.2495 0.0015 0.0022 0.0014 0.0098 0.0028 0.0014

1988 0.0082 0.0025 0.0190 0.0025 0.0020 0.0013 0.0045 0.0011 0.0014

1989 0.0759 0.0052 0.0478 0.0045 0.0071 0.0240 0.0311 0.0480 0.0604

1990 0.0088 0.0029 0.0308 0.0027 0.0033 0.0036 0.0442 0.0038 0.0095 0.0009

1991 0.2753 0.0018 0.0340 0.0031 0.0064 0.0029 0.0311 0.0000 0.0199 0.0013

1992 0.0133 0.0098 0.2007 0.0071 0.0065 0.0346 0.0781 0.0213 0.0078 0.0034

1993 0.0180 0.0044 0.0352 0.0071 0.0079 0.0069 0.2398 0.0151 0.0373 0.0746

1994

1995 0.0036 0.0010 0.0216 0.0014 0.0045 0.0011 0.0346 0.0033 0.0036 0.0010

1996 0.0666 0.0000 0.0268 0.0015 0.0041 0.0400 0.0542 0.0051 0.0017 0.0003

1997 0.0010 0.0023 0.0010 0.0001

1998 0.0416 0.0060 0.0084 0.0010 0.0024 0.0653 0.0008

1999

2000 0.0075 0.0239 0.0052 0.0018 0.0366 0.0030 0.2509

2001 0.0404 0.0011 0.0103 0.0166 0.0061 0.0086 0.0015 0.0447 0.0008

Year

Area Site 

average

Site 

15 16 18 19 20 23 25 27 28 maximum

1971 0.0171 0.0071 0.0201 0.0083 0.0075 0.0035 0.0200 0.0640

1972 0.0150 0.0129 0.2424 0.0309 0.0147 0.0045 0.0677 0.3810

1973 0.0039 0.0034 0.0063 0.0050 0.0033 0.0029 0.0130 0.0763

1974 0.0053 0.0059 0.0075 0.0076 0.0064 0.0104 0.0077 0.0187

ATTACHMENT A

AMBIENT ENVIRONMENTAL INTAKES AT THE NEVADA TEST SITE

BASED ON AIR S AMPLING AND SOIL CONTAMINATION DATA 

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1975 0.0038 0.0030 0.0039 0.0051 0.0060 0.0062 0.0056 0.0149

1976 0.0139 0.0088 0.0504 0.0178 0.0122 0.0056 0.0428 0.2824

1977 0.0023 0.0020 0.0024 0.0029 0.0028 0.0020 0.0062 0.0226

1978 0.0045 0.0055 0.0052 0.0062 0.0083 0.0035 0.0118 0.0481

1979 0.0046 0.0019 0.0017 0.0020 0.0016 0.0014 0.0020 0.0089 0.0465

1980 0.0057 0.0085 0.0031 0.0027 0.0019 0.0014 0.0070 0.0401

1981 0.0083 0.0023 0.0020 0.0025 0.0022 0.0022 0.0014 0.0052 0.0286

Year

Area Site 

average

Site 

15 16 18 19 20 23 25 27 28 maximum

1982 0.0037 0.0015 0.0019 0.0019 0.0015 0.0045 0.0191

1983 0.0020 0.0013 0.0026 0.0025 0.0027 0.0052 0.0190

1984 0.0028 0.0025 0.0022 0.0025 0.0028 0.0043 0.0102 0.0906

1985 0.0036 0.0019 0.0031 0.0020 0.0023 0.0028 0.0115 0.1181

1986 0.0035 0.0025 0.0020 0.0058 0.0018 0.0020 0.0092 0.0484

1987 0.0016 0.0012 0.0015 0.0014 0.0020 0.0000 0.0189 0.2495

1988 0.0022 0.0012 0.0031 0.0013 0.0013 0.0037 0.0190

1989 0.0021 0.0010 0.0008 0.0104 0.0008 0.0004 0.0213 0.0759

1990 0.0073 0.0005 0.0003 0.0075 0.0010 0.0031 0.0081 0.0442

1991 0.0109 0.0026 0.0008 0.0012 0.0023 0.0007 0.0246 0.2753

1992 0.0977 0.0017 0.0008 0.0011 0.0016 0.0142 0.0312 0.2007

1993 0.0293 0.0012 0.0007 0.0006 0.0003 0.0010 0.0300 0.2398

1994 0.0000 0.0000

1995 0.0142 0.0012 0.0001 0.0007 0.0004 0.0061 0.0346

1996 0.0213 0.0002 0.0003 0.0000 0.0004 0.0006 0.0004 0.0002 0.0124 0.0666

1997 0.0026 0.0000 0.0018 0.0018 0.0001 0.0001 0.0001 0.0010 0.0026

1998 0.0056 0.0027 0.0029 0.0010 0.0006 0.0115 0.0653

1999 0.0010 0.0010 0.0018 0.0006 0.0011 0.0018

2000 0.0229 0.0009 0.0006 0.0010 0.0322 0.2509

2001 0.0012 0.0008 0.0015 0.0007 0.0002 0.0004 0.0090 0.0447

It should be noted that some covered employees remained on the site continuously for weeks at a 

time. However, because most of the nonworking hours were spent indoors where ambient air 

particulate loadings would be much less than the outdoor loadings and because of the conservative 

assumptions that were used to estimate the values in Table A-2, adjustment of the tabular data is not 

required to ensure intakes are not underestimated for these individuals. In addition, employees who 

lived on the site during their work week would have been housed in Area 12 or Area 23 (Mercury). 

For most years the values in Table A-1 for these locations are less than the site average values.

It is assumed that plutonium could be any of absorption Type S, Super S, or M dependent on which

type delivers the maximum organ dose. 

A.3 Soil Contamination and Decay Corrections

Extensive studies were performed in the 1980s to quantify residual contamination at NTS (McArthur 

1983, 1985, 1987, 1988, 1989, 1991). Table A-3 lists the results of these studies (McArthur 1991). 

Table A-4 lists the total areal depositions based on the inventory values in Table A-3 divided by the 

respective areal size. It should be noted that the results shown in Table A-4 are representative of 

areas of the NTS that have been shown to contain measurable levels of contamination. These area 

actually represent only about one third of the total area within the boundaries of the NTS.

Because the data shown in Table A-3 is representative of the soil contamination in 1991, these values 

were decay-corrected to the beginning of 1963 – the first year after the cessation of atmospheric 

testing. Table A-5 presents the decay-corrected areal soil deposition.

ATTACHMENT A

AMBIENT ENVIRONMENTAL INTAKES AT THE NEVADA TEST SITE

BASED ON AIR S AMPLING AND SOIL CONTAMINATION DATA 

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Table A-3. Inventory of contaminated soil (Ci). 

Area Area (mi2) Am-241 Pu-238 Pu-239,240 Co-60 Cs 137 Sr-90 Eu-152 Eu-154 Eu-155

1 26.5 4.2 6.5 24 1.1 8.8 15 15 0.1 0.5

2 19.7 2.9 8.6 22 1.2 24 46 14 0.4

3 32.3 4.6 3.1 37 1 12 33 18 0.1 0.5

4 16 6.6 13 40 1.6 12 13 9.1 0.2

5 2.9 0.6 0.1 4.8 0.6 0.4 0.9 10 0.2 0

6 32.3 1.7 3.3 8.4 0.2 2.8 3.5 0

7 19.3 2.2 0.6 16 1 5.2 9.2 22 0.2 0.3

8 13.9 17 8 110 5.7 42 25 4.4 0.6

9 20 4.2 2.2 89 0.7 8.7 13 23 0.2 0.3

10 20 19 19 110 9.7 84 55 2.2 0.3 5

11 4 3.3 0.5 29 0 0.5 0.3

12 39.6 5.7 8.5 39 1.2 20 17

15 35.3 8 7.8 63 0.3 19 22

16 14.3 0.7 1.5 3.7 0.1 2.9 3.7

17 31.4 2.8 4.5 18 1.0 15 19

18 27.3 19 5.6 100 0.7 10 17 1.1 0.1 0.8

19 148.3 21 32 140 1.1 36 31

20 6.2 23 30 41 7.9 5.5 4.3 13 1.6 4.8

25 0.9 0.2 0.1 0.4

26 0.2

30 0.03 3.2 4.5 14 0.8 1.5 1.3 0.7 0.1 0.2

Source: McArthur (1991).

Table A-4. Radionuclide areal soil deposition (Bq/m2).

Area Am-241 Pu-238 Pu-239,240 Co-60 Cs 137 Sr-90 Eu-152 Eu-154 Eu-155

1 2.26E+03 3.50E+03 1.29E+04 5.93E+02 4.74E+03 8.09E+03 8.09E+03 5.39E+01 2.70E+02

2 2.10E+03 6.24E+03 1.60E+04 8.70E+02 1.74E+04 3.34E+04 1.02E+04 0.00E+00 2.90E+02

3 2.03E+03 1.37E+03 1.64E+04 4.42E+02 5.31E+03 1.46E+04 7.96E+03 4.42E+01 2.21E+02

4 5.89E+03 1.16E+04 3.57E+04 1.43E+03 1.07E+04 1.16E+04 8.13E+03 0.00E+00 1.79E+02

5 2.96E+03 4.93E+02 2.36E+04 2.96E+03 1.97E+03 4.43E+03 4.93E+04 9.85E+02 0.00E+00

6 7.52E+02 1.46E+03 3.72E+03 8.85E+01 1.24E+03 1.55E+03 0.00E+00 0.00E+00 0.00E+00

7 1.63E+03 4.44E+02 1.18E+04 7.40E+02 3.85E+03 6.81E+03 1.63E+04 1.48E+02 2.22E+02

8 1.75E+04 8.22E+03 1.13E+05 5.86E+03 4.32E+04 2.57E+04 4.52E+03 0.00E+00 6.17E+02

9 3.00E+03 1.57E+03 6.36E+04 5.00E+02 6.21E+03 9.29E+03 1.64E+04 1.43E+02 2.14E+02

10 1.36E+04 1.36E+04 7.86E+04 6.93E+03 6.00E+04 3.93E+04 1.57E+03 2.14E+02 3.57E+03

11 1.18E+04 1.79E+03 1.04E+05 0.00E+00 1.79E+03 1.07E+03 0.00E+00 0.00E+00 0.00E+00

12 2.06E+03 3.07E+03 1.41E+04 4.33E+02 7.22E+03 6.13E+03 0.00E+00 0.00E+00 0.00E+00

15 3.24E+03 3.16E+03 2.55E+04 1.21E+02 7.69E+03 8.90E+03 0.00E+00 0.00E+00 0.00E+00

16 6.99E+02 1.50E+03 3.70E+03 9.99E+01 2.90E+03 3.70E+03 0.00E+00 0.00E+00 0.00E+00

17 1.27E+03 2.05E+03 8.19E+03 4.55E+02 6.82E+03 8.64E+03 0.00E+00 0.00E+00 0.00E+00

18 9.94E+03 2.93E+03 5.23E+04 3.66E+02 5.23E+03 8.90E+03 5.76E+02 5.23E+01 4.19E+02

19 2.02E+03 3.08E+03 1.35E+04 1.06E+02 3.47E+03 2.99E+03 0.00E+00 0.00E+00 0.00E+00

20 5.30E+04 6.91E+04 9.45E+04 1.82E+04 1.27E+04 9.91E+03 3.00E+04 3.69E+03 1.11E+04

25 0.00E+00 0.00E+00 0.00E+00 0.00E+00 3.17E+03 1.59E+03 6.35E+03 0.00E+00 0.00E+00

26 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00

30 1.52E+05 2.14E+05 6.67E+05 3.81E+04 7.14E+04 6.19E+04 3.33E+04 4.76E+03 9.52E+03

ATTACHMENT A

AMBIENT ENVIRONMENTAL INTAKES AT THE NEVADA TEST SITE

BASED ON AIR S AMPLING AND SOIL CONTAMINATION DATA 

Page 7 of 27
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Table A-5. Radionuclide areal soil deposition (Bq/m2) decay-corrected to 1963. Area Am-241 Pu-238 Pu-239,240 Co-60 Cs 137 Sr-90 Eu-152 Eu-154 Eu-155

1 2.37E+03 4.37E+03 1.29E+04 2.35E+04 9.06E+03 1.57E+04 3.47E+04 4.89E+02 1.35E+04

2 2.20E+03 7.78E+03 1.60E+04 3.45E+04 3.32E+04 6.49E+04 4.35E+04 0.00E+00 1.45E+04

3 2.13E+03 1.71E+03 1.76E+04 1.01E+04 2.84E+04 3.41E+04 4.01E+02 1.11E+04

4 6.16E+03 1.45E+04 3.57E+04 5.67E+04 2.05E+04 2.26E+04 3.48E+04 8.93E+03

5 3.09E+03 6.15E+02 2.37E+04 1.17E+05 3.76E+03 8.63E+03 2.11E+05 8.94E+03 0.00E+00

6 7.86E+02 1.82E+03 3.72E+03 3.51E+03 2.36E+03 3.01E+03

7 1.70E+03 5.54E+02 1.19E+04 2.94E+04 7.35E+03 1.33E+04 6.98E+04 1.34E+03 1.11E+04

8 1.83E+04 1.03E+04 1.13E+05 2.33E+05 8.24E+04 5.00E+04 1.94E+04 3.08E+04

9 3.14E+03 1.96E+03 6.36E+04 1.98E+04 1.19E+04 1.81E+04 7.04E+04 1.30E+03 1.07E+04

10 1.42E+04 1.69E+04 7.86E+04 2.75E+05 1.15E+05 7.65E+04 6.74E+03 1.94E+03 1.79E+05

11 1.23E+04 2.23E+03 1.04E+05 0.00E+00 3.41E+03 2.09E+03

12 2.15E+03 3.83E+03 1.41E+04 1.72E+04 1.38E+04 1.19E+04

15 3.39E+03 3.94E+03 2.55E+04 4.82E+03 1.47E+04 1.73E+04

16 7.31E+02 1.87E+03 3.70E+03 3.97E+03 5.53E+03 7.20E+03

17 1.33E+03 2.55E+03 8.20E+03 1.81E+04 1.30E+04 1.68E+04

18 1.04E+04 3.66E+03 5.24E+04 1.45E+04 9.99E+03 1.73E+04 2.47E+03 4.75E+02 2.09E+04

19 2.12E+03 3.85E+03 1.35E+04 4.21E+03 6.62E+03 5.81E+03

20 5.54E+04 8.62E+04 9.45E+04 7.23E+05 2.42E+04 1.93E+04 1.28E+05 3.34E+04 5.53E+05

25 6.06E+03 3.09E+03 2.72E+04

26

30 1.59E+05 2.67E+05 6.67E+05 1.51E+06 1.36E+05 1.21E+05 1.43E+05 4.32E+04 4.76E+05

A.4 Scaling Factors for Inhalation Intake Estimates

Because the air sampling program did not provide isotopic analyses for all of the identified 

radionuclides in NTS soils, scaling factors were developed to estimate potential intakes for these 

radionuclides based on the relative abundances of these radionuclides in comparison with 239Pu with 

soil contamination data (McArthur 1991) decay-corrected to 1963. These area-specific ratios are 

provided in Table A-6.

The scaling factors in Table A-6 were used in conjunction with the 239Pu intakes in Table A-2 to 

determine the potential environmental intakes of all the radionuclides important to dose (Table A-6) 

that have been identified as persistent in NTS soils. For purposes of dose reconstruction, maximum 

intakes of these radionuclides were calculated by selecting the maximum annual intake of plutonium 

(i.e., 0.4131 Bq/yr derived for Area 9 in 1972) and multiplying this value by the maximum scaling 

factor for each of the radionuclides in Table A-6. 

A.5 Correction for Resuspension for Early Times After Atmospheric Tests

Anspaugh (2002) stated that, based on empirical observations, concentration of resuspended 

radionuclides in air have been noted to display a strong time dependence at early times after 

deposition and that this pathway might be important for reoccupation of contaminated property. He 

goes on to state that there has not been universal agreement that resuspension is an important 

pathway but that it is now generally accepted that there are a few instances where the pathway could 

be the dominant one. Many observations have shown that the rate of resuspension decreases rapidly 

with time, and that for accident situations resuspension is only of importance (in comparison with the 

inhalation exposure from the initial cloud passage) over short periods. For this reason, Anspaugh 

(2002) states that the resuspension factor model has been widely used to predict the concentration of 

ATTACHMENT A

AMBIENT ENVIRONMENTAL INTAKES AT THE NEVADA TEST SITE

BASED ON AIR S AMPLING AND SOIL CONTAMINATION DATA 

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resuspended radionuclides at early times after initial deposition while the mass loading model (which 

uses measurements of dust loading in air and soil contamination data to predict air concentration of 

Table A-6. Abundance of radionuclides in NTS soils relative to 239Pu decay-corrected to 1963. 

Area Am-241 Pu-238 Pu-239, 240 Co-60 Cs 137 Sr-90 Eu-152 Eu-154 Eu-155

1 0.183 0.338 1.000 1.818 0.700 1.216 2.677 0.038 1.041

2 0.138 0.487 1.000 2.164 2.081 4.067 2.726 0.908

3 0.130 0.104 1.000 1.072 0.619 1.735 2.084 0.024 0.675

4 0.172 0.405 1.000 1.587 0.572 0.632 0.975 0.250

5 0.131 0.026 1.000 4.958 0.159 0.365 8.925 0.378

6 0.212 0.490 1.000 0.944 0.636 0.810

7 0.144 0.047 1.000 2.479 0.620 1.118 5.890 0.113 0.937

8 0.162 0.091 1.000 2.055 0.728 0.442 0.171 0.273

9 0.049 0.031 1.000 0.312 0.187 0.284 1.107 0.020 0.168

10 0.181 0.215 1.000 3.498 1.457 0.973 0.086 0.025 2.271

11 0.119 0.021 1.000 0.033 0.020

12 0.153 0.272 1.000 1.220 0.978 0.848

15 0.133 0.154 1.000 0.189 0.575 0.679

16 0.198 0.505 1.000 1.072 1.495 1.945

17 0.163 0.312 1.000 2.204 1.590 2.053

18 0.199 0.070 1.000 0.278 0.191 0.331 0.047 0.009 0.400

19 0.157 0.285 1.000 0.312 0.491 0.431

20 0.586 0.912 1.000 7.643 0.256 0.204 1.358 0.354 5.849

25

26

30 0.229 0.321 1.000 0.057 0.107 0.093 0.050 0.007 0.014

Maximum scaling factor 0.586 0.912 1.000 7.64 2.08 4.07 8.93 0.378 5.85

Scaled maximum intake, 

Bq/yr

0.223 0.347 0.381 2.91 0.792 1.55 3.40 0.144 2.23

radionuclides) has generally been preferred for times long after deposition. However, Anspaugh also 

states that it is always preferable to rely on actual measurements that are performed over long 

periods (such as those in Section A.2 of this report).

Anspaugh (2002) presents several resuspension models that have been proposed but concludes that 

they can be over- or under-predictive at various times after deposition in comparison with empirical 

observations. However, with expanded datasets from Hicks (1981e) and others in the 1980s, 

Anspaugh has proposed a resuspension model that more accurately describes the observed results 

over the entire time span of the expanded dataset for the NTS:

[ ] 07.05 003.09 9 1 10 106 1010 −− −− − ± = ×+ ×+ t t

f eS e (A-1) 

A graphical depiction of Equation A-1 is provided in Figure A-1 from time t equal zero to 1,000 d after 

detonation. As Figure A-1 shows, the resuspension factor Sf ranges from about 10-5 at times early

after deposition, falls rapidly during the first 100 or so days to a value of about 10-8, and then 

approaches a value of 10-9 after a few years. The factor of 10 at the end of the equation is a 

statement of uncertainty in the model. If it is true that the mass loading approach is more predictive at

times long after initial deposition and that the resuspension proposed by Anspaugh (2002) is 

predictive of the observed results over the expanded dataset (including those that were developed in 

the 1980s), then the factor of 10-9 could be taken to be the resuspension factor that would be 

predictive of the mass loading process that is thought to be more important during the times when the 

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AMBIENT ENVIRONMENTAL INTAKES AT THE NEVADA TEST SITE

BASED ON AIR S AMPLING AND SOIL CONTAMINATION DATA 

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air monitoring data are available (i.e., 1971 through 2001, see Section A.2). When Anspaugh’s 

proposed resuspension model (i.e., Equation A-1) is integrated from 180 to 545 days and compared 

to the result of the integral of the constant 10-9 over the same period, a factor is developed that can be

used to correct the intakes that are derived from air sampling data (i.e., times long after initial 

deposition) for the early resuspension phenomenon that has been observed at NTS. The period of 

integration was selected to begin at day 180 because the last atmospheric tests at NTS were in July 

of 1962 and therefore about 180 days had passed before the beginning of 1963 (the starting date for 

reconstruction of internal dose at NTS in the absence of bioassay results for the employee). For the 

Anspaugh model, the early resuspension correction factor has been determined to be 3.12. In a 

similar manner, correction factors for 1964 and 1965 were determined to be 1.72 and 1.24, 

respectively. Therefore, for dose reconstruction purposes, the scaled maximum inhalation intakes in 

Table A-6 should be multiplied by these factors to account for early resuspension from 1963 through 

1965. These resultant intakes are shown in Table A-7. If necessary, these intakes can be prorated

for time less than a year for a best estimate if the worker was on the site for only a fraction of the year.

Resuspension Factor

1.3 E-09 m-1

1.E-09

1.E-08

1.E-07

1.E-06

1.E-05

0 100 200 300 400 500 600 700 800 900 1000

Days

meter-1

Figure A-1. Resuspension factor as a function of time after initial deposition. 

Table A-7. Scaled inhalation intakes corrected for early resuspension in 1963 (Bq/yr). 

Year of intake Am-241 Pu-238 Pu-239,240 Co-60 Cs 137 Sr-90 Eu-152 Eu-154 Eu-155

1963 0.70 1.08 1.19 9.08 2.47 4.84 10.61 0.45 6.96

1964 0.38 0.59 0.65 4.98 1.35 2.65 5.81 0.25 3.81

1965 0.28 0.43 0.47 3.61 0.98 1.92 4.22 0.18 2.77

All subsequent years 0.223 0.347 0.381 2.91 0.792 1.55 3.40 0.144 2.23

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A.6 Corrections for Inhalation Dose From Short-lived Fission and Activation Products

During the early and mid-1960s, workers could have been exposed to fallout from atmospheric testing 

that contained short-lived fission and activation products (e.g., 144Ce, 106Ru, etc.) that have not 

persisted in NTS soils in measurable quantities. For purposes of dose reconstruction, two methods 

are described to account for possible exposure to these radionuclides. 

Potential inhalation dose from short-lived fission and activation products can be estimated using data 

that were developed by Hicks (1981e). Hicks (1981e) provides estimates of ground deposition of 

152 fission products and 25 neutron-induced nuclides from nuclear weapons tests. The data for these 

calculations were for the last atmospheric tests NTS conducted in 1962 as a function of time after 

detonation. Spreadsheets were developed to multiply the relative abundance of each of these 

radionuclides by its associated International Commission on Radiological Protection (ICRP) 

Publication 68 (ICRP 1995) inhalation organ dose conversion factor (Bunker 1999) to determine the 

relative importance of each of the 177 radionuclides to total organ dose. Except for 90Sr, the 

inhalation dose conversion factors were those for the absorption type (i.e., S, M, or F) that produced 

the largest dose to the organ of interest. For 90Sr, only absorption Type F was used. In addition, 

because the Publication 68 organ dose factors represent the 50-yr committed dose, the IMBA 

computer program was used to develop the annual organ-specific dose conversion factor for 241Am,

which was used in the calculations to better represent the relative importance 241Am, which delivers 

dose to the affected organ over long periods of time. Similarly, annual organ-specific dose conversion 

factors were developed for 90Sr to provide a more accurate time-dependant correction factor.

The development of the short-lived fission and activation product correction factors based on the 

Hicks data must be adjusted for fractionation. Fractionation is a phenomenon due to both chemical 

and physical separation of the radionuclides in the fireball in first few minutes after detonation. Within 

the first minute after detonation, the vaporized soil components condense with other refractory 

elements and begin to fall to the surface. The volatile elements (except for krypton and xenon) and 

their progeny condense in 6 to 8 minutes and begin falling to the surface (Hicks 1981e). Because the 

Hicks data was developed to estimate offsite levels of fallout and resultant dose, fractionation effects 

were simulated in the Hicks data by the removal of a fraction of the refractory nuclides from the 

calculated abundances. In general, air drops were assumed to be unfractionated and offsite fallout 

from surface and cratering tests was assumed to have 0.4 of the refractory elements. For all other 

types of tests, offsite fallout was assumed to have 0.5 of the refractory elements present. Therefore, 

the refractory elements (e.g., beryllium, sodium, manganese, iron, cobalt, copper, yttrium, zirconium, 

niobium, barium, rare earths, thorium, uranium, neptunium, plutonium, americium, and curium) in the 

Hicks data must be adjusted to produce the best estimate of their enriched abundances in the onsite 

environment to which the workers could have been exposed. Adjustment factors for each 

radionuclide were determined from data in Hicks (1984). This report provided relative abundances of 

radionuclides assuming no fraction, 50% fraction, and 90% fraction of refractory elements. From 

these data, ratios were developed for the 50%-fractionation case (Table A-8). These ratios were used 

to deplete the refractory elements in the far-field (i.e., offsite) environment to estimate doses to offsite 

individuals. Therefore, to enrich the near-field (i.e., onsite) environment, the inverse of these ratios 

was applied to the Hicks Small Boy data (see below). These inverse ratios were applied twice 

because the Hicks Small Boy data was provided to estimate fallout in the offsite environment. The 

first application results in the data that represent no fractionation while the second application results 

in data that is enriched with refractory elements. 

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In Table A-8, the radionuclides with ratios less than 1.0 represent the radionuclides that are 

considered to condense as refractory elements while the radionuclides with ratios greater than 1.0 are 

assumed to behave as volatile elements. It should be noted that although strontium is considered to 

be a refractory element, for purposes of these calculation, the assumption is made that it behaves like 

a volatile element because its fission precursors are noble gases (i.e., volatile elements). 

Table A-8. Far-field refractory enrichment ratios. Radionuclide Ratio

Radionuclide RatioRadionuclide RatioRadionuclide RatioRadionuclide RatioRadionuclide Ratio

Be-7 0.68 Br-82 1.36 Tc-101 1.36 Cd-118 1.36 I-132 1.36 Pr-145 0.68

Na-24 0.68 Se-83 1.36 Mo-102 1.36 In-118 1.36 Te-133m 1.36 Ce-146 0.68

Mn-54 0.68 Br-83 1.36 Tc-102m 1.36 Cd-119 1.36 Te-133 1.36 Pr-146 0.68

Fe-55 0.68 Kr-83m 1.37 Tc-102 1.36 In-119m 1.36 I-133 1.36 Pr-147 0.68

Fe-59 0.68 Br-84 1.36 Ru-103 1.36 In-119 1.36 Xe-133m 1.36 Nd-147 0.68

Cu-64 0.68 Kr85m 1.36 Rh-103m 1.36 Sn-121 1.36 Xe-133 1.36 Nd-149 0.68

Cu-67 0.68 Kr-87 1.36 Tc-104 1.36 Sn-123m 1.37 Te-134 1.36 Pm-149 0.68

W-181 1.36 Kr-88 1.36 Ru-105 1.36 Sn-123 1.36 I-134 1.36 Pm-150 0.68

W-185 1.35 Rb-88 1.36 Rh-105m 1.36 Sn-125 1.36 I-135 1.36 Nd-151 0.68

W-187 1.36 Rb-89 1.36 Rh-105 1.36 Sb-125 1.36 Xe-135m 1.36 Pm-151 0.68

W-188 1.36 Sr-89 1.36 Ru-106 1.36 Sb-126 1.36 Xe-135 1.36 Pm-152 0.68

U-237 0.68 Sr-90 1.36 Rh-106 1.36 Sn-127 1.36 Cs-136 1.36 Sm-153 0.68

U-239 0.68 Sr-91 1.19 Rh-107 1.36 Sb-127 1.36 Cs-137 1.36 Sm-155 0.68

U-240 0.68 Y-91m 1.19 Pd-107m 1.36 Te-127 1.36 Ba-137m 1.36 Eu-155 0.68

Np-239 0.68 Y-91 1.19 Pd-109 1.36 Sn-128 1.36 Xe-138 1.36 Sm-156 0.68

Np240m 0.69 Sr-92 0.68 Ag-109m 1.36 Sb-128m 1.37 Cs-138 1.36 Eu-156 0.68

Np-240 0.68 Y-92 0.68 Pd-111m 1.37 Sb-128 1.37 Cs-139 1.36 Eu-157 0.68

Am-241 0.68 Sr-93 0.68 Pd-111 1.36 Sn-129m 1.36 Ba-139 1.36 Eu-158 0.68

Cm-242 0.68 Y-93 0.68 Ag-111m 1.36 Sn-129 1.36 Ba-140 1.16 Eu-159 0.68

Ge-75 1.36 Y-94 0.68 Ag-111 1.36 Sb-129 1.36 La-140 1.16 Gd-159 0.68

Ge-77 1.36 Y-95 0.68 Pd-112 1.36 Te-129m 1.36 Ba-141 0.90 Tb-161 0.68

As-77 1.36 Zr-95 0.68 Ag-112 1.36 Te-129 1.36 La-141 0.90

Se-77m 1.36 Nb-95 0.68 Ag-113 1.36 Sb-130m 1.36 Ce-141 0.90

Ge-78 1.36 Zr-97 0.68 Ag-115 1.36 Sb-130 1.36 Ba-142 0.68

As-78 1.36 Nb-97m 0.68 Cd-115m 1.36 I-130 1.36 La-142 0.68

As-79 1.37 Nb-97 0.68 Cd-115 1.36 Sb-131 1.36 La-143 0.68

Se-79m 1.36 Nb-98 0.68 In-115m 1.37 Te-131m 1.36 Ce-143 0.68

Br-80 1.36 Mo-99 0.68 Cd-117 1.36 Te-131 1.36 Pr-143 0.68

Se-81m 1.36 Tc-99m 0.68 In-117m 1.36 I-131 1.36 Ce-144 0.68

Se-81 1.36 Mo-101 1.36 In-117 1.36 Te-132 1.36 Pr-144 0.68

The fission and activation product correction factor was developed based on the relative contribution 

of 90Sr to total organ dose. Strontium-90 was chosen because Hicks provided time-dependant 

abundances for this radionuclide and 90Sr continues to persist in the NTS environment. Therefore, by 

determining the relative importance of organ dose from 90Sr to the total dose from all 177 

radionuclides, a multiplication factor was developed, by which the various organ doses from the 90Sr 

intakes (Table A-7) can be multiplied, to account for short-lived fission and activation products. For 

example, using the Hicks data for Storax Small Boy, which was the next-to-last atmospheric test at 

NTS on July 14, 1962 (Storax Little Feller I was the last atmospheric test on July 17, 1962), it was 

determined that the relative importance of 90Sr dose to the lung to the total dose from all 177 

radionuclides varied, in a mostly linear fashion, from 0.00284 to 0.0589 from 1 to 365 days after 

detonation, respectively (see Figure A-2). Using the trend line function, an expression for the relative 

importance of 90Sr dose y to total dose as a function of the number of days after detonation x was 

developed:

y = 0.0001x + 0.0074 (A-2) 

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y = 0.0001x + 0.0037

R2 = 0.9824

0.00

0.05

0.10

0.15

0.20

0.25

0.30

0.35

0.40

0.45

0 500 1000 1500 2000 2500 3000 3500 4000

Days After Detonation

Fraction of Total Dose

Figure A-2. Small Boy – 90Sr fraction of total dose lungs. 

To ensure favorability to claimants, the relative importance of 90Sr lung dose to the total lung dose 

from all fission and activation products was examined for several tests including Teapot Turk in 1955 

and Little Feller I in 1962 (Hicks 1981a,b,c,d). The slope of the trend lines that predict the relative 

importance of 90Sr dose was determined to be 0.0001x for Storax Small Boy, 0.0002x for Storax Little 

Feller I (Figure A-3) Teapot Turk (see Figure A-4). Because the slope of the trend line is directly 

proportional to the relative importance of the 90Sr dose to total dose (i.e., the larger the slope, the 

larger the relative importance of 90Sr dose), the tests with the smallest slopes result in the highest 

multiplicative correction factors for fission and activation products. Therefore, to ensure the organ 

dose from short-lived fission and activation products is not underestimated, the Hicks data for the 

Storax Small Boy test was selected to be used to determine the fission and activation product dose 

correction factor. This assumption is also justified by the fact that the test was very near the last 

atmospheric test (i.e., Storax Little Feller I) and would therefore have been the test most likely to 

produce the short-lived fission and activation product intakes for workers at NTS after 1962 (the 

period of time for which organ dose from environmental intakes is calculated).

Integrating Equation A-2 for Small Boy from 0 to 365 d and dividing the result by 365 (the value that 

represents the integrated total dose for 1 yr), it was determined that for the first year after detonation 

the lung dose from 90Sr represented 0.0000738 or about 0.00738% of the dose from all 177 

radionuclides. Therefore, the inverse of the value would produce a factor of 13,600 that the lung dose 

from the 90Sr intake (Table A-7) could be multiplied by to account for inhalation dose from short-lived 

fission and activation products. Similar integrations were performed for subsequent years through 

1972. 

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Figure A-3. Little Feller I – 90Sr fraction of total dose to lungs. 

y = 0.0002x - 0.0124

R2

 = 0.9926

0

0.1

0.2

0.3

0.4

0.5

0.6

0.7

0.8

0.9

0 500 1000 1500 2000 2500 3000 3500 4000

Days After Detonation

Fraction of Totla dose

Figure A-4. Turk – 90Sr fraction of total dose to lungs. 

Correction factors to account for inhalation intakes of short-lived fission and activation products have 

been developed for all organs using the Hicks data for Storax Small Boy (Table A-9). These 

correction factors were based on the relationships that Figures A-5 through A-11 show. 

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Table A-9 Organ-specific inhalation dose fission and activation product correction factors. Organ Fission and activation product correction factor

Year 1963 1964 1965 1966 1967 1968 1969 1970 1971 1972

Skin Adrenals

Thymus SI 

Spleen Skin 

Muscle Uterus

Pancreas Kidneys

Breast Testes

Esophagus Ovaries

Brain Stomach

Thyroid Gall Bladder

730 364 242 182 145 121 104 90.7 80.6 72.5

ULI 458 179 99.2 64.0 45.0 33.4 25.9 20.6 16.8 14.0

Urinary Bladder 335 149 91.3 63.6 47.5 37.2 30.1 24.9 21.0 18.0

Lungs 34,900 14,200 7,960 5,150 3,630 2,700 2,100 1,660 1,360 1,130

ET ET1 ET2 LN(TH)

LN(ET)

1,570 827 598 492 438 412 412 412 412 412

LLI 420 142 70.8 42.4 28.2 20.1 15.1 11.7 9.4 7.6

Colon 390 148 79.4 50.0 34.5 25.2 19.3 15.2 12.3 10.2

Liver 9,260 4,620 1,540 1,190 988 858 769 706 661 629

Red Bone Marrow 37.9 18.2 12.8 10.4 10.4 10.4 10.4 10.4 10.4 10.4

Bone Surfaces 78.5 40.1 28.1 22.4 22.4 22.4 22.4 22.4 22.4 22.4

 

y = 4E-06x - 4E-05

R2

 = 0.9936

0.00000

0.00500

0.01000

0.01500

0.02000

0.02500

0.03000

0.03500

0 500 1000 1500 2000 2500 3000 3500 4000

Days After Detonation

Fraction of Total Dose

Adrenals Brain Breast Esophagus

Stomach Small Intestine Kidneys Muscle

Ovaries Pancreas Skin Spleen

Testes Thymus Thyroid Uterus

Linear (Kidneys)

Figure A-5. Strontium-90 fraction of total dose. 

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y = 4E-09x2

 + 5E-06x - 0.0002

R2

 = 0.9999

y = 2E-09x2

 + 8E-06x - 0.0002

R2

 = 0.9997

-0.01000

0.00000

0.01000

0.02000

0.03000

0.04000

0.05000

0.06000

0.07000

0.08000

0 500 1000 1500 2000 2500 3000 3500 4000

Days After Detonation

Fraction of Total Dose

Urinary Bladder Upper Large Intestine Poly. (Upper Large Intestine) Poly. (Urinary Bladder)

Figure A-6. Strontium-90 fraction of total dose. 

 

y = -2E-11x2

 + 2E-07x - 9E-06

R2 = 0.995

y = -4E-10x2

 + 2E-06x - 3E-05

R2

 = 0.9964

0.00000

0.00050

0.00100

0.00150

0.00200

0.00250

0.00300

0.00350

0 500 1000 1500 2000 2500 3000 3500 4000

Days After Detonation

Fraction of Total Dose

Extrathoracic Airways Lungs Poly. (Lungs) Poly. (Extrathoracic Airways)

Figure A-7. Strontium-90 fraction of total dose. 

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y = 9E-09x2

 + 3E-06x + 1E-05

R2

 = 1

y = 6E-09x2

 + 5E-06x - 0.0001

R2

 = 1

0.00000

0.02000

0.04000

0.06000

0.08000

0.10000

0.12000

0.14000

0 500 1000 1500 2000 2500 3000 3500 4000

Days After Detonation

Fraction of Total Dose

Lower Large Intestine Colon Poly. (Lower Large Intestine) Poly. (Colon)

Figure A-8. Strontium-90 fraction of total dose. 

y = -7E-11x2 + 7E-07x - 3E-05

R2

 = 0.9942

0.000

0.000

0.000

0.001

0.001

0.001

0.001

0.001

0.002

0 500 1000 1500 2000 2500 3000 3500 4000

Days After Detonation

Fraction of Totoal Dose

Figure A-9. Strontium-90 fraction of total dose to the liver. 

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y = -2E-08x2

 + 0.0001x - 0.007

R2

 = 0.9958

0.00

0.05

0.10

0.15

0.20

0.25

0.30

0.35

0 500 1000 1500 2000 2500 3000 3500 4000

Days After Detonation

Fraction of Total Dose

Figure A-10. Strontium-90 fraction of total dose red marrow. 

 

y = -6E-09x2 + 4E-05x - 0.0009

R2

 = 0.9983

-0.01

0.00

0.01

0.02

0.03

0.04

0.05

0.06

0.07

0 500 1000 1500 2000 2500 3000 3500 4000

Days AfterDetonation

Fraction of Total Dose

Figure A-11. Strontium-90 fraction of total dose to bone surface. 

The IMBA computer program was used to determine the organ doses from the scaled intakes 

provided in Table A-7 for a period of 10 yr. These doses were then multiplied by the correction factors 

in Table A-9 to determine the additional dose that should be added to account for potential dose from 

inhalation of short-lived fission and activation products. The organ-specific fission and activation 

doses are provided in Table A-10. 

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Table A-10. Inhalation dose from short-lived fission and activation products (rem). Year 1963 1964 1965 1966 1967 1968 1969 1970 1971 1972

Adrenals 1.52E-04 6.76E-05 3.92E-05 2.89E-05 2.47E-05 2.17E-05 1.94E-05 1.74E-05 1.57E-05 1.42E-05

Bladder 1.86E-04 6.99E-05 3.45E-05 2.26E-05 1.81E-05 1.51E-05 1.29E-05 1.12E-05 9.73E-06 8.53E-06

Brain 1.52E-04 6.76E-05 3.92E-05 2.89E-05 2.47E-05 2.17E-05 1.94E-05 1.74E-05 1.57E-05 1.42E-05

Breast 2.08E-07 3.57E-08 1.65E-08 1.37E-08 1.18E-08 1.01E-08 8.70E-09 7.50E-09 6.49E-09 5.58E-09

Gall bladder 2.08E-07 3.57E-08 1.65E-08 1.37E-08 1.18E-08 1.01E-08 8.70E-09 7.50E-09 6.49E-09 5.58E-09

Heart wall 1.52E-04 6.76E-05 3.92E-05 2.89E-05 2.47E-05 2.17E-05 1.94E-05 1.74E-05 1.57E-05 1.42E-05

Kidney 1.52E-04 6.76E-05 3.92E-05 2.89E-05 2.47E-05 2.17E-05 1.94E-05 1.74E-05 1.57E-05 1.42E-05

Liver 1.93E-03 8.57E-04 3.71E-04 2.70E-04 2.33E-04 2.07E-04 1.85E-04 1.67E-04 1.52E-04 1.39E-04

Muscle 1.52E-04 6.76E-05 3.92E-05 2.89E-05 2.47E-05 2.17E-05 1.94E-05 1.74E-05 1.57E-05 1.42E-05

Ovaries 1.52E-04 6.76E-05 3.92E-05 2.89E-05 2.47E-05 2.17E-05 1.94E-05 1.74E-05 1.57E-05 1.42E-05

Pancreas 1.52E-04 6.76E-05 3.92E-05 2.89E-05 2.47E-05 2.17E-05 1.94E-05 1.74E-05 1.57E-05 1.42E-05

Testes 1.52E-04 6.76E-05 3.92E-05 2.89E-05 2.47E-05 2.17E-05 1.94E-05 1.74E-05 1.57E-05 1.42E-05

Thyroid 1.52E-04 6.76E-05 3.92E-05 2.89E-05 2.47E-05 2.17E-05 1.94E-05 1.74E-05 1.57E-05 1.42E-05

RBM 4.08E-04 7.00E-04 6.65E-04 6.09E-04 5.59E-04 5.18E-04 4.84E-04 4.57E-04 4.35E-04 4.16E-04

Bone surface 1.23E-03 2.03E-03 2.00E-03 1.91E-03 1.83E-03 1.77E-03 1.72E-03 1.69E-03 1.66E-03 1.64E-03

Stomach 1.69E-04 7.22E-05 4.14E-05 3.02E-05 2.57E-05 2.26E-05 2.01E-05 1.81E-05 1.63E-05 1.47E-05

SI 1.85E-04 7.67E-05 4.36E-05 3.15E-05 2.68E-05 2.35E-05 2.09E-05 1.87E-05 1.69E-05 1.53E-05

ULI 4.70E-04 1.36E-04 5.98E-05 3.59E-05 2.73E-05 2.19E-05 1.81E-05 1.52E-05 1.29E-05 1.11E-05

LLI 1.38E-03 3.26E-04 1.24E-04 6.66E-05 4.74E-05 3.63E-05 2.90E-05 2.37E-05 1.98E-05 1.67E-05

Skin 1.52E-04 6.76E-05 3.92E-05 2.89E-05 2.47E-05 2.17E-05 1.94E-05 1.74E-05 1.57E-05 1.42E-05

Spleen 1.52E-04 6.76E-05 3.92E-05 2.89E-05 2.47E-05 2.17E-05 1.94E-05 1.74E-05 1.57E-05 1.42E-05

Thymus 1.52E-04 6.76E-05 3.92E-05 2.89E-05 2.47E-05 2.17E-05 1.94E-05 1.74E-05 1.57E-05 1.42E-05

Uterus 1.52E-04 6.76E-05 3.92E-05 2.89E-05 2.47E-05 2.17E-05 1.94E-05 1.74E-05 1.57E-05 1.42E-05

ET 2.22E-03 6.97E-04 3.78E-04 2.60E-04 2.32E-04 2.19E-04 2.17E-04 2.15E-04 2.14E-04 2.13E-04

Lung 7.67E-03 2.96E-03 1.55E-03 1.08E-03 8.84E-04 7.46E-04 6.40E-04 5.52E-04 4.80E-04 4.18E-04

Colon 7.82E-04 2.08E-04 8.57E-05 4.85E-05 3.57E-05 2.78E-05 2.26E-05 1.86E-05 1.57E-05 1.34E-05

ET1 1.63E+00 4.74E-01 2.48E-01 1.64E-01 1.46E-01 1.38E-01 1.38E-01 1.38E-01 1.38E-01 1.38E-01

ET2 5.83E-04 2.24E-04 1.30E-04 9.60E-05 8.63E-05 8.10E-05 7.92E-05 7.78E-05 7.66E-05 7.55E-05

LN(ET) 3.27E-04 1.50E-04 9.15E-05 7.03E-05 6.34E-05 5.95E-05 5.77E-05 5.63E-05 5.51E-05 5.40E-05

LN(TH) 3.27E-04 1.50E-04 9.15E-05 7.03E-05 6.34E-05 5.95E-05 5.77E-05 5.63E-05 5.51E-05 5.40E-05

Esophagus 1.52E-04 6.76E-05 3.92E-05 2.89E-05 2.47E-05 2.17E-05 1.94E-05 1.74E-05 1.57E-05 1.42E-05

The application of the early fission and activation product correction factor through 1972 intakes is 

favorable to claimants because the Hicks data verifies that after 10 years, over 90% of the total organ 

dose for all organs is delivered by 241Am, 137Cs, and/or 90Sr. Further, as times after detonation 

become greater, the relative importance of 241Am, 137Cs, and 90Sr become greater and, because the 

organ doses from these radionuclides are already accounted for by the scaled intakes (Table A-7), the 

dose from these radionuclides is doubled. 

A.7 Ingestion Pathway

To account for potential intakes from inadvertent ingestion of contaminated soil, the area-specific 

radionuclide soil deposition data in Table A-5 were converted to volumetric data (i.e., Bq/mg) by 

assuming a radionuclide relaxation depth of 2.3 cm (DOE 2003) and a soil density of 1.5 g/cm3

 (DOE 

2003). The area-specific radionuclide soil concentrations are presented in Table A-11.

If the assumption is made that the workers ingested 100 mg of soil each day, which is favorable to 

claimants (EPA [1989] recommends a value of 50 mg/d), and the assumption is made that full-time 

employment was 250 d/yr, annual ingestion can be calculated. The area-specific annual ingestion 

rates are shown in Table A-12. 

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Table A-11. Radionuclide soil concentration by area (Bq/g). 

Area Am-241 Pu-238 Pu-239,240 Co-60 Cs 137 Sr-90 Eu-152 Eu-154 Eu-155

1 0.069 0.127 0.375 0.682 0.263 0.456 1.005 0.014 0.391

2 0.064 0.226 0.463 1.001 0.963 1.882 1.262 0.420

3 0.062 0.050 0.475 0.509 0.294 0.824 0.989 0.012 0.321

4 0.179 0.420 1.036 1.644 0.593 0.655 1.010 0.259

5 0.090 0.018 0.686 3.401 0.109 0.250 6.122 0.259

6 0.023 0.053 0.108 0.102 0.069 0.087 0.000 0.00

7 0.049 0.016 0.344 0.852 0.213 0.384 2.024 0.039 0.322

8 0.530 0.297 3.280 6.741 2.389 1.450 0.562 0.894

9 0.091 0.057 1.844 0.575 0.344 0.524 2.042 0.038 0.311

10 0.411 0.491 2.279 7.973 3.321 2.217 0.195 0.056 5.177

11 0.357 0.065 3.005 0.099 0.060

12 0.062 0.111 0.408 0.498 0.399 0.346

15 0.098 0.114 0.740 0.140 0.426 0.502

16 0.021 0.054 0.107 0.115 0.160 0.209

17 0.039 0.074 0.238 0.524 0.378 0.488

18 0.301 0.106 1.518 0.421 0.290 0.502 0.072 0.014 0.607

19 0.061 0.111 0.391 0.122 0.192 0.169

20 1.607 2.500 2.741 20.946 0.701 0.559 3.723 0.969 16.031

25 0.176 0.090 0.789

26

30 4.620 7.749 19.340 43.835 3.953 3.493 4.143 1.252 13.804

Table A-12. Area-specific and maximum annual ingestion rates (Bq/yr). 

Area Am-241 Pu-238 Pu-239,240 Co-60 Cs 137 Sr-90 Eu-152 Eu-154 Eu-155

1 1.72 3.17 9.38 17.06 6.56 11.41 25.12 0.35 9.77

2 1.59 5.64 11.57 25.03 24.08 47.06 31.54 10.51

3 1.54 1.24 11.87 12.72 7.34 20.59 24.73 0.29 8.01

4 4.47 10.49 25.90 41.10 14.83 16.37 25.24 6.47

5 2.24 0.45 17.15 85.03 2.73 6.25 153.05 6.48 0.00

6 0.57 1.32 2.69 2.54 1.71 2.18

7 1.23 0.40 8.59 21.29 5.33 9.61 50.59 0.97 8.05

8 13.24 7.43 81.99 168.52 59.73 36.25 14.05 0.00 22.34

9 2.27 1.42 46.11 14.38 8.60 13.10 51.04 0.94 7.76

10 10.29 12.27 56.98 199.31 83.02 55.42 4.88 1.41 129.41

11 8.93 1.61 75.12 0.00 2.47 1.51

12 1.56 2.77 10.20 12.45 9.98 8.65

15 2.45 2.85 18.49 3.49 10.64 12.56

16 0.53 1.35 2.68 2.87 4.01 5.21

17 0.97 1.85 5.94 13.09 9.44 12.19

18 7.54 2.65 37.95 10.54 7.24 12.55 1.79 0.34 15.17

19 1.53 2.79 9.78 3.05 4.80 4.21

20 40.17 62.49 68.51 523.64 17.54 13.98 93.06 24.23 400.77

25 4.39 2.24 19.73

26

30 115.50 193.72 483.50 1095.89 98.83 87.33 103.56 31.30 345.10

Maxa 40.17 62.49 81.99 523.64 83.02 55.42 153.05 24.23 400.77

a. Maximum value with Area 30 excluded.

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It should be noted that for most radionuclides, Area 30 provided the highest areal deposition and 

resultant intakes. Area 30 is relatively small (150 km2

), inaccessible, and is located on the Western 

edge of NTS. It has rugged terrain and includes the northern reaches of Fortymile Canyon. In 1968, 

it was the site of Project BUGGY, the first nuclear row-charge experiment in the PLOWSHARE 

Program. As a result of the test, a trench 255 m long, 77 m wide, and 206 m deep was created. The 

test resulted in large quantities of vitrified glass. Because of the bias that is introduced when Area 30 

is included, the maximum annual intakes in Table A-12 have been provided with Area 30 areal

concentrations excluded. The Area 30 intakes should only be used if it can be determined that the 

worker was assigned to and primarily worked in Area 30.

To determine the relative importance of the ingestion pathway, a hypothetical ingestion scenario was

used that assumed 30 years of the Table A-12 maximum intakes and used the IMBA computer 

program to determine organ doses from these intakes. The annual organ doses from ingestion of 

238,239

Pu and 241Am greater than 0.001 rem are presented in Table A-13. With the exception of the red 

bone marrow and bone surfaces, the annual doses from the Table A-12 maximum ingestion rates of 

60Co, 137Cs, and 152,154,155Eu were all less than 0.001 rem. For the red bone marrow and bone 

surfaces, the annual doses were all less than 0.002 rem.

Similar to what was done for inhalation intakes discussed is Section A.5 of this report, ingestion doses 

need to be adjusted for potential dose from short-lived fission and activation products that are no 

longer persistent in NTS soils in measurable amounts. The organ-specific fission and activation 

product correction factors were again developed (for the reasons that are provided in Section A.5) 

based on the relative contribution of 90Sr to the total ingestion dose using the Hicks [1981c] data from 

the Storax Small Boy test in July of 1962. As was done in Section A.5 for inhalation dose, organspecific relationships were developed for the 90Sr fraction of total dose. These relationships are 

provided in Figures A-12 through A-17. With the exception of fractional dose to bone surfaces and 

red bone marrow, these relationships were valid out to 10 years after detonation. For bone surfaces 

and red bone marrow, the time dependant relationships were determined for the first year and for the 

last 9 years. 

To account for fractionation, the refractory elements included in the Storax Small Boy Hicks data were 

multiplied by a factor of 2. Increasing the abundance of the refractory elements provides a more 

reasonable estimate of their relative contribution to total dose. As was done for the inhalation 

correction factors, these radionuclide- and time-dependant relative abundances are then multiplied by 

their organ-specific ingestion dose conversion factors (Bunker 1999) to determine the radionuclidespecific, relative importance to total dose as a function of time after detonation. It should be noted 

that because of the large difference between the 50-year committed dose conversion factor and the 

annual dose conversion factor for 241Am, the IMBA computer program was used to calculate the 

annual ingestion dose conversion factors which were then used instead of the 50-yr committed dose 

conversion factors. The same method was also used for 90Sr annual dose conversion factors. To 

ensure that organ doses were not underestimated, the f1 factor for each radionuclide was chosen to 

provide the largest dose to the specific organ of interest.

It should be noted that the ingestion correction factors were evaluated over a 10-yr period. This is 

due to the fact that 144Ce and 106Ru continue to provide relatively large ingestion doses through 5 

years after detonation. Therefore, to capture their contributions to total dose, the correction factor 

integration times were extended to 10 years where the relative contribution of 144Ce and 106Ru to total 

dose was less than three percent for all organs. Also, at the end of 10 years, over 95% of the total 

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Table A-13. Annual organ doses from ingestion of 238,239Pu and 241Am

(rem). 

Year Liver Ovaries Testes RBM

Bone 

surface LLI

1963 0.002 0.001

1964 0.001 0.007 0.001

1965 0.002 0.001 0.011 0.001

1966 0.003 0.001 0.016 0.001

1967 0.004 0.002 0.020 0.001

1968 0.005 0.002 0.024 0.001

1969 0.005 0.002 0.028 0.001

1970 0.006 0.002 0.032 0.001

1971 0.007 0.003 0.035 0.001

1972 0.008 0.003 0.039 0.001

1973 0.008 0.003 0.043 0.001

1974 0.009 0.003 0.046 0.001

1975 0.010 0.003 0.050 0.001

1976 0.010 0.004 0.053 0.001

1977 0.011 0.004 0.057 0.001

1978 0.011 0.004 0.060 0.001

1979 0.012 0.004 0.063 0.001

1980 0.013 0.001 0.001 0.004 0.066 0.001

1981 0.013 0.001 0.001 0.004 0.070 0.001

1982 0.014 0.001 0.001 0.004 0.073 0.001

1983 0.014 0.001 0.001 0.004 0.076 0.001

1984 0.015 0.001 0.001 0.005 0.079 0.001

1985 0.015 0.001 0.001 0.005 0.082 0.001

1986 0.016 0.001 0.001 0.005 0.085 0.001

1987 0.016 0.001 0.001 0.005 0.088 0.001

1988 0.017 0.001 0.001 0.005 0.091 0.001

1989 0.017 0.001 0.001 0.005 0.094 0.001

1990 0.018 0.001 0.001 0.005 0.096 0.001

1991 0.018 0.002 0.002 0.005 0.099 0.001

1992 0.019 0.002 0.002 0.005 0.102 0.001

1993 0.019 0.002 0.002 0.005 0.105 0.001

1994 0.019 0.002 0.002 0.005 0.105

1995 0.019 0.002 0.002 0.005 0.103

1996 0.018 0.002 0.002 0.005 0.101

1997 0.018 0.002 0.002 0.004 0.099

1998 0.017 0.002 0.002 0.004 0.098

1999 0.017 0.001 0.001 0.004 0.096

2000 0.017 0.001 0.001 0.004 0.095

2001 0.016 0.001 0.001 0.004 0.093

2002 0.016 0.001 0.001 0.003 0.092

2003 0.016 0.001 0.001 0.003 0.090

2004 0.015 0.001 0.001 0.003 0.089

2005 0.015 0.001 0.001 0.003 0.088

2006 0.015 0.001 0.001 0.003 0.086

2007 0.014 0.001 0.001 0.003 0.085

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y = 2E-05x - 0.001

R2

 = 0.9845

y = 0.0684Ln(x) - 0.1775

R2

 = 0.88

0.00000

0.10000

0.20000

0.30000

0.40000

0.50000

0.60000

0 500 1000 1500 2000 2500 3000 3500 4000

Days After Detonation

Fraction of Total Dose

Bone Surface Upper Large Intestine Lower Large Intestine

Colon Red Marrow Linear (Upper Large Intestine)

Log. (Bone Surface)

Figure A-12. Strontium-90 fraction of total ingestion dose. 

 

y = 0.0008x - 0.0107

R2

 = 0.9879

-0.05000

0.00000

0.05000

0.10000

0.15000

0.20000

0.25000

0.30000

0 50 100 150 200 250 300 350 400

Days After Detonation

Fraction of Total Dose

Figure A-13. Strontium-90 fraction of total ingestion dose to red bone marrow. 

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y = -4E-08x2

 + 0.0002x + 0.196

R2

 = 1

0

0.1

0.2

0.3

0.4

0.5

0.6

0 500 1000 1500 2000 2500 3000 3500 4000

Days After Detonation

Fraction of Total Dose

Figure A-14. Strontium-90 fraction of total ingestion dose to red bone marrow. 

Figure A-15. Strontium-90 fraction of total ingestion dose. 

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y = -2E-09x2

 + 3E-05x - 0.0022

R2

 = 0.9908

0.00

0.01

0.02

0.03

0.04

0.05

0.06

0.07

0.08

0.09

0 500 1000 1500 2000 2500 3000 3500 4000

Days After Detonation

Fraction of Total Dose

Figure A-16. Strontium-90 fraction of ingestion dose to the upper large intestine. 

y = 4E-05x - 0.0024

R2

 = 0.9929

0.00

0.02

0.04

0.06

0.08

0.10

0.12

0.14

0.16

0.18

0 500 1000 1500 2000 2500 3000 3500 4000

Days After Detonation

Fraction of Total Dose

Lower Large Intestine Colon Linear (Colon)

Figure A-17. Strontium-90 fraction of total ingestion dose. 

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ingestion dose is delivered by 90Sr and 137Cs; their dose accounted for by the ingestion intakes in 

Table A-12. 

Using the relationships shown in Figures A-12 through A-17, short-lived fission and activation 

ingestion correction factors were developed (Table A-14). As with the inhalation fission and activation 

factors, these correction factors are multiplied by the organ-specific 90Sr annual doses (from the 

intakes in Table A-12 [i.e., 55.42 Bq/yr]) to calculate additional ingestion dose from short-lived fission

and products. 

Table A-14. Organ-specific ingestion fission and activation correction factors. 

Organ Correction factor

Year 1963 1964 1965 1966 1967 1968 1969 1970 1971 1972

Adrenals Breast 

Brain Skin Bladder

Stomach Kidneys Muscle 

Pancreas Brain Esophagus

SI Liver Ovaries ET ET1

ET2 LM(ET) LN(TH) Lungs 

Skin Spleen Testes 

Thymus Thyroid Uterus

Gall Bladder

416 219 155 123 106 95.0 88.1 84.0 82.0 81.8

ULI 514 184 95.2 58.2 39.3 18.6 16.3 14.6 13.4 12.4

Bone Surface / RBM 3.8 3.1 2.7 2.5 2.3 2.3 2.3 2.3 2.3 2.3

LLI / Colon 417 208 138 25.6 20.4 17.4 14.6 12.8 11.3 10.2

To simplify the application of organ-specific ingestion dose from short-lived fission and activation 

products, the IMBA computer program was used to determine the organ specific annual doses for the 

90Sr intake of 55.42 Bq/yr for 1963 through 1972 (Table A-15). These doses are then multiplied by the 

Table A-14 organ-specific correction factors to provide the annual doses from ingestion of short-lived 

fission and activation products (Table A-16). 

Table A-15. Organ-specific annual ingestion doses (rem) for the 90Sr intake of 55.42 Bq/yr. Organ 1963 1964 1965 1966 1967 1968 1969 1970 1971 1972

Adrenals 2.11E-06 2.47E-06 2.64E-06 2.77E-06 2.89E-06 3.00E-06 3.08E-06 3.16E-06 3.22E-06 3.28E-06

Urinary 

bladder 5.68E-06 6.42E-06 6.67E-06 6.87E-06 7.04E-06 7.18E-06 7.31E-06 7.41E-06 7.50E-06 7.58E-06

Bone surface 1.58E-04 3.74E-04 5.47E-04 6.98E-04 8.32E-04 9.50E-04 1.06E-03 1.15E-03 1.23E-03 1.31E-03

Brain 2.11E-06 2.47E-06 2.64E-06 2.77E-06 2.89E-06 3.00E-06 3.08E-06 3.16E-06 3.22E-06 3.28E-06

Breast 2.11E-06 2.47E-06 2.64E-06 2.77E-06 2.89E-06 3.00E-06 3.08E-06 3.16E-06 3.22E-06 3.28E-06

Esophagus 2.11E-06 2.47E-06 2.64E-06 2.77E-06 2.89E-06 3.00E-06 3.08E-06 3.16E-06 3.22E-06 3.28E-06

Stomach 3.42E-06 3.78E-06 3.95E-06 4.09E-06 4.21E-06 4.31E-06 4.40E-06 4.47E-06 4.54E-06 4.59E-06

Small intestine 4.71E-06 5.07E-06 5.24E-06 5.38E-06 5.50E-06 5.60E-06 5.69E-06 5.76E-06 5.83E-06 5.88E-06

Upper large 

intestine 3.42E-05 3.43E-05 3.43E-05 3.43E-05 3.43E-05 3.43E-05 3.43E-05 3.43E-05 3.43E-05 3.43E-05

Lower large 

intestine 1.45E-04 1.45E-04 1.45E-04 1.45E-04 1.45E-04 1.45E-04 1.45E-04 1.45E-04 1.45E-04 1.45E-04

Colon 4.36E-03 2.26E-03 1.51E-03 1.14E-03 9.16E-04 7.71E-04 6.61E-04 5.84E-04 5.21E-04 4.71E-04

Kidneys 2.11E-06 2.47E-06 2.64E-06 2.77E-06 2.89E-06 3.00E-06 3.08E-06 3.16E-06 3.22E-06 3.28E-06

Liver 2.11E-06 2.47E-06 2.64E-06 2.77E-06 2.89E-06 3.00E-06 3.08E-06 3.16E-06 3.22E-06 3.28E-06

Muscle 2.11E-06 2.47E-06 2.64E-06 2.77E-06 2.89E-06 3.00E-06 3.08E-06 3.16E-06 3.22E-06 3.28E-06

Ovaries 2.11E-06 2.47E-06 2.64E-06 2.77E-06 2.89E-06 3.00E-06 3.08E-06 3.16E-06 3.22E-06 3.28E-06

Pancreas 2.11E-06 2.47E-06 2.64E-06 2.77E-06 2.89E-06 3.00E-06 3.08E-06 3.16E-06 3.22E-06 3.28E-06

Red marrow 1.08E-04 2.66E-04 3.87E-04 4.86E-04 5.69E-04 6.37E-04 6.94E-04 7.42E-04 7.81E-04 8.14E-04

Extrathoracic 2.11E-06 2.47E-06 2.64E-06 2.77E-06 2.89E-06 3.00E-06 3.08E-06 3.16E-06 3.22E-06 3.28E-06

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Organ 1963 1964 1965 1966 1967 1968 1969 1970 1971 1972

airways

Lungs 2.11E-06 2.47E-06 2.64E-06 2.77E-06 2.89E-06 3.00E-06 3.08E-06 3.16E-06 3.22E-06 3.28E-06

Skin 2.11E-06 2.47E-06 2.64E-06 2.77E-06 2.89E-06 3.00E-06 3.08E-06 3.16E-06 3.22E-06 3.28E-06

Spleen 2.11E-06 2.47E-06 2.64E-06 2.77E-06 2.89E-06 3.00E-06 3.08E-06 3.16E-06 3.22E-06 3.28E-06

Testes 2.11E-06 2.47E-06 2.64E-06 2.77E-06 2.89E-06 3.00E-06 3.08E-06 3.16E-06 3.22E-06 3.28E-06

Thymus 2.11E-06 2.47E-06 2.64E-06 2.77E-06 2.89E-06 3.00E-06 3.08E-06 3.16E-06 3.22E-06 3.28E-06

Thyroid 2.11E-06 2.47E-06 2.64E-06 2.77E-06 2.89E-06 3.00E-06 3.08E-06 3.16E-06 3.22E-06 3.28E-06

Uterus 2.11E-06 2.47E-06 2.64E-06 2.77E-06 2.89E-06 3.00E-06 3.08E-06 3.16E-06 3.22E-06 3.28E-06

Table A-16. Organ specific doses (rem) from ingestion of fission and activation products. Organ 1963 1964 1965 1966 1967 1968 1969 1970 1971 1972

Adrenals 8.78E-04 6.12E-04 4.75E-04 4.09E-04 3.73E-04 3.49E-04 3.32E-04 3.20E-04 3.13E-04 1.40E-04

Bladder 2.36E-03 1.55E-03 1.15E-03 9.50E-04 8.45E-04 7.76E-04 7.31E-04 7.02E-04 6.85E-04 2.21E-04

Brain 8.78E-04 6.12E-04 4.75E-04 4.09E-04 3.73E-04 3.49E-04 3.32E-04 3.20E-04 3.13E-04 1.40E-04

Breast 8.78E-04 6.12E-04 4.75E-04 4.09E-04 3.73E-04 3.49E-04 3.32E-04 3.20E-04 3.13E-04 1.40E-04

Gall bladder 8.78E-04 6.12E-04 4.75E-04 4.09E-04 3.73E-04 3.49E-04 3.32E-04 3.20E-04 3.13E-04 1.40E-04

Heart wall 8.78E-04 6.12E-04 4.75E-04 4.09E-04 3.73E-04 3.49E-04 3.32E-04 3.20E-04 3.13E-04 1.40E-04

Kidney 8.78E-04 6.12E-04 4.75E-04 4.09E-04 3.73E-04 3.49E-04 3.32E-04 3.20E-04 3.13E-04 1.40E-04

Liver 8.78E-04 6.12E-04 4.75E-04 4.09E-04 3.73E-04 3.49E-04 3.32E-04 3.20E-04 3.13E-04 1.40E-04

Muscle 8.78E-04 6.12E-04 4.75E-04 4.09E-04 3.73E-04 3.49E-04 3.32E-04 3.20E-04 3.13E-04 1.40E-04

Ovaries 8.78E-04 6.12E-04 4.75E-04 4.09E-04 3.73E-04 3.49E-04 3.32E-04 3.20E-04 3.13E-04 1.40E-04

Pancreas 8.78E-04 6.12E-04 4.75E-04 4.09E-04 3.73E-04 3.49E-04 3.32E-04 3.20E-04 3.13E-04 1.40E-04

Testes 8.78E-04 6.12E-04 4.75E-04 4.09E-04 3.73E-04 3.49E-04 3.32E-04 3.20E-04 3.13E-04 1.40E-04

Thyroid 8.78E-04 6.12E-04 4.75E-04 4.09E-04 3.73E-04 3.49E-04 3.32E-04 3.20E-04 3.13E-04 1.40E-04

RBM 4.12E-04 2.66E-04 3.87E-04 4.86E-04 5.69E-04 6.37E-04 6.94E-04 7.42E-04 7.81E-04 8.14E-04

Bone surface 6.01E-04 1.31E-03 1.75E-03 2.09E-03 2.35E-03 2.56E-03 2.76E-03 2.94E-03 3.11E-03 3.05E-03

Stomach 1.42E-03 9.00E-04 6.79E-04 5.71E-04 5.13E-04 4.74E-04 4.48E-04 4.31E-04 4.21E-04 1.41E-04

SI 1.96E-03 1.18E-03 8.79E-04 7.29E-04 6.49E-04 5.96E-04 5.61E-04 5.39E-04 5.27E-04 1.43E-04

ULI 1.53E-02 5.87E-03 3.11E-03 1.96E-03 1.36E-03 7.26E-04 6.30E-04 5.64E-04 5.14E-04 1.34E-04

LLI 4.84E-02 2.50E-02 1.66E-02 3.47E-03 2.66E-03 2.23E-03 1.92E-03 1.68E-03 1.49E-03 1.34E-03

Skin 8.78E-04 6.12E-04 4.75E-04 4.09E-04 3.73E-04 3.49E-04 3.32E-04 3.20E-04 3.13E-04 1.40E-04

Spleen 8.78E-04 6.12E-04 4.75E-04 4.09E-04 3.73E-04 3.49E-04 3.32E-04 3.20E-04 3.13E-04 1.40E-04

Thymus 2.11E-06 2.47E-06 2.64E-06 2.77E-06 2.89E-06 3.00E-06 3.08E-06 3.16E-06 3.22E-06 3.28E-06

Uterus 8.78E-04 6.12E-04 4.75E-04 4.09E-04 3.73E-04 3.49E-04 3.32E-04 3.20E-04 3.13E-04 1.40E-04

ET 8.78E-04 6.12E-04 4.75E-04 4.09E-04 3.73E-04 3.49E-04 3.32E-04 3.20E-04 3.13E-04 1.40E-04

Lung 8.78E-04 6.12E-04 4.75E-04 4.09E-04 3.73E-04 3.49E-04 3.32E-04 3.20E-04 3.13E-04 1.40E-04

Colon 2.80E-02 1.45E-02 9.66E-03 2.07E-03 1.59E-03 1.33E-03 1.15E-03 1.01E-03 8.94E-04 8.06E-04

ET1 8.78E-04 6.12E-04 4.75E-04 4.09E-04 3.73E-04 3.49E-04 3.32E-04 3.20E-04 3.13E-04 1.40E-04

ET2 8.78E-04 6.12E-04 4.75E-04 4.09E-04 3.73E-04 3.49E-04 3.32E-04 3.20E-04 3.13E-04 1.40E-04

LN(ET) 8.78E-04 6.12E-04 4.75E-04 4.09E-04 3.73E-04 3.49E-04 3.32E-04 3.20E-04 3.13E-04 1.40E-04

LN(TH) 8.78E-04 6.12E-04 4.75E-04 4.09E-04 3.73E-04 3.49E-04 3.32E-04 3.20E-04 3.13E-04 1.40E-04

Esophagus 2.89E-04 2.07E-04 1.62E-04 1.41E-04 1.29E-04 1.21E-04 1.16E-04 1.13E-04 1.11E-04 5.02E-05

A.8 Instruction to Dose Reconstructors for Assignment of Environmental Intakes

With the exception of cases that can be worked with the use of the bounding assumptions from 

ORAUT-OTIB-0002 (ORAUT 2007d) or ORAUT-OTIB-0018 (ORAUT 2005), environmental inhalation 

and ingestion intakes in Tables A-7 and A-12, respectively, should be applied for all cases. In 

addition, for applicable years of employment and affected organs, the annual doses from 30- to 250-

keV photons in Tables A-10 and A-16 should be applied to account for dose from inhalation and 

ingestion of short-lived fission and activation products. These intakes and resultant doses should be 

entered into the IREP analysis with a constant distribution because they are considered to be 

reasonable overestimates of the actual intakes and doses.

ATTACHMENT A

AMBIENT ENVIRONMENTAL INTAKES AT THE NEVADA TEST SITE

BASED ON AIR S AMPLING AND SOIL CONTAMINATION DATA 

Page 27 of 27
Document No. ORAUT-TKBS-0008-4 Revision No. 01 Effective Date: 05/27/2008 Page 74 of 77

ATTACHMENT B

TOTAL ANNUAL ORGAN DOSES FROM 30 YEARS OF INHALATION AND INGESTION INTAKES 

AND SOURCES OF OVERESTIMATED ORGAN DOSE

Page 1 of 4 

In the previous sections, methods were presented to account for organ dose from ambient 

environmental inhalation and ingestion intakes of radioactive materials in NTS atmosphere and soils. 

To illustrate the importance of these pathways, the IMBA computer program was used to determine 

organ doses from 30 years of the inhalation intakes in Table A-7 and the ingestion intakes in 

Table A-12. These 30-year organ doses and the fractional contribution for each of the radionuclides 

persistent in the NTS soils are presented in Table B-1. To correct for exposure to short-lived fission 

and activation, the dose reconstructor should add annual doses that were greater than 0.001 rem 

from Tables A-10 and A-16 as 30- to 250-keV photons with a constant distribution. 
Document No. ORAUT-TKBS-0008-4 Revision No. 01 Effective Date: 05/27/2008 Page 75 of 77

ATTACHMENT B

TOTAL ANNUAL ORGAN DOSES FROM 30 YEARS OF INHALATION AND INGESTION INTAKES 

AND SOURCES OF OVERESTIMATED ORGAN DOSE

Page 2 of 4 

Table B-1. Total annual organ doses greater than 0.0001 rem from 30 years of inhalation and ingestion intakes (rem). Year Liver Ovaries Testes RBM Bone surface LLI ET Lung Colon ET1 ET2 LN(ET) LN(TH)

1963 Alpha 0.001 0.006 0.001 0.004 0.007 0.001 0.005 0.004 0.001

1964 Alpha 0.003 0.002 0.017 0.001 0.009 0.006 0.001 0.003 0.009 0.001 0.003

1965 Alpha 0.005 0.002 0.026 0.001 0.009 0.005 0.001 0.002 0.009 0.001 0.004

1966 Alpha 0.006 0.003 0.033 0.001 0.009 0.004 0.001 0.002 0.009 0.002 0.006

1967 Alpha 0.008 0.003 0.040 0.001 0.009 0.004 0.001 0.002 0.009 0.003 0.007

1968 Alpha 0.009 0.004 0.046 0.001 0.008 0.004 0.001 0.002 0.008 0.003 0.008

1969 Alpha 0.010 0.004 0.052 0.001 0.008 0.004 0.001 0.002 0.008 0.004 0.009

1970 Alpha 0.011 0.004 0.058 0.001 0.008 0.004 0.001 0.002 0.008 0.004 0.011

1971 Alpha 0.012 0.005 0.065 0.001 0.008 0.004 0.001 0.002 0.008 0.005 0.012

1972 Alpha 0.013 0.001 0.001 0.005 0.070 0.001 0.008 0.004 0.001 0.002 0.008 0.005 0.013

1973 Alpha 0.015 0.001 0.001 0.005 0.076 0.001 0.008 0.004 0.001 0.002 0.008 0.006 0.014

1974 Alpha 0.016 0.001 0.001 0.006 0.082 0.001 0.008 0.004 0.001 0.002 0.008 0.006 0.016

1975 Alpha 0.017 0.001 0.001 0.006 0.088 0.001 0.008 0.005 0.001 0.002 0.008 0.007 0.017

1976 Alpha 0.018 0.001 0.001 0.006 0.093 0.001 0.008 0.005 0.001 0.002 0.008 0.007 0.018

1977 Alpha 0.019 0.001 0.001 0.006 0.098 0.001 0.008 0.005 0.001 0.002 0.008 0.007 0.019

1978 Alpha 0.020 0.002 0.002 0.006 0.104 0.001 0.008 0.005 0.001 0.002 0.008 0.008 0.020

1979 Alpha 0.021 0.002 0.002 0.007 0.109 0.001 0.008 0.005 0.001 0.002 0.008 0.008 0.022

1980 Alpha 0.021 0.002 0.002 0.007 0.114 0.001 0.008 0.005 0.001 0.002 0.008 0.008 0.023

1981 Alpha 0.022 0.002 0.002 0.007 0.119 0.001 0.008 0.005 0.001 0.002 0.008 0.009 0.024

1982 Alpha 0.023 0.002 0.002 0.007 0.124 0.001 0.008 0.005 0.001 0.002 0.008 0.009 0.025

1983 Alpha 0.024 0.002 0.002 0.007 0.129 0.001 0.008 0.005 0.001 0.002 0.008 0.009 0.026

1984 Alpha 0.025 0.002 0.002 0.008 0.134 0.001 0.008 0.005 0.001 0.002 0.008 0.009 0.027

1985 Alpha 0.026 0.002 0.002 0.008 0.139 0.001 0.008 0.005 0.001 0.002 0.008 0.010 0.028

1986 Alpha 0.027 0.002 0.002 0.008 0.144 0.001 0.008 0.005 0.001 0.002 0.008 0.010 0.029

1987 Alpha 0.027 0.002 0.002 0.008 0.149 0.001 0.008 0.005 0.001 0.002 0.008 0.010 0.030

1988 Alpha 0.028 0.002 0.002 0.008 0.153 0.001 0.008 0.005 0.001 0.002 0.008 0.010 0.031

1989 Alpha 0.029 0.002 0.002 0.008 0.158 0.001 0.008 0.005 0.001 0.002 0.008 0.011 0.032

1990 Alpha 0.030 0.002 0.002 0.008 0.162 0.001 0.008 0.005 0.001 0.002 0.008 0.011 0.033

1991 Alpha 0.030 0.003 0.003 0.009 0.167 0.001 0.008 0.005 0.001 0.002 0.008 0.011 0.034

1992 Alpha 0.031 0.003 0.003 0.009 0.171 0.001 0.008 0.005 0.001 0.002 0.008 0.011 0.035

1993 Alpha 0.031 0.003 0.003 0.009 0.172 0.006 0.003 0.006 0.011 0.035

1994 Alpha 0.030 0.003 0.003 0.008 0.169 0.004 0.002 0.004 0.011 0.035

1995 Alpha 0.030 0.003 0.003 0.008 0.166 0.003 0.002 0.003 0.011 0.035

1996 Alpha 0.029 0.003 0.003 0.007 0.163 0.002 0.001 0.002 0.011 0.035

1997 Alpha 0.028 0.002 0.003 0.007 0.160 0.001 0.001 0.001 0.011 0.035

1998 Alpha 0.028 0.002 0.002 0.006 0.158 0.010 0.034

1999 Alpha 0.027 0.002 0.002 0.006 0.155 0.010 0.034

2000 Alpha 0.026 0.002 0.002 0.006 0.153 0.010 0.033
Document No. ORAUT-TKBS-0008-4 Revision No. 01 Effective Date: 05/27/2008 Page 76 of 77

ATTACHMENT B

TOTAL ANNUAL ORGAN DOSES FROM 30 YEARS OF INHALATION AND INGESTION INTAKES 

AND SOURCES OF OVERESTIMATED ORGAN DOSE

Page 3 of 4 

Year Liver Ovaries Testes RBM Bone surface LLI ET Lung Colon ET1 ET2 LN(ET) LN(TH)

2001 Alpha 0.026 0.002 0.002 0.006 0.151 0.009 0.033

2002 Alpha 0.025 0.002 0.002 0.005 0.148 0.009 0.032

2003 Alpha 0.025 0.002 0.002 0.005 0.146 0.009 0.032

2004 Alpha 0.024 0.002 0.002 0.005 0.144 0.008 0.031

2005 Alpha 0.024 0.002 0.002 0.005 0.142 0.008 0.031

2006 Alpha 0.023 0.002 0.002 0.005 0.140 0.008 0.030

2007 Alpha 0.023 0.002 0.002 0.005 0.138 0.007 0.029

1963 Photons E>250 keV 0.002

1964 Photons E>250 keV 0.001

1965 Photons E>250 keV 0.001

1963 Electrons E>15 keV 0.002 

1964 Electrons E>15 keV 0.001

1965 Electrons E>15 keV 

1966 Electrons E>15 keV

1967 Electrons E>15 keV 0.001

1968 Electrons E>15 keV 0.001

1969 Electrons E>15 keV 0.001

1970 Electrons E>15 keV 0.001

1971 Electrons E>15 keV 0.001

1972 Electrons E>15 keV 0.002 

1973 Electrons E>15 keV 0.002

1974 Electrons E>15 keV 0.001 0.002 

1975 Electrons E>15 keV 0.001 0.002

1976 Electrons E>15 keV 0.001 0.002

1977 Electrons E>15 keV 0.001 0.002

1978 Electrons E>15 keV 0.001 0.002

1979 Electrons E>15 keV 0.001 0.002

1980 Electrons E>15 keV 0.001 0.002

1981 Electrons E>15 keV 0.001 0.002 

1982 Electrons E>15 keV 0.001 0.002

1983 Electrons E>15 keV 0.001 0.002 

1984 Electrons E>15 keV 0.001 0.002

1985 Electrons E>15 keV 0.001 0.002

1986 Electrons E>15 keV 0.001 0.002

1987 Electrons E>15 keV 0.001 0.002

1988 Electrons E>15 keV 0.001 0.002

1989 Electrons E>15 keV 0.001 0.002

1990 Electrons E>15 keV 0.001 0.002

1991 Electrons E>15 keV 0.001 0.002
Document No. ORAUT-TKBS-0008-4 Revision No. 01 Effective Date: 05/27/2008 Page 77 of 77

ATTACHMENT B

TOTAL ANNUAL ORGAN DOSES FROM 30 YEARS OF INHALATION AND INGESTION INTAKES 

AND SOURCES OF OVERESTIMATED ORGAN DOSE

Page 4 of 4 

Year Liver Ovaries Testes RBM Bone surface LLI ET Lung Colon ET1 ET2 LN(ET) LN(TH)

1992 Electrons E>15 keV 0.001 0.002

1993 Electrons E>15 keV 0.002

1994 Electrons E>15 keV 0.002

1995 Electrons E>15 keV 0.002

1996 Electrons E>15 keV 0.001 

1997 Electrons E>15 keV 0.001

1998 Electrons E>15 keV 0.001 

1999 Electrons E>15 keV 0.001

2000 Electrons E>15 keV 0.001