Page 1 of 94

DOE Review Release: 03/06/2020

Nevada Test Site – Occupational 

Environmental Dose

ORAUT-TKBS-0008-4 Rev. 04

Effective Date: 02/19/2020

Supersedes: Revision 03

Subject Expert(s): Eugene M. Rollins

Document Owner 

Approval: Signature on File

Eugene M. Rollins, Document Owner

Approval Date: 02/07/2020

Concurrence: Signature on File

John M. Byrne, Objective 1 Manager

Concurrence Date: 02/06/2020

Concurrence: Signature on File

Joel Arana, Objective 3 Representative

Concurrence Date: 02/06/2020

Concurrence: Vickie S. Short Signature on File for

Kate Kimpan, Project Director

Concurrence Date: 02/06/2020

Approval: Signature on File

Timothy D. Taulbee, Associate Director for Science

Approval Date: 02/19/2020

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

REVISION AND DISCARD / DESTROY ALL COPIES OF THE PRIOR REVISION.

New Total Rewrite Revision Page Change
Document No. ORAUT-TKBS-0008-4 Revision No. 04 Effective Date: 02/19/2020 Page 2 of 94

 

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:

Signature on File 12/05/2006

Eugene M. Rollins, Document Owner

Signature on File 10/24/2006

John M. Byrne, Task 3 Manager

Signature on File 10/24/2006

Edward F. Maher, Task 5 Manager

Signature on File 11/15/2006

Kate Kimpan, Project Director

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

James W. Neton, Associate Director for Science

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-OTIB0018 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 ID: 37805) 
Document No. ORAUT-TKBS-0008-4 Revision No. 04 Effective Date: 02/19/2020 Page 3 of 94

EFFECTIVE 

DATE

REVISION 

NUMBER DESCRIPTION

05/27/2008 

(continued)

01 

(continued)

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.

04/15/2010 02 Revision initiated to incorporate definitions and directions for dose 

reconstruction for nonpresumptive cancers that are excluded from 

the 1963 through 1992 Special Exposure Cohort. Added Section 

4.1.2. Added years 2002 through 2008 ambient external doses to 

Tables 4-15 and 4-16. Incorporates formal internal and NIOSH 

review comments. Updated ORAUT references. Training required: 

As determined by the Objective Manager. Initiated by Eugene M. 

Rollins. 

08/24/2012 03 Revision extended measured ambient radiation doses in Tables 4-15 

and 4-16 to 2010 in Section 4.3.1. Also in Section 4.3.1, clarification 

was added to ensure that ambient dose was not to be assigned for 

years when no data was available; in particular, prior to 1958. 

Instructions clarified that for unmonitored workers likely to be 

exposed prior to 1958, the 50% coworker dose in Table 6-11 of 

ORAUT-TKBS-0008-6 (the NTS External TBD) should be assigned. 

Incorporates formal internal and NIOSH review comments. Training 

required: As determined by the Objective Manager. Initiated by 

Eugene M. Rollins. 

02/19/2020 04 Clarified in Section 4.3.1 that site maximum values would be entered 

into IREP as constants. Clarified in Section 4.4.4 that 220Rn should 

be assumed to be present at twice the concentrations of 222Rn in 

Table 4-20. Added discussion in Sections 4.3.1 to provide maximum 

and best-estimate methods, respectively, for estimating onsite 

ambient external doses. In Section 4.5.2, clarified that the intake 

values in Tables 4-7 and 4-11 are to be considered bounding and 

should be applied in IREP as constants. Also for a best estimate, 

10% of the intakes in Tables 4-7 and 4-11 should be assigned. The 

best-estimate intakes should be applied in IREP as a lognormal 

distribution with a GSD of 3.0. Added to Section 4.4.4 that doses for 

all organs should be adjusted in accordance with DCAS-TIB-011 for 

radon (DCAS 2018). Added instructions in Section 4.4.4 for the 

application of underground occupancy factors for work in the gravel 

gerties and tunnels. Incorporates formal internal and NIOSH review 

comments. Constitutes a total rewrite of the document. Training 

required: As determined by the Objective Manager. Initiated by 

Eugene M. Rollins. 
Document No. ORAUT-TKBS-0008-4 Revision No. 04 Effective Date: 02/19/2020 Page 4 of 94

TABLE OF CONTENTS

SECTION TITLE PAGE

Acronyms and Abbreviations ............................................................................................................... 7

4.1 Introduction .............................................................................................................................. 9

4.1.1 Purpose...................................................................................................................... 10

4.1.2 Scope......................................................................................................................... 10

4.1.3 Special Exposure Cohort............................................................................................ 10

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........................................... 16

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

4.2.1.2.3 Annual Intakes of Other Radionuclides...................................... 19

4.2.1.2.4 Scaling Factors for Inhalation Intakes........................................ 19

4.2.1.2.5 Correction for Resuspension for Early Times after 

Atmospheric Tests..................................................................... 27

4.2.1.2.6 Correction for Inhalation Dose from Short-Lived Fission 

and Activation Products............................................................. 28

4.2.2 Annual Intakes from Ingestion .................................................................................... 28

4.3 Occupational External Dose................................................................................................... 32

4.3.1 Ambient Radiation ...................................................................................................... 32

4.3.2 Releases of Noble Gases ........................................................................................... 33

4.4 Internal Inhalation and Ingestion Dose to Underground Workers............................................ 37

4.4.1 Underground Activities................................................................................................ 37

4.4.2 Radon Measurements ................................................................................................ 38

4.4.3 Underground Radon Concentrations .......................................................................... 39

4.4.4 Underground Worker Exposure to Radon................................................................... 40

4.4.4.1 Gravel Gerties.............................................................................................. 42

4.4.4.2 Uncertainty Associated with Radon Measurements...................................... 42

4.4.5 Underground Worker Exposure to Other Radionuclides ............................................. 43

4.5 Instruction to Dose Reconstructors for Assignment of External Ambient Doses and 

Environmental Intakes and Resultant Doses.......................................................................... 43

4.5.1 Assigning External Ambient Dose............................................................................... 43

4.5.2 Assigning Internal Ambient Dose................................................................................ 44

4.6 Attributions and Annotations .................................................................................................. 44

References ........................................................................................................................................ 45
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Glossary ............................................................................................................................................ 53

ATTACHMENT A AMBIENT ENVIRONMENTAL INTAKES AT THE NEVADA TEST SITE 

BASED ON AIR SAMPLING AND SOIL CONTAMINATION DATA................. 59

ATTACHMENT B TOTAL ANNUAL ORGAN DOSES FROM 30 YEARS OF INHALATION 

AND INGESTION INTAKES AND SOURCES OF OVERESTIMATED 

ORGAN DOSE................................................................................................ 91

LIST OF TABLES

TABLE TITLE PAGE

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

average annual organ dose.................................................................................................... 17

4-2 Atmospheric concentrations of 239Pu for sampled areas......................................................... 20

4-3 Annual inhalation intakes from 239Pu for sampled areas ......................................................... 22

4-4 Inventory of contaminated soil................................................................................................ 24

4-5 Radionuclide areal soil deposition decay, corrected to 1963 .................................................. 25

4-6 Abundance of radionuclides in NTS soils in relation to 239Pu decay, corrected to 1963 .......... 26

4-7 Scaled inhalation intakes corrected for early resuspension .................................................... 29

4-8 Organ-specific inhalation dose fission and activation product correction factors..................... 29

4-9 Inhalation dose from short-lived fission and activation products ............................................. 29

4-10 Radionuclide soil concentration by area................................................................................. 30

4-11 Area-specific and maximum annual ingestion rates................................................................ 30

4-12 Organ-specific ingestion fission and activation product correction factors .............................. 31

4-13 Organ-specific annual ingestion doses for the 90Sr intake of 55.42 Bq/year ........................... 31

4-14 Organ-specific doses from ingestion of fission and activation products .................................. 32

4-15 Ambient radiation by area ...................................................................................................... 34

4-16 85Kr average background and the Area 20 net concentrations and annual organ doses 

from 1983 through 1993......................................................................................................... 36

4-17 Results of experiment to determine the effects of ventilation conditions on RDCs in the 

G-Tunnel inclined drift............................................................................................................ 39

4-18 RDCs for N-Tunnel in 1991 and 1992 .................................................................................... 39

4-19 RDCs for P-Tunnel in 1991 and 1992 .................................................................................... 40

4-20 Annual 222Rn exposures and uncertainties for internal dose reconstruction for miners and 

tunnel workers........................................................................................................................41

A-1 Atmospheric concentrations of 239Pu for sampled areas......................................................... 63

A-2 Annual inhalation intakes from 239Pu for sampled areas ......................................................... 65

A-3 Inventory of contaminated soil................................................................................................ 67

A-4 Radionuclide areal soil deposition.......................................................................................... 68

A-5 Radionuclide areal soil deposition decay, corrected to 1963 .................................................. 68

A-6 Abundance of radionuclides in NTS soils relative to 239Pu decay, corrected to 1963 .............. 69

A-7 Scaled inhalation intakes corrected for early resuspension in 1963........................................ 71

A-8 Far-field refractory enrichment ratios...................................................................................... 73

A-9 Organ-specific inhalation dose fission and activation product correction factors..................... 79

A-10 Inhalation dose from short-lived fission and activation products ............................................. 80

A-11 Radionuclide soil concentration by area................................................................................. 81

A-12 Area-specific and maximum annual ingestion rates................................................................ 82

A-13 Annual organ doses from ingestion of 238,239Pu and 241Am...................................................... 83

A-14 Organ-specific ingestion fission and activation product correction factors .............................. 87

A-15 Organ-specific annual ingestion doses for the 90Sr intake of 55.42 Bq/yr ............................... 88

A-16 Organ-specific doses from ingestion of fission and activation products .................................. 89
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B-1 Total annual organ doses greater than or equal to 0.001 rem from 30 years of inhalation 

and ingestion intakes .............................................................................................................92

LIST OF FIGURES

FIGURE TITLE PAGE

4-1 Air sampling stations in 1997 ................................................................................................. 16

4-2 Resuspension factor as a function of time after initial deposition............................................ 28

A-1 Resuspension factor as a function of time after initial deposition............................................ 70

A-2 SMALL BOY – 90Sr fraction of total dose to lungs................................................................... 74

A-3 LITTLE FELLER I – 90Sr fraction of total dose to lungs........................................................... 74

A-4 TURK – 90Sr fraction of total dose to lungs............................................................................. 75

A-5 Strontium-90 fraction of total dose.......................................................................................... 76

A-6 Strontium-90 fraction of total dose.......................................................................................... 76

A-7 Strontium-90 fraction of total dose.......................................................................................... 77

A-8 Strontium-90 fraction of total dose.......................................................................................... 77

A-9 Strontium-90 fraction of total dose to the liver ........................................................................ 78

A-10 Strontium-90 fraction of total dose to the red bone marrow .................................................... 78

A-11 Strontium-90 fraction of total dose to the bone surface .......................................................... 79

A-12 Strontium-90 fraction of total ingestion dose .......................................................................... 84

A-13 Strontium-90 fraction of total ingestion dose to red bone marrow........................................... 84

A-14 Strontium-90 fraction of total ingestion dose to red bone marrow........................................... 85

A-15 Strontium-90 fraction of total ingestion dose .......................................................................... 85

A-16 Strontium-90 fraction of ingestion dose to the upper large intestine ....................................... 86

A-17 Strontium-90 fraction of total ingestion dose .......................................................................... 86
Document No. ORAUT-TKBS-0008-4 Revision No. 04 Effective Date: 02/19/2020 Page 7 of 94

ACRONYMS AND ABBREVIATIONS

AWE Atomic Weapons Employer

Bq becquerel

CFR Code of Federal Regulations

Ci curie

cm centimeter

d day

DCF dose conversion factor

DOE U.S. Department of Energy

DOL U.S. Department of Labor

EEOICPA Energy Employees Occupational Illness Compensation Program Act of 2000

EPA U.S. Environmental Protection Agency

ET extrathoracic

ET1 extrathoracic region 1

ET2 extrathoracic region 2 (posterior nasal passage, larynx, pharynx, and mouth)

ft foot

g gram

GSD geometric standard deviation

hr hour

HT elemental tritium (tritiated gas)

HTO tritiated water vapor

ICRP International Commission on Radiological Protection

IMBA Integrated Modules for Bioassay Analysis

in. inch

IREP Interactive RadioEpidemiological Program

JASPER Joint Actinide Shock Physics Experimental Research

keV kiloelectron-volt, 1,000 electron-volts

kg kilogram

L liter

lb pound

LLI lower large intestine

LLNL Lawrence Livermore National Laboratory

LN lymph nodes

LN(ET) extrathoracic lymph nodes

LN(TH) thoracic lymph nodes (trachea, bronchi, bronchioles, alveolar ducts, and sacs)

m meter

MeV megaelectron-volt, 1 million electron-volts

mg milligram

mi mile

min minute
Document No. ORAUT-TKBS-0008-4 Revision No. 04 Effective Date: 02/19/2020 Page 8 of 94

mL milliliter

mR milliroentgen

mrem millirem

NIOSH National Institute for Occupational Safety and Health

NNSS Nevada National Security Site

NTS Nevada Test Site

ORAU Oak Ridge Associated Universities

pCi picocurie

PER program evaluation report

PERM passive environmental radon monitor

RBM red bone marrow

RDC radon daughter concentration

REECo Reynolds Electrical & Engineering Company

RPISU radon progeny integrating sampling unit

s second

SEC Special Exposure Cohort

SI small intestine

SRDB Ref ID Site Research Database Reference Identification (number)

Sv sievert

TBD technical basis document

TLD thermoluminescent dosimeter

TRU transuranic

TTR Tonopah Test Range

ULI upper large intestine

U.S.C. United States Code

wk week

WL working level

WLM working level month

yr year

μCi microcurie

μm micrometer
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4.1 INTRODUCTION

Technical basis documents (TBDs) 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 historical background information and guidance to assist in the preparation of 

dose reconstructions at particular U.S. Department of Energy (DOE) or Atomic Weapons Employer 

(AWE) facilities or categories of DOE or AWE facilities. They will be revised in the event additional 

relevant information is obtained about the affected DOE or AWE facility(ies), such as changing 

scientific understanding of operations, processes, or procedures involving radioactive materials. 

These documents may be used to assist NIOSH staff in the evaluation of Special Exposure Cohort 

(SEC) petitions and the completion of individual dose reconstructions under Part B of the Energy 

Employees Occupational Illness Compensation Program Act of 2000 (EEOICPA). 

In this document the word “facility” is used to refer to an area, building, or group of buildings that 

served a specific purpose at a DOE or AWE facility. It does not mean nor should it be equated to an 

“AWE facility” or a “DOE facility.” The term “AWE facility” is defined in EEOICPA to mean “a facility, 

owned by an atomic weapons employer, that is or was used to process or produce, for use by the 

United States, material that emitted radiation and was used in the production of an atomic weapon, 

excluding uranium mining or milling” [42 United States Code (USC) 7384I(5)]. On the other hand, a 

DOE facility is defined as “any building, structure, or premise, including the grounds upon which such 

building, structure, or premise is located—(A) in which operations are, or have been, conducted by, or 

on behalf of, the [DOE] (except for buildings, structures, premises, grounds, or operations ... 

pertaining to the Naval Nuclear Propulsion Program); and (B) with regard to which the [DOE] has or 

had—(i) a proprietary interest; or (ii) entered into a contract with an entity to provide management and 

operation, management and integration, environmental remediation services, construction, or 

maintenance services” [42 USC 7384I(12)]. The DOE determines whether a site meets the statutory 

definition of an AWE facility and the U.S. Department of Labor (DOL) determines if a site is a DOE 

facility and, if it is, designates it as such.

Under EEOICPA, a Part B cancer claim for benefits must be based on an energy employee’s eligible 

employment and occupational radiation exposure at a DOE or AWE facility during the facility’s 

designated time period and location (i.e., a “covered employee with cancer”). After DOL determines 

that a claim meets the eligibility requirements under Part B of EEOICPA, DOL transmits the claim to 

NIOSH for a dose reconstruction. EEOICPA provides, among other things, guidance on eligible 

employment and the types of radiation exposure to be included in an individual dose reconstruction. 

Under EEOICPA, eligible employment at a DOE facility includes individuals who are or were 

employed by DOE and its predecessor agencies, as well as their contractors and subcontractors at 

the facility [42 USC 7384I(11)]. Also under EEOICPA, the types of exposure to be included in dose 

reconstructions for DOE employees are those radiation exposures incurred in the performance of 

duty. As such, NIOSH includes all radiation exposures received as a condition of employment at DOE 

facilities in its dose reconstructions for covered employees, which may include radiation exposures 

related to the Naval Nuclear Propulsion Program at DOE facilities, if applicable. This is because 

NIOSH does not determine the fraction of total measured radiation exposure at a DOE facility that is 

contributed by the Naval Nuclear Propulsion Program at the DOE facility during a specified period of 

time for inclusion in dose reconstruction.

NIOSH does not consider the following types of exposure as those incurred in the performance of 

duty as a condition of employment at a DOE facility. Therefore these exposures are not included in 

dose reconstructions for covered employees [NIOSH 2010]: 

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

conventional structures, and

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

reasons. 
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4.1.1 Purpose

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

environmental dose is the dose individuals received at the Nevada Test Site (NTS), now known as the 

Nevada National Security Site (NNSS), 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 the aboveground and underground working 

environments.

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 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 and Kordas 1983, 

1985; McArthur and Mead 1987, 1988, 1989; McArthur 1991). Because the air monitoring data were 

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 NTS 

soils (e.g., 241Am, 60Co, 137Cs, 90Sr, 152Eu, 154Eu, 155Eu) are scaled to those of plutonium based on their

relative abundance in NTS soils. Ingestion intakes can be estimated by assuming consumption of 

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

relative abundances of the long-lived radionuclides in 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 were not underestimated, correction factors were developed to account 

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

provided by Hicks (1981a, 1981b, 1981c, 1981d). In addition, a correction factor was developed for 

inhalation intakes that accounts for the phenomenon of early resuspension (Anspaugh et al. 2002).

4.1.2 Scope

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 to 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 

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

environmental intakes. Attributions and annotations, indicated by bracketed callouts and used to 

identify the source, justification, or clarification of the associated information, are presented in 

Section 4.6. 

4.1.3 Special Exposure Cohort

January 27, 1951 through December 31, 1962

The Secretary of the U.S. Department of Health and Human Services has added the following class to 

the SEC (Leavitt 2006):

Department of Energy (DOE) employees or DOE contractor or subcontractor employees 

who worked at the Nevada Test Site from January 27, 1951 through December 31, 1962 
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for a number of work days aggregating at least 250 work days, either solely under this 

employment or in combination with work days within the parameters (excluding 

aggregate work day requirements) established for other classes of employees included 

in the SEC, and who were monitored or should have been monitored.

NIOSH has determined that, in the absence of bioassay results for the worker, internal doses cannot 

be reconstructed from January 27, 1951 through December 31, 1962. Based on the SEC petition 

evaluation, internal dose is not to be reconstructed for work before 1963 unless a worker has specific 

bioassay results that can be directly related to an event or incident (NIOSH 2006).

Any bioassay results in the DOE files for NTS workers before 1963 should be assumed to be valid 

because they were determined using HASL-300 (Harley 1976) procedures in the early bioassay 

program; therefore, these results can be used to evaluate internal dose. Much of the internal 

monitoring for individuals during the SEC period was event related. However, certain job 

classifications required routine monitoring. These included radiation safety personnel, industrial 

hygienist, and security personnel.

January 1, 1963 through December 31, 1992

The Secretary has added the following class to the SEC (Sebelius 2010):

All employees of the Department of Energy, its predecessor agencies, and its contractors 

and subcontractors who worked at the Nevada Test Site from January 1, 1963 through 

December 31, 1992 for a number of work days aggregating at least 250 work days, 

occurring either solely under this employment or in combination with work days within 

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

NIOSH has determined that it lacks sufficient information that would allow it to adequately estimate 

internal exposures during the period from January 1, 1963, through December 31, 1992 

(NIOSH 2010b). 

NIOSH believes that the cessation of nuclear testing, coupled with the implementation of the 1993 

NTS internal dose TBD that demonstrates NTS compliance with 10 CFR Part 835, supports NIOSH’s 

ability to bound internal dose for the evaluated class starting in 1993.

Dose reconstruction guidance in this TBD for periods before January 1, 1993, is presented to provide 

a technical basis for partial dose reconstructions for nonpresumptive cancers not covered in the SEC 

classes through December 31, 1992. Although NIOSH found that it is not possible to bound total 

internal dose for the classes, it intends to use internal and external monitoring data that might become 

available for an individual claim (and that can be interpreted using its existing dose reconstruction 

processes or procedures). Therefore, dose reconstructions for individuals who were employed at 

NTS during this period, but who do not qualify for inclusion in the SEC, can be performed using these 

data as appropriate.

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.
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4.2.1 Onsite Releases to Air

4.2.1.1 Source Description

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 mid1960s, 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 about 

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 site. 

However, much of the nonvolatile, long-lived radionuclides settled in the soils at various locations (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 after an underground test 

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

of the release (DOE 1996a):

• 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 enable 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.
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Handling of radioactive materials includes 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 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, and

• Other diffuse sources. 

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

Hardesty (2018). 

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 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 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, small amounts of contaminated water continue to drain from only one 

tunnel. 

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

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

the 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

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

Building 650, and Los Alamos National Laboratory conducted similar analyses in Building 701 at 

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

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

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

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

radioiodines, or noble gases.

Continuous stack monitoring has been conducted in the past at one facility on the NNSS, the Joint 

Actinide Shock Physics Experimental Research (JASPER) facility in Area 27 (Hardesty 2018). In 
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2013, the potential air emissions from the facility were re-evaluated and determined to result in a 

potential dose that is much less than the 0.1 mrem/yr threshold at which stack monitoring is required 

under NESHAP. Therefore, only periodic sampling is recommended to verify low emissions. In 2017, 

one sample was taken from January 18 to January 25 for this purpose. No manmade radionuclides 

were detected in the sample, which confirms continued low emissions.

4.2.1.1.7 Radioactive Waste Management Sites

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

storage of transuranic (TRU) and mixed TRU wastes as well as the Greater Confinement Disposal 

Test Unit and 12 accompanying boreholes, of which 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 houses a glovebox with high-efficiency 

particulate air filtration that is used to examine and repack 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 in 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 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 to measure radiological 

conditions throughout the site without regard to nuclear tests (Glora and Brown 1964a). That is, the 

collected data were not meant 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 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 dose reconstruction, 90Sr is assumed to be the 
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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 

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 (Lantz 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 solidstate surface barrier detectors (Lantz 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 elemental tritium (HT) and 

HTO. The sampler was portable and capable of unattended operation for as long as 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 enabled the determination of HT and HTO activities. The typical minimum detection limit 

for this analysis was 3 × 10-7 μCi/m3 (Lantz 1978a).

The 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 (EPA; 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 References section) and their 

importance to dose reconstruction for unmonitored employees.
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Figure 4-1. Air sampling stations in 1997 (Black and Townsend 1998). 

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 (Lantz 

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

samples in Area 10 (near the Sedan Crater) demonstrated the highest HTO concentrations with a 

high of 1.11 × 10-11 Bq/m3 (Lantz 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).

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

given area in a given year, unless these values were determined to be adversely affected by nearby 
 

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Table 4-1. HTO atmospheric concentrations by year and area with estimated maximum and average annual organ dose.

a,b

Year Area 1 Area 3 Area 5 Area 6 Area 10 Area 12 Area 15 Area 16 Area 18 Area 20 Area 23 Area 25

Site 

average

(Bq/m3)

Site 

maximum

(Bq/m3)

Site 

maximum 

annual 

dose

(rem)

Site 

average 

annual 

dose

(rem)

1977 ND ND ND ND 1.11E+01 ND ND ND ND ND 4.44E-02 ND 5.55E+00 1.11E+01 6.90E-05 3.50E-05

1978 ND ND 1.07E+02 ND 4.81E+00 7.03E-01 7.03E-01 ND ND ND 1.11E+00 ND 2.29E+01 1.07E+02 6.70E-04 1.40E-04

1979 ND ND ND ND ND ND ND ND ND ND ND ND 2.29E+01c 1.07E+02c 6.70E-04c 1.40E-04c

1980 ND ND 9.62E+00 ND ND ND ND ND ND ND 2.63E+00 ND 6.29E+00 9.62E+00 6.00E-05 3.80E-05

1981 ND ND 1.33E+02 ND ND ND ND ND ND ND 3.63E+01 ND 8.51E+01 1.33E+02 8.30E-04 5.30E-04

1982 2.78E+01 ND 8.51E+01 ND ND 1.63E+02 2.81E+01 ND ND ND 1.85E+02 ND 9.62E+01 1.85E+02 1.20E-03 6.10E-04

1983 3.03E+00 ND 4.07E+00 ND ND 3.29E+00 1.44E+02 ND ND ND 2.41E+01 ND 3.55E+01 1.44E+02 8.90E-04 2.20E-04

1984 6.29E+00 ND 1.96E+01 ND ND 1.22E+00 1.30E+00 ND ND ND 1.85E+02 ND 4.44E+01 1.85E+02 1.20E-03 2.70E-04

1985 5.18E+00 ND 3.18E+00 ND ND 1.74E+02 1.30E+00 ND ND ND 1.74E+01 ND 4.07E+01 1.74E+02 1.10E-03 2.50E-04

1986 4.44E+00 ND 2.07E+00 ND ND 2.44E+00 2.22E+00 ND ND ND 3.11E+01 ND 8.51E+00 3.11E+01 1.90E-04 5.30E-05

1987 1.67E+00 ND 6.29E+00 ND ND 2.33E+00 1.70E+01 ND ND ND 9.62E+00 ND 7.40E+00 1.70E+01 1.00E-04 4.60E-05

1988 1.63E+00 ND 1.44E+00 ND ND 7.40E-01 1.92E+01 ND ND ND 1.18E+00 ND 4.81E+00 1.92E+01 1.20E-04 3.00E-05

1989 6.29E+00 ND 1.15E+00 ND 4.81E-01 7.03E-01 3.03E+00 ND ND ND 1.70E-01 ND 1.96E+00 6.29E+00 3.90E-05 1.20E-05

1990 2.74E-01 ND 8.14E-01 ND 4.81E-01 1.48E-01 8.51E-01 ND ND ND 2.78E+00 7.77E-01 8.88E-01 2.78E+00 1.70E-05 5.50E-06

1991 3.37E-01 ND 1.04E+00 ND 2.33E-01 3.11E-01 6.29E-01 ND ND ND 1.59E-01 7.77E-02 4.07E-01 1.04E+00 6.50E-06 2.50E-06

1992 3.70E-01 ND 1.30E+00 ND 1.22E-01 2.29E+00 2.29E+00 ND ND ND 2.26E-01 3.26E+00 1.41E+00 3.26E+00 2.00E-05 8.80E-06

1993 3.22E-01 ND 1.04E+00 ND 2.92E-01 1.37E-01 7.40E-01 ND ND ND 1.48E-01 7.03E-02 4.07E-01 1.04E+00 6.50E-06 2.40E-06

1994 1.48E-01 ND 1.74E+00 1.89E-01 8.14E-02 9.62E-02 1.37E+00 ND ND ND 1.52E-01 1.07E-01 3.92E-01 1.74E+00 6.74E-06 2.55E-06

1995 1.37E-01 5.18E-02 7.77E-01 3.52E-02 2.59E-01 1.59E-01 3.70E-01 ND ND ND 9.62E-02 5.18E-02 2.15E-01 7.77E-01 4.90E-06 1.30E-06

1996 1.15E-01 1.52E-01 5.92E-01 2.07E+00 9.62E-01 7.40E-01 5.18E-01 ND ND ND 4.07E-02 4.07E-02 5.92E-01 2.07E+00 1.30E-05 3.60E-06

1997 8.14E-02 1.41E-01 2.78E-01 2.07E+00 1.07E+00 8.51E-01 7.40E-01 ND ND ND 4.81E-02 4.81E-02 5.92E-01 2.07E+00 1.30E-05 3.70E-06

1998 1.26E-01 ND 6.29E-01 6.66E+00 1.07E+00 2.04E+00 5.18E-01 ND ND 1.70E+01 ND ND 4.07E+00 1.70E+01 1.10E-04 2.50E-05

1999 7.40E-01 ND 6.66E-01 1.55E-01 1.52E+00 1.04E+00 9.99E-01 ND ND 2.78E+01 ND ND 4.81E+00 2.78E+01 1.70E-04 2.90E-05

2000 1.52E-01 ND 5.55E-01 ND 1.89E+00 2.00E+00 7.03E-01 ND ND 3.59E+01 ND ND 7.03E+00 3.59E+01 2.20E-04 4.30E-05

2001 1.55E-01 ND 1.33E-01 2.29E-01 1.55E+00 9.25E-01 1.41E-01 8.51E-02 4.81E-02 4.07E+01 2.63E-02 2.26E-01 4.07E+00 4.07E+01 2.50E-04 2.50E-05

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

b. ND = no data. 

c. Value not provided in annual reports; assumed same as that for 1978.
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point-source releases (e.g., Building 790 in 1983; Scoggins 1983, p. 29). If maximum values were not 

provided, the average concentrations were reported.

In addition to the annual area tritium concentrations, Table 4-1 lists 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 dose reconstruction, these average and maximum site 

concentrations have been converted to annual organ dose using the dose conversion factor (DCF) for 

inhalation of 1.80 × 10-9 rem/Bq (ICRP 1997) 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 reports until 2001. Based on the values for that year, the standard deviation would be 

typically less than 10% of the measured values. Because sampling methods changed over time and 

uncertainty estimates for previous years were unavailable, the assumption that Table 4-1 lists the 

50th-percentile expected intakes and that the 95th-percentile values are twice the Table 4-1 values is 

favorable to claimants.

For dose reconstruction, the geometric standard deviation (GSD) of the values in Table 4-1 would be 

estimated by using the following expression for a lognormal distribution (Till 2002): 

(4-1)

1

1.645

( ) ( ) 95th percentile GSD

50th percentile =

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

be shown using Equation 4-1 to be 1.52.

If the assumption that is favorable to claimants 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 be less 

than 2 mrem/yr. 

For purposes of dose reconstruction, these small doses would be inconsequential to estimates of 

causation and should not be included in the IREP evaluations.

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 (Lantz 1978a). Six additional sampling stations were added in 1978 

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

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

populated areas on the site and 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 increased as a result of the requirements 

of DOE Order 5400.1 (DOE 1990). Known and potential effluent sources throughout NTS were 

assessed for their potential to contribute to public dose (Black and Townsend 1997).
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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. For cases in which maximum values were not 

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

to annual area concentrations, Table 4-2 lists 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 

occupationally exposed for 2,000 hours per year and had a breathing rate of 2,400 m3

/yr. Table 4-3 

lists these calculated intakes.

Some covered employees remained on the site continuously for weeks at a time. However, because 

most nonworking hours were spent indoors where ambient air particulate loadings would be much 

less than 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 

workweek 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 dose reconstruction, the 

inhalation intake values in Table 4-3 can be assumed to be bounding; therefore, the organ dose from 

these intakes should be applied in the IREP evaluation with a constant distribution.

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

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

evaluation for type Super S (ORAUT 2010) 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 

and Kordas 1983, 1985; McArthur and Mead 1987, 1988, 1989; McArthur 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 areal size. The results in Table 4-5 are representative of 

areas of NTS that contain measurable levels of contamination. These areas actually represent only 

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

In addition, because the data in Table 4-4 are representative of 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.

4.2.1.2.4 Scaling Factors for Inhalation Intakes

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

in NTS soils, scaling factors were developed to estimate potential intakes for these radionuclides 

based on their relative abundances in comparison with 239Pu when the soil contamination data 

(McArthur 1991) have been decay-corrected to 1963. These area-specific ratios are listed in 

Table 4-6.

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

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

have been identified as persistent in NTS soils. For dose reconstruction, maximum intakes of these 

radionuclides are calculated by selecting the maximum annual intake of plutonium (e.g., 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.
 

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Table 4-2. Atmospheric concentrations of 239Pu for sampled areas (Bq/m3).a,b

Year Area 1 Area 2 Area 3 Area 4 Area 5 Area 6 Area 7 Area 9 Area 10 Area 11 Area 12

1971 1.37E-05 2.45E-05 6.11E-06 ND 4.37E-06 6.36E-06 ND 2.67E-05 7.70E-06 3.37E-06 6.03E-06

1972 5.48E-06 8.29E-06 1.36E-05 ND 5.18E-06 4.96E-05 ND 1.59E-04 1.07E-05 8.99E-06 2.91E-05

1973 3.22E-06 8.25E-06 7.62E-06 ND 3.23E-06 4.44E-06 ND 3.18E-05 1.63E-06 8.99E-06 1.59E-06

1974 2.96E-06 2.52E-06 4.66E-06 ND 1.85E-06 2.93E-06 ND 7.81E-06 2.11E-06 2.96E-06 2.15E-06

1975 1.07E-06 1.41E-06 5.40E-06 ND 1.59E-06 2.32E-06 ND 6.22E-06 2.11E-06 1.81E-06 1.22E-06

1976 1.06E-05 6.36E-06 3.85E-05 ND 5.55E-06 1.64E-05 ND 1.18E-04 7.88E-06 8.73E-06 1.03E-05

1977 9.99E-07 2.41E-06 9.44E-06 ND 1.24E-06 2.05E-06 ND 9.36E-06 1.70E-06 4.48E-06 7.77E-07

1978 1.74E-06 4.92E-06 1.04E-05 ND 2.26E-06 3.66E-06 ND 2.01E-05 1.07E-05 3.96E-06 2.29E-06

1979 2.74E-06 1.72E-05 6.11E-08 ND 7.59E-07 1.46E-06 1.81E-06 1.94E-05 1.07E-05 1.37E-06 8.88E-07

1980 8.51E-07 9.58E-06 4.14E-08 ND 1.08E-06 1.78E-06 1.67E-06 1.67E-05 ND 1.63E-06 7.03E-07

1981 9.62E-07 2.35E-06 5.18E-06 ND 9.18E-07 1.07E-06 1.22E-06 1.19E-05 ND 1.70E-06 7.40E-07

1982 2.26E-06 1.48E-06 2.31E-06 ND 7.36E-07 1.30E-06 2.74E-06 7.96E-06 ND 1.96E-06 9.99E-07

1983 1.15E-06 2.39E-06 3.42E-06 ND 8.40E-07 1.20E-06 1.22E-06 7.92E-06 ND 6.73E-06 8.51E-07

1984 7.03E-07 3.68E-06 8.44E-06 ND 1.40E-06 7.40E-07 1.41E-06 3.77E-05 ND 2.07E-06 6.29E-07

1985 5.55E-07 2.13E-06 7.88E-06 ND 1.13E-06 9.99E-07 9.99E-07 4.92E-05 ND 1.92E-06 8.51E-07

1986 5.48E-06 2.29E-06 2.02E-05 ND 1.02E-06 6.88E-06 1.78E-06 1.04E-05 ND 9.25E-07 1.18E-06

1987 2.02E-06 8.51E-07 1.04E-04 ND 6.11E-07 9.36E-07 5.92E-07 4.07E-06 ND 1.18E-06 5.92E-07

1988 3.40E-06 1.04E-06 7.92E-06 ND 1.02E-06 8.14E-07 5.55E-07 1.89E-06 ND 4.44E-07 5.92E-07

1989 3.16E-05 2.15E-06 1.99E-05 ND 1.86E-06 2.96E-06 9.99E-06 1.30E-05 ND 2.00E-05 2.52E-05

1990 3.66E-06 1.20E-06 1.28E-05 ND 1.14E-06 1.37E-06 1.50E-06 1.84E-05 1.57E-06 3.96E-06 3.85E-07

1991 1.15E-04 7.59E-07 1.42E-05 ND 1.30E-06 2.68E-06 1.20E-06 1.30E-05 1.99E-08 8.29E-06 5.29E-07

1992 5.55E-06 4.07E-06 8.36E-05 ND 2.97E-06 2.69E-06 1.44E-05 3.26E-05 8.88E-06 3.26E-06 1.41E-06

1993 7.51E-06 1.81E-06 1.47E-05 ND 2.96E-06 3.31E-06 2.89E-06 9.99E-05 6.29E-06 1.55E-05 3.11E-05

1994 ND ND ND ND ND ND ND ND ND ND ND

1995 1.50E-06 4.07E-07 8.99E-06 ND 5.66E-07 1.86E-06 4.44E-07 1.44E-05 1.39E-06 1.48E-06 4.07E-07

1996 2.78E-05 ND 1.12E-05 ND 6.07E-07 1.70E-06 1.67E-05 2.26E-05 2.13E-06 7.03E-07 1.37E-07

1997 ND 3.96E-07 ND ND ND ND ND ND 9.73E-07 4.07E-07 5.55E-08

1998 1.74E-05 ND 2.49E-06 ND 3.52E-06 4.07E-07 1.01E-06 2.72E-05 ND 3.50E-07 ND

1999 ND ND ND 2.18E-06 ND ND ND ND ND ND ND

2000 3.14E-06 ND 9.95E-06 2.18E-06 7.59E-07 1.52E-05 1.26E-06 1.05E-04 ND ND ND

2001 1.68E-05 4.62E-07 4.29E-06 6.92E-06 2.53E-06 3.57E-06 6.28E-07 1.86E-05 3.13E-07 ND ND
 

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Year Area 15 Area 16 Area 18 Area 19 Area 20 Area 23 Area 25 Area 27 Area 28

Site 

average

Site 

maximum

1971 ND 7.14E-06 2.96E-06 8.36E-06 ND 3.46E-06 ND 3.11E-06 1.44E-06 8.35E-06 2.67E-05

1972 ND 6.25E-06 5.37E-06 1.01E-04 ND 1.29E-05 ND 6.14E-06 1.89E-06 2.82E-05 1.59E-04

1973 ND 1.63E-06 1.41E-06 2.65E-06 ND 2.07E-06 ND 1.37E-06 1.22E-06 5.41E-06 3.18E-05

1974 ND 2.22E-06 ND 2.44E-06 ND 3.15E-06 3.18E-06 2.66E-06 4.33E-06 3.20E-06 7.81E-06

1975 ND 1.59E-06 ND 1.24E-06 ND 1.63E-06 2.11E-06 2.52E-06 2.59E-06 2.32E-06 6.22E-06

1976 ND 5.77E-06 ND 3.66E-06 ND 2.10E-05 7.44E-06 5.07E-06 2.33E-06 1.78E-05 1.18E-04

1977 ND 9.62E-07 ND 8.51E-07 ND 9.99E-07 1.22E-06 1.15E-06 8.51E-07 2.56E-06 9.44E-06

1978 ND 1.89E-06 ND 2.28E-06 ND 2.15E-06 2.58E-06 3.48E-06 1.44E-06 4.92E-06 2.01E-05

1979 1.92E-06 7.77E-07 ND 7.22E-07 ND 8.25E-07 6.66E-07 5.92E-07 8.51E-07 3.69E-06 1.94E-05

1980 2.36E-06 ND ND 3.53E-06 ND 1.31E-06 1.13E-06 7.77E-07 5.92E-07 2.73E-06 1.67E-05

1981 3.44E-06 9.62E-07 ND 8.51E-07 ND 1.06E-06 8.99E-07 9.25E-07 5.92E-07 2.17E-06 1.19E-05

1982 1.55E-06 6.29E-07 ND 8.03E-07 ND 7.77E-07 6.29E-07 ND ND 1.87E-06 7.96E-06

1983 8.25E-07 5.55E-07 ND 1.09E-06 ND 1.04E-06 1.11E-06 ND ND 2.17E-06 7.92E-06

1984 1.18E-06 1.04E-06 ND 9.25E-07 ND 1.06E-06 1.18E-06 1.78E-06 ND 4.26E-06 3.77E-05

1985 1.48E-06 7.77E-07 ND 1.30E-06 ND 8.14E-07 9.73E-07 1.15E-06 ND 4.81E-06 4.92E-05

1986 1.44E-06 1.04E-06 ND 8.14E-07 ND 2.42E-06 7.59E-07 8.51E-07 ND 3.83E-06 2.02E-05

1987 6.66E-07 5.18E-07 ND 6.29E-07 ND 5.92E-07 8.51E-07 ND ND 8.43E-06 1.04E-04

1988 8.99E-07 ND ND 5.00E-07 ND 1.30E-06 5.55E-07 5.55E-07 ND 1.53E-06 7.92E-06

1989 8.62E-07 4.07E-07 ND 3.33E-07 ND 4.33E-06 3.50E-07 1.78E-07 ND 8.87E-06 3.16E-05

1990 3.03E-06 2.16E-07 ND 1.45E-07 ND 3.13E-06 4.18E-07 1.30E-06 ND 3.39E-06 1.84E-05

1991 4.55E-06 1.07E-06 ND 3.26E-07 ND 4.92E-07 9.51E-07 2.86E-07 ND 1.03E-05 1.15E-04

1992 4.07E-05 7.03E-07 ND 3.29E-07 ND 4.70E-07 6.70E-07 5.92E-06 ND 1.30E-05 8.36E-05

1993 1.22E-05 4.81E-07 ND 2.85E-07 ND 2.31E-07 1.26E-07 4.07E-07 ND 1.25E-05 9.99E-05

1994 ND ND ND ND ND ND ND ND ND ND ND

1995 5.92E-06 5.18E-07 ND 3.70E-08 ND 2.75E-07 1.67E-07 ND ND 2.56E-06 1.44E-05

1996 8.88E-06 6.66E-08 1.30E-07 ND 1.67E-07 2.31E-07 1.55E-07 9.62E-08 ND 5.82E-06 2.78E-05

1997 1.07E-06 7.03E-09 7.40E-07 ND 7.40E-07 2.22E-08 3.74E-08 5.92E-08 ND 4.11E-07 1.07E-06

1998 2.35E-06 ND 1.14E-06 ND 1.21E-06 4.29E-07 2.32E-07 ND ND 4.81E-06 2.72E-05

1999 4.07E-07 ND 4.07E-07 ND 7.40E-07 ND 2.29E-07 ND ND 7.93E-07 2.18E-06

2000 9.56E-06 ND 3.78E-07 ND 2.43E-07 ND 4.25E-07 ND ND 1.34E-05 1.05E-04

2001 5.18E-07 3.53E-07 6.37E-07 ND 2.80E-07 7.51E-08 1.63E-07 ND ND 3.75E-06 1.86E-05

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

b. ND = no data. 
 

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

Year Area 1 Area 2 Area 3 Area 4 Area 5 Area 6 Area 7 Area 9 Area 10 Area 11 Area 12

1971 0.0329 0.0587 0.0147 ND 0.0105 0.0153 ND 0.0640 0.0185 0.0081 0.0145

1972 0.0131 0.0199 0.0327 ND 0.0124 0.1190 ND 0.3810 0.0258 0.0216 0.0698

1973 0.0077 0.0198 0.0183 ND 0.0078 0.0107 ND 0.0763 0.0039 0.0216 0.0038

1974 0.0071 0.0060 0.0112 ND 0.0044 0.0070 ND 0.0187 0.0051 0.0071 0.0052

1975 0.0026 0.0034 0.0130 ND 0.0038 0.0056 ND 0.0149 0.0051 0.0044 0.0029

1976 0.0255 0.0153 0.0924 ND 0.0133 0.0392 ND 0.2824 0.0189 0.0210 0.0248

1977 0.0024 0.0058 0.0226 ND 0.0030 0.0049 ND 0.0225 0.0041 0.0107 0.0019

1978 0.0042 0.0118 0.0250 ND 0.0054 0.0088 ND 0.0481 0.0258 0.0095 0.0055

1979 0.0066 0.0413 0.0001 ND 0.0018 0.0035 0.0044 0.0465 0.0258 0.0033 0.0021

1980 0.0020 0.0230 0.0001 ND 0.0026 0.0043 0.0040 0.0401 ND 0.0039 0.0017

1981 0.0023 0.0056 0.0124 ND 0.0022 0.0026 0.0029 0.0286 ND 0.0041 0.0018

1982 0.0054 0.0036 0.0055 ND 0.0018 0.0031 0.0066 0.0191 ND 0.0047 0.0024

1983 0.0028 0.0057 0.0082 ND 0.0020 0.0029 0.0029 0.0190 ND 0.0162 0.0020

1984 0.0017 0.0088 0.0202 ND 0.0034 0.0018 0.0034 0.0906 ND 0.0050 0.0015

1985 0.0013 0.0051 0.0189 ND 0.0027 0.0024 0.0024 0.1181 ND 0.0046 0.0020

1986 0.0131 0.0055 0.0484 ND 0.0025 0.0165 0.0043 0.0249 ND 0.0022 0.0028

1987 0.0048 0.0020 0.2495 ND 0.0015 0.0022 0.0014 0.0098 ND 0.0028 0.0014

1988 0.0082 0.0025 0.0190 ND 0.0025 0.0020 0.0013 0.0045 ND 0.0011 0.0014

1989 0.0759 0.0052 0.0478 ND 0.0045 0.0071 0.0240 0.0311 ND 0.0480 0.0604

1990 0.0088 0.0029 0.0308 ND 0.0027 0.0033 0.0036 0.0442 0.0038 0.0095 0.0009

1991 0.2753 0.0018 0.0340 ND 0.0031 0.0064 0.0029 0.0311 0.0000 0.0199 0.0013

1992 0.0133 0.0098 0.2007 ND 0.0071 0.0065 0.0346 0.0781 0.0213 0.0078 0.0034

1993 0.0180 0.0044 0.0352 ND 0.0071 0.0079 0.0069 0.2398 0.0151 0.0373 0.0746

1994 ND ND ND ND ND ND ND ND ND ND ND

1995 0.0036 0.0010 0.0216 ND 0.0014 0.0045 0.0011 0.0346 0.0033 0.0036 0.0010

1996 0.0666 0.0000 0.0268 ND 0.0015 0.0041 0.0400 0.0542 0.0051 0.0017 0.0003

1997 ND 0.0010 ND ND ND ND ND ND 0.0023 0.0010 0.0001

1998 0.0416 ND 0.0060 ND 0.0084 0.0010 0.0024 0.0653 ND 0.0008 ND

1999 ND ND ND ND ND ND ND ND ND ND ND

2000 0.0075 ND 0.0239 0.0052 0.0018 0.0366 0.0030 0.2509 ND ND ND

2001 0.0404 0.0011 0.0103 0.0166 0.0061 0.0086 0.0015 0.0447 0.0008 ND ND
 

Document No. ORAUT-TKBS-0008-4 Revision No. 04 Effective Date: 02/19/2020 Page 23 of 94

Year Area 15 Area 16 Area 18 Area 19 Area 20 Area 23 Area 25 Area 27 Area 28

Site 

average

Site 

maximum

1971 ND 0.0171 0.0071 0.0201 ND 0.0083 ND 0.0075 0.0035 0.0200 0.0640

1972 ND 0.0150 0.0129 0.2424 ND 0.0309 ND 0.0147 0.0045 0.0677 0.3810

1973 ND 0.0039 0.0034 0.0063 ND 0.0050 ND 0.0033 0.0029 0.0130 0.0763

1974 ND 0.0053 ND0.0059 ND 0.0075 0.0076 0.0064 0.0104 0.0077 0.0187

1975 ND 0.0038 ND 0.0030 ND 0.0039 0.0051 0.0060 0.0062 0.0056 0.0149

1976 ND 0.0139 ND 0.0088 ND 0.0504 0.0178 0.0122 0.0056 0.0428 0.2824

1977 ND 0.0023 ND 0.0020 ND 0.0024 0.0029 0.0028 0.0020 0.0062 0.0226

1978 ND 0.0045 ND 0.0055 ND 0.0052 0.0062 0.0083 0.0035 0.0118 0.0481

1979 0.0046 0.0019 ND 0.0017 ND 0.0020 0.0016 0.0014 0.0020 0.0089 0.0465

1980 0.0057 ND ND 0.0085 ND 0.0031 0.0027 0.0019 0.0014 0.0070 0.0401

1981 0.0083 0.0023 ND 0.0020ND 0.0025 0.0022 0.0022 0.0014 0.0052 0.0286

1982 0.0037 0.0015 ND 0.0019 ND 0.0019 0.0015 ND ND 0.0045 0.0191

1983 0.0020 0.0013 ND 0.0026 ND 0.0025 0.0027 ND ND 0.0052 0.0190

1984 0.0028 0.0025 ND 0.0022 ND 0.0025 0.0028 0.0043 ND 0.0102 0.0906

19850.0036 0.0019 ND 0.0031 ND 0.0020 0.0023 0.0028 ND 0.0115 0.1181

1986 0.0035 0.0025 ND 0.0020 ND 0.0058 0.0018 0.0020 ND 0.0092 0.0484

1987 0.0016 0.0012 ND 0.0015 ND 0.0014 0.0020 0.0000 ND 0.0189 0.2495

1988 0.0022 ND ND 0.0012 ND 0.0031 0.0013 0.0013 ND 0.0037 0.0190

1989 0.0021 0.0010 ND 0.0008 ND 0.0104 0.0008 0.0004 ND 0.0213 0.0759

1990 0.0073 0.0005 ND 0.0003 ND 0.0075 0.0010 0.0031 ND 0.0081 0.0442

1991 0.0109 0.0026 ND 0.0008 ND 0.0012 0.0023 0.0007 ND 0.0246 0.2753

1992 0.0977 0.0017 ND 0.0008 ND 0.0011 0.0016 0.0142 ND 0.0312 0.2007

1993 0.0293 0.0012 ND 0.0007 ND 0.0006 0.0003 0.0010 ND 0.0300 0.2398

1994 ND ND ND ND ND ND ND ND ND ND ND

1995 0.0142 0.0012 ND 0.0001 ND 0.0007 0.0004 ND ND 0.0061 0.0346

1996 0.0213 0.0002 0.0003 0.0000 0.0004 0.0006 0.0004 0.0002 ND 0.0124 0.0666

1997 0.0026 0.0000 0.0018 ND 0.0018 0.0001 0.0001 0.0001 ND 0.0010 0.0026

1998 0.0056 ND 0.0027 ND 0.0029 0.0010 0.0006 ND ND 0.0115 0.0653

1999 0.0010 ND 0.0010 ND 0.0018 ND 0.0006 ND ND 0.0011 0.0018

2000 0.0229 ND 0.0009 ND 0.0006 ND 0.0010 ND ND 0.0322 0.2509

2001 0.0012 0.0008 0.0015 ND 0.0007 0.0002 0.0004 ND ND 0.0090 0.0447

a. ND = no data.
 

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Table 4-4. Inventory of contaminated soil (Bq).a,b

Area Area (m2) Am-241 Pu-238 Pu-239,240 Co-60 Cs-137 Sr-90 Eu-152 Eu-154 Eu-155

1 6.86E+07 1.554E+11 2.41E+11 8.88E+11 4.07E+10 3.26E+11 5.55E+11 5.55E+11 3.7E+09 1.85E+10

2 5.10E+07 1.073E+11 3.18E+11 8.14E+11 4.44E+10 8.88E+11 1.7E+12 5.18E+11 ND 1.48E+10

3 8.37E+07 1.702E+11 1.15E+11 1.37E+12 3.7E+10 4.44E+11 1.22E+12 6.66E+11 3.7E+09 1.85E+10

4 4.14E+07 2.442E+11 4.81E+11 1.48E+12 5.92E+10 4.44E+11 4.81E+11 3.37E+11 ND 7.4E+09

5 7.51E+06 2.22E+10 3.7E+09 1.78E+11 2.22E+10 1.48E+10 3.33E+10 3.7E+11 7.4E+09 ND

6 8.37E+07 6.29E+10 1.22E+11 3.11E+11 7.4E+09 1.04E+11 1.3E+11 ND ND ND

7 5.00E+07 8.14E+10 2.22E+10 5.92E+11 3.7E+10 1.92E+11 3.4E+11 8.14E+11 7.4E+09 1.11E+10

8 3.60E+07 6.29E+11 2.96E+11 4.07E+12 2.11E+11 1.55E+12 9.25E+11 1.63E+11 ND 2.22E+10

9 5.18E+07 1.554E+11 8.14E+10 3.29E+12 2.59E+10 3.22E+11 4.81E+11 8.51E+11 7.4E+09 1.11E+10

10 5.18E+07 7.03E+11 7.03E+11 4.07E+12 3.59E+11 3.11E+12 2.04E+12 8.14E+10 1.11E+10 1.85E+11

11 1.04E+07 1.221E+11 1.85E+10 1.07E+12 ND 1.85E+10 1.11E+10 ND ND ND

12 1.03E+08 2.109E+11 3.15E+11 1.44E+12 4.44E+10 7.4E+11 6.29E+11 ND ND ND

15 9.14E+07 2.96E+11 2.89E+11 2.33E+12 1.11E+10 7.03E+11 8.14E+11 ND ND ND

16 3.70E+07 2.59E+10 5.55E+10 1.37E+11 3.7E+09 1.07E+11 1.37E+11 ND ND ND

17 8.13E+07 1.036E+11 1.67E+11 6.66E+11 3.7E+10 5.55E+11 7.03E+11 ND ND ND

18 7.07E+07 7.03E+11 2.07E+11 3.7E+12 2.59E+10 3.7E+11 6.29E+11 4.07E+10 3.7E+09 2.96E+10

19 3.84E+08 7.77E+11 1.18E+12 5.18E+12 4.07E+10 1.33E+12 1.15E+12 ND ND ND

20 1.61E+07 8.51E+11 1.11E+12 1.52E+12 2.92E+11 2.04E+11 1.59E+11 4.81E+11 5.92E+10 1.78E+11

25 2.33E+06 ND ND ND ND 7.4E+09 3.7E+09 1.48E+10 ND ND

26 5.18E+05 ND ND ND ND ND ND ND ND ND

30 7.77E+04 1.184E+11 1.67E+11 5.18E+11 2.96E+10 5.55E+10 4.81E+10 2.59E+10 3.7E+09 7.4E+09

a. Source: McArthur (1991).

b. ND = no data.
 

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

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 ND 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 ND 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 ND

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

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 ND 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 ND 3.41E+03 2.09E+03 ND ND ND

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

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

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

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

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 ND ND ND

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 ND ND ND ND 6.06E+03 3.09E+03 2.72E+04 ND ND

26 ND ND ND ND ND ND ND ND ND

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. ND = no data.
 

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Table 4-6. Abundance of radionuclides in NTS soils in relation 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 NDa 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 ND 0.250

5 0.131 0.026 1.000 4.958 0.159 0.365 8.925 0.378 ND

6 0.212 0.490 1.000 0.944 0.636 0.810 ND ND ND

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 ND 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 ND 0.033 0.020 ND ND ND

12 0.153 0.272 1.000 1.220 0.978 0.848 ND ND ND

15 0.133 0.154 1.000 0.189 0.575 0.679 ND ND ND

16 0.198 0.505 1.000 1.072 1.495 1.945 ND ND ND

17 0.163 0.312 1.000 2.204 1.590 2.053 ND ND ND

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 ND ND ND

20 0.586 0.912 1.000 7.643 0.256 0.204 1.358 0.354 5.849

25 ND ND ND ND ND ND ND ND ND

26 ND ND ND ND ND ND ND ND ND

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

a. ND = no data.
Document No. ORAUT-TKBS-0008-4 Revision No. 04 Effective Date: 02/19/2020 Page 27 of 94

 

4.2.1.2.5 Correction for Resuspension for Early Times after Atmospheric Tests

Anspaugh et al. (2002) stated that, based on empirical observations, concentrations of resuspended 

radionuclides in air have been noted to display a strong time dependence early after deposition and 

that this pathway could be important for reoccupation of contaminated property. Anspaugh et al. also 

stated that there had not been universal agreement that resuspension is an important pathway but 

that it is now generally accepted that there are a few instances in which 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 et al. stated 

that the resuspension factor model has been widely used to predict the concentration of resuspended 

radionuclides early 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 et al. also stated 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 et al. (2002) presented several resuspension models that have been proposed but 

concluded that they can be over- or under-predictive at various times after deposition in comparison 

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

others in the 1980s, Anspaugh et al. proposed a resuspension model that more accurately describes 

the observed results over the entire timespan of the expanded data set for NTS:

(4-2) 5 0.07 9 0.003 9 1 1 (1 10 )e (6 10 )e (1 10 ) (1 10 ) m t t Sf

− − − − − ± − = × + × +× ××    

A graphical depiction of Equation 4-2 is in Figure 4-2 from time t equal zero to 1,000 days 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 times long after initial deposition and 

that the resuspension proposed by Anspaugh et al. is predictive of the observed results over the 

expanded dataset (including those developed in the 1980s), 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 air monitoring data are available (i.e., 1971 through 2001, see 

Section 4.2.1.2.2).

When Anspaugh et al.’s proposed resuspension model (Equation 4-2) is integrated over 180 to 

545 days and compared with 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 NTS.

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

were in July 1962 and, therefore, about 180 days had passed before the beginning of 1963.

For the Anspaugh et al. (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, 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.
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Figure 4-2. Resuspension factor as a function of time after initial deposition 

(from Attachment A). 

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

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 listed in Table 4-8.

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., 

4.84 Bq for 1963, 2.65 Bq for 1964, 1.92 for 1965, and 1.55 for all subsequent years 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 product doses greater than 0.001 rem 

are listed in Table 4-9. Other organs are not listed in Table 4-9 because all were less than 0.001 rem. 

These doses should be added to the input for the 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 Annual Intakes from Ingestion

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 the radionuclides are deposited within the top 2.3 cm of soil and a soil density of 1.5 g/cm3

(DOE 2003). The area-specific radionuclide soil concentrations are listed in Table 4-10.
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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

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

Organ(s) 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

RBM 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). 

Organ 1963 1964 1965 1966 1967 1968 1969 1970 1971 1972

Liver 0.002 <0.001 <0.001 <0.001 <0.001 <0.001 <0.001 <0.001 <0.001 <0.001

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 <0.001 <0.001 <0.001 <0.001 <0.001 <0.001 <0.001 <0.001 <0.001

ET 0.002 <0.001 <0.001 <0.001 <0.001 <0.001 <0.001 <0.001 <0.001 <0.001

Lung 0.008 0.003 0.002 0.001 <0.001 <0.001 <0.001 <0.001 <0.001 <0.001

ET1 1.635 0.474 0.248 0.164 0.146 0.138 0.138 0.138 0.138 0.138

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 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 

Table 4-11.

For most radionuclides, Area 30 provided the highest areal deposition and resultant intakes. Area 30 

is relatively small (150 km2

) and inaccessible, and is 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 4-11 have been provided without the Area 30 areal concentrations.
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Table 4-10. Radionuclide soil concentration by area (Bq/g).

a

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 ND 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 ND 0.259

5 0.090 0.018 0.686 3.401 0.109 0.250 6.122 0.259 ND

6 0.023 0.053 0.108 0.102 0.069 0.087 ND ND ND

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 ND 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 ND 0.099 0.060 ND ND ND

12 0.062 0.111 0.408 0.498 0.399 0.346 ND ND ND

15 0.098 0.114 0.740 0.140 0.426 0.502 ND ND ND

16 0.021 0.054 0.107 0.115 0.160 0.209 ND ND ND

17 0.039 0.074 0.238 0.524 0.378 0.488 ND ND ND

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 ND ND ND

20 1.607 2.500 2.741 20.946 0.701 0.559 3.723 0.969 16.031

25 ND ND ND ND 0.176 0.090 0.789 ND ND

26 ND ND ND ND ND ND ND ND ND

30 4.620 7.749 19.340 43.835 3.953 3.493 4.143 1.252 13.804

a. ND = no data.

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

a

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 ND 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 ND 6.47

5 2.24 0.45 17.15 85.03 2.73 6.25 153.05 6.48 ND

6 0.57 1.32 2.69 2.54 1.71 2.18 ND ND ND

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 ND 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 ND 2.47 1.51 ND ND ND

12 1.56 2.77 10.20 12.45 9.98 8.65 ND ND ND

15 2.45 2.85 18.49 3.49 10.64 12.56 ND ND ND

16 0.53 1.35 2.68 2.87 4.01 5.21 ND ND ND

17 0.97 1.85 5.94 13.09 9.44 12.19 ND ND ND

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 ND ND ND

20 40.17 62.49 68.51 523.64 17.54 13.98 93.06 24.23 400.77

25 ND ND ND ND 4.39 2.24 19.73 ND ND

26 ND ND ND ND ND ND ND ND ND

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

Maximumb 40.17 62.49 81.99 523.64 83.02 55.42 153.05 24.23 400.77

a. ND = no data.

b. Maximum value with Area 30 excluded.

As with inhalation intakes, a method was developed to adjust ingestion doses 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 developed based 
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on the relative contribution of 90Sr to the total ingestion dose. These correction factors are listed in 

Table 4-12.

Table 4-12. Organ-specific ingestion fission and activation product correction factors (from 

Attachment A). 

Organ 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 organ-specific annual doses for the 90Sr 

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

Table 4-13 are 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 organspecific doses that were greater than or equal to 0.001 rem are listed 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 bone 

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
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Table 4-14. Organ-specific doses (rem) from ingestion of fission and activation products. Organ 1963 1964 1965 1966 1967 1968 1969 1970 1971 1972

Bladder 0.002 0.002 0.001 <0.001 <0.001 <0.001 <0.001 <0.001 <0.001 <0.001

Bone surface <0.001 0.001 0.002 0.002 0.002 0.003 0.003 0.003 0.003 0.003

Stomach 0.001 <0.001 <0.001 <0.001 <0.001 <0.001 <0.001 <0.001 <0.001 <0.001

SI 0.002 0.001 <0.001 <0.001 <0.001 <0.001 <0.001 <0.001 <0.001 <0.001

ULI 0.015 0.006 0.003 0.002 0.001 <0.001 <0.001 <0.001 <0.001 <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 <0.001 <0.001

4.3 OCCUPATIONAL EXTERNAL DOSE

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

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

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

4.3.1 Ambient Radiation

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 between 1951 

and 1967, most of these were to characterize the effects of weapons tests and, therefore, were not 

appropriate for use in estimating external environmental dose for unmonitored employees.

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 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 at the sample locations for measurements and another stored fully charged 

in the laboratory. Each week these sets were exchanged, the fresh set was recharged and a record 

was kept of the amount of drift that occurred while one was stored in the laboratory. A specially 

designed shockproof 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 to be the result of shock or other 

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

subtraction 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; ambient 

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

1978a). The dosimeters 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 

(1,030 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.
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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 more than 150 and eventually 

covered all populated areas at NTS. For dose reconstruction, because the values in Table 4-15 

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

exposure (i.e., for a maximizing and best estimate value, assume 2,600 hr/yr and 2,500 hr/yr, 

respectively) and added to the dose of record for unmonitored employees after 1992 (see below). If 

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

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

Some covered employees remained on site continuously for weeks at a time. However, because 

most of the nonworking 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 4-15, adjustment of the tabular data is not required to ensure ambient exposures are 

not underestimated for these individuals.

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 2012). However, for earlier measurement methods, the overall uncertainty could have 

been greater. Therefore, for dose reconstructions that are favorable to claimants, the assumption is 

made that the 95th-percentile values of the expected values (50th percentile) in Table 4-15 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. Site maximum values would be 

entered into IREP as constants.

Assignment of dosimeters to all persons at NTS began in April 1957 and ended in December 1992 

(ORAUT 2012). During this time, readings from control badges at Mercury were used to subtract 

ambient background dose from the personnel dosimeters (ORAUT 2012). For this reason, from April 

1957 to December 1992, ambient background external radiation dose should not be included in dose 

reconstructions. Because no documentation is available before 1967 (see Table 4-15), ambient 

external doses should not be assigned under any circumstances from 1951 through 1956. After 1992, 

ambient background external dose should be assigned to all unmonitored workers in accordance with 

the guidelines for Table 4-15. For years after 2010, assume 2010 exposure values.

4.3.2 Releases of Noble Gases

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 

created by a nuclear test. This process, sometimes referred to as “atmospheric pumping,” creates a 

diffuse source of radiological effluents.

In 1982, REECo assumed responsibility for six noble gas sampling stations EPA previously ran and 

replaced them with new samplers. These sampling units were housed in a metal toolbox with three 

metal air bottles attached with quick disconnect hoses. A vacuum was maintained on the first bottle, 

which caused a steady flow of air to be collected in the other two bottles. The flow rate was 
 

Document No. ORAUT-TKBS-0008-4 Revision No. 04 Effective Date: 02/19/2020 Page 34 of 94

Table 4-15. Ambient radiation by area (rem/yr).

a,b

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

1967 ND ND 0.318 ND ND 0.194 ND ND ND ND ND 0.205 ND ND

1968 ND ND 0.285 ND 0.183 0.175 ND ND ND ND ND 0.190 ND ND

1969 ND ND ND ND ND ND ND ND ND ND ND ND ND ND

1970 ND ND ND ND ND ND ND ND ND ND ND ND ND ND

1971 ND ND ND ND ND ND ND ND ND ND ND ND ND ND

1972 ND ND ND ND ND ND ND ND ND ND ND ND ND ND

1973 ND ND ND ND ND ND ND ND ND ND ND ND ND ND

1974 ND ND ND ND ND ND ND ND ND ND ND ND ND ND

1975 ND ND ND ND ND ND ND ND ND ND ND ND ND ND

1976 ND ND ND ND ND ND ND ND ND ND ND ND ND ND

1977 ND ND ND ND ND ND ND ND ND ND NDND ND ND

1978 0.130 0.167 0.200 ND 0.138 0.110 0.110 0.120 0.150 0.320 0.140 0.150 ND ND

1979 0.119 0.147 0.190 0.140 0.133 0.104 0.125 0.115 0.140 0.168 0.155 0.191 ND ND

1980 0.123 0.156 0.199 0.155 0.133 0.106 0.130 0.135 0.145 0.192 0.165 0.196 ND ND

1981 0.120 0.155 0.218 0.150 0.166 0.113 ND 0.125 0.15 0.180 0.160 0.187 ND ND

1982 0.126 0.149 0.188 0.145 0.141 0.108 ND 0.125 0.14 0.209 0.140 0.189 ND ND

1983 0.115 0.141 0.181 0.143 0.144 0.103 ND 0.095 0.135 0.201 0.140 0.174 ND ND

1984 0.097 0.149 0.162 0.116 0.111 0.086 0.327 0.125 0.102 0.166 0.112 0.155 ND ND

1985 ND 0.141 0.078 0.101 ND 0.107 0.347 0.094 0.114 0.225 ND 0.180 ND ND

1986 0.093 0.126 0.150 0.092 0.092 0.077 0.318 0.094 0.096 0.12 0.107 0.106 ND ND

1987 0.146 0.189 0.184 0.164 0.191 0.126 ND 0.12 0.149 0.181 0.133 0.168 ND ND

1988 0.155 0.209 0.207 0.164 0.177 0.134 ND 0.15 0.179 0.203 0.158 0.194 ND ND

1989 0.140 0.217 0.205 0.132 0.141 0.106 ND 0.126 0.151 0.179 0.153 0.139 ND ND

1990 0.134 0.167 0.187 0.142 0.142 0.109ND 0.122 0.148 0.170 0.132 0.140 ND ND

1991 0.128 0.168 0.194 0.146 0.165 0.116 0.132 0.124 0.132 0.15 0.13 0.190 ND ND

1992 0.123 0.167 0.173 0.141 0.150 0.104 0.145 0.128 0.136 0.157 0.158 0.170 ND ND

1993 0.147 0.178 0.200 0.158 0.198 0.134 0.165 0.153 0.156 0.172 0.184 0.201 ND ND

1994 ND 0.062 0.062 ND 0.089 0.084 ND ND ND 0.070 0.102 0.097 ND ND

1995 0.102 0.136 0.150 0.121 0.212 0.087 0.176 0.102 0.104 0.179 0.124 0.130 ND ND

1996 0.102 0.131 0.151 0.110 0.225 0.105 0.164 0.098 0.099 0.176 0.124 0.135 ND ND

1997 ND ND 0.131 ND 0.086 0.088 ND ND 0.081 ND ND ND ND ND

1998 0.106 0.132 0.137 0.110 0.134 0.100 0.201 0.127 0.101 0.168 0.121 0.148 ND ND

1999 0.108 0.133 0.135 0.122 0.110 0.105 0.191 0.129 0.099 0.160 0.122 0.152 ND ND

2000 0.127 0.155 0.161 0.129 0.134 0.122 0.217 0.147 0.190 0.187 0.144 0.176 ND ND

2001 0.121 0.299 0.151 0.120 0.136 0.107 0.214 0.143 0.115 0.180 0.14 0.163 ND ND

2002 0.113 0.272 0.168 0.341 0.122 0.098 0.174 0.135 0.106 0.167 0.131 0.162 ND ND

2003 0.18 0.26 0.257 0.334 0.145 0.104 0.159 0.209 0.189 0.188 0.133 0.169 ND ND

2004 0.162 0.251 0.239 0.323 0.148 0.106 0.157 0.204 0.190 0.185 0.130 0.166 ND ND

2005 0.159 0.246 0.236 0.320 0.148 0.108 0.159 0.205 0.191 0.182 0.132 0.166 ND ND

2006 0.154 0.243 0.228 0.309 0.144 0.108 0.156 0.202 0.186 0.182 0.133 0.167 ND ND

2007 0.155 0.242 0.228 0.306 0.147 0.112 0.158 0.199 0.190 0.184 0.138 0.171 ND ND

2008 0.148 0.227 0.219 0.298 0.138 0.107 0.15 0.189 0.179 0.174 0.129 0.159 ND ND

2009 0.151 0.229 0.218 0.2830.140 0.109 0.152 0.191 0.180 0.175 0.134 0.164 ND ND

2010 0.146 0.216 0.250 0.269 0.144 0.098 0.149 0.184 0.173 0.170 0.132 0.156 ND ND
 

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4 Effective Date: 02/19/2020 Page 35 of 94

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

1967 ND ND ND ND ND ND ND ND 0.142 ND ND ND 0.175 ND ND ND 0.207 0.318

1968 ND ND ND 0.186 ND ND ND ND 0.131 ND ND ND 0.172 ND ND ND 0.189 0.285

1969 ND ND ND ND ND ND ND ND ND ND ND ND ND ND ND ND 0.155c 0.320c

1970 ND ND ND ND ND ND ND ND ND ND ND ND ND ND ND ND 0.155c 0.320c

1971 ND ND ND ND ND ND ND ND ND ND ND ND ND ND ND ND 0.155c 0.320c

1972 ND ND ND ND ND ND ND ND ND ND ND ND ND ND ND ND 0.155c 0.320c

1973 ND ND ND ND ND ND ND ND ND ND ND ND ND ND ND ND 0.155c 0.320c

1974 ND ND ND ND ND ND ND ND ND ND ND ND ND ND ND ND 0.155c 0.320c

1975 ND ND ND ND ND ND ND ND ND ND ND ND ND ND ND ND 0.155c 0.320c

1976 ND ND ND ND ND ND ND ND ND ND ND ND ND ND ND ND 0.155c 0.320c

1977 ND ND ND ND ND ND ND ND ND ND ND ND ND ND ND ND 0.155c 0.320c

1978 0.129 ND ND 0.173 0.185 ND ND ND 0.150 ND 0.162 ND 0.140 0.120 ND ND 0.155 0.320

1979 0.121 ND 0.150 0.145 0.155 0.160 ND 0.115 0.117 ND 0.152 ND 0.130 0.120 ND 0.125 0.140 0.191

1980 0.134 ND 0.165 0.163 0.185 0.207 ND 0.070 0.098 ND 0.163 ND 0.135 0.130 ND 0.205 0.152 0.207

1981 0.138 ND 0.155 0.168 0.174 0.163 ND 0.075 0.126 ND 0.175 ND 0.150 ND ND 0.140 0.152 0.218

1982 0.134 ND 0.150 0.173 0.176 0.200 ND 0.075 0.095 ND 0.150 ND 0.135 0.135 ND 0.140 0.147 0.209

1983 0.124 ND 0.135 0.151 0.158 0.206 ND 0.125 0.163 ND 0.149 ND 0.140 ND ND 0.185 0.148 0.206

1984 0.108 ND 0.129 0.124 0.132 0.157 ND 0.058 0.154 ND 0.129 ND ND ND ND 0.173 0.137 0.327

1985 ND ND 0.115 0.123 0.131 0.127 ND ND ND ND 0.103 ND ND ND ND 0.153 0.143 0.347

1986 0.081 ND ND 0.124 0.134 0.160 ND 0.055 0.071 ND 0.106 ND 0.089 ND ND 0.135 0.116 0.318

1987 0.135 ND 0.151 0.174 0.246 0.204 ND 0.068 0.096 ND 0.129 ND 0.139 ND ND ND 0.155 0.246

1988 0.155 ND 0.172 0.212 0.223 0.211 ND 0.084 0.123 ND 0.163 ND ND ND ND ND 0.172 0.223

1989 0.162 ND 0.160 0.168 0.178 0.189 ND 0.078 0.116 ND 0.146 ND 0.118 ND ND ND 0.150 0.217

1990 0.124 ND 0.164 0.174 0.178 0.192 ND 0.083 0.177 ND 0.139 ND 0.146 ND ND ND 0.149 0.192

1991 0.122 ND 0.165 0.175 0.173 0.189 ND 0.081 0.079 ND 0.143 ND 0.154 0.110 ND 0.155 0.144 0.194

1992 0.134 ND 0.152 0.168 0.189 0.185 ND 0.079 0.075 ND 0.132 ND 0.143 ND ND 0.179 0.145 0.189

1993 0.145 ND 0.177 0.195 0.192 0.213 ND 0.091 0.096 ND 0.158 ND 0.169 ND ND 0.195 0.167 0.213

1994 0.143 ND ND ND 0.141 0.131 ND 0.058 0.061 ND 0.115 ND 0.121 ND ND 0.104 0.098 0.143

1995 0.109 ND 0.129 0.137 0.149 0.153 ND 0.058 0.070 ND 0.117 ND 0.121 ND ND ND 0.127 0.212

1996 0.103 ND 0.137 0.144 0.153 0.150 ND 0.055 0.056 ND 0.114 ND 0.124 ND ND 0.165 0.128 0.225

1997 0.102 ND ND ND 0.150 0.153 ND 0.083 0.063 ND 0.127 ND 0.135 ND ND 0.170 0.114 0.170

1998 0.099 ND ND 0.135 0.151 0.156 ND 0.076 0.058 ND 0.113 ND 0.123 ND ND 0.165 0.127 0.201

1999 0.108 ND ND 0.137 0.153 0.151 ND 0.075 0.057 ND 0.111 ND 0.126 ND ND 0.170 0.126 0.191

2000 0.117 ND ND 0.16 0.185 0.182 ND 0.092 0.081 ND 0.139 ND 0.158 ND ND 0.205 0.153 0.217

2001 0.112 ND ND 0.167 0.159 0.131 ND 0.090 0.074 ND 0.135 ND 0.145 ND ND 0.193 0.147 0.299

2002 0.103 ND ND 0.144 0.172 0.157 ND 0.081 0.068 ND 0.122 ND 0.128 ND ND ND 0.148 0.341

2003 0.116 0.133 ND 0.146 0.165 0.334 ND ND 0.065 ND 0.127 ND 0.113 ND ND ND 0.176 0.334

2004 0.111 0.137 ND 0.145 0.158 0.309 ND ND 0.093 ND 0.123 ND 0.112 ND ND ND 0.172 0.323

2005 0.114 0.144 ND 0.146 0.164 0.305 ND ND 0.116 ND 0.125 ND 0.115 ND ND ND 0.174 0.320

2006 0.117 0.143 ND 0.149 0.161 0.290 ND ND 0.076 ND 0.125 ND 0.115 ND ND ND 0.169 0.309

2007 0.116 0.145 ND 0.151 0.165 0.287 ND ND 0.098 ND 0.128 ND 0.117 ND ND ND 0.172 0.306

2008 0.111 0.137 ND 0.144 0.161 0.272 ND ND 0.068 ND 0.124 ND 0.114 ND ND ND 0.162 0.298

2009 0.115 0.133 ND 0.148 0.160 0.255 ND ND 0.063 ND 0.117 ND 0.116 ND ND ND 0.162 0.283

2010 0.111 0.140 ND 0.142 0.157 0.247 ND ND 0.068 ND 0.116 ND 0.114 ND ND ND 0.159 0.269

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

b. ND = no data.

c. Ambient radiation not provided for these years; assumed equal to value for 1978.
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Table 4-16. 85Kr average background and the Area 20 net concentrations and annual organ doses 

from 1983 through 1993.a

Year

NTS 

background 

(Bq/m3)

Area 20 camp 

(Bq/m3)

Net elevated 

(Bq/m3)

Sourceb

elevated 

concentration 

(Bq/m3)

Other organc

dosed at 

sourceb

(rem/yr)

Skin dosee at 

sourceb

(rem/yr)

1983 0.93 0.85 0.00 0.00 0.00E+00 0.00E+00

1984 1.04 1.15 0.11 3.00 4.67E-07 3.00E-05

1985 1.22 1.70 0.48 12.99 2.02E-06 1.30E-04

1986 1.30 2.15 0.85 22.98 3.58E-06 2.30E-04

1987 1.04 1.44 0.41 10.99 1.71E-06 1.10E-04

1988 0.93 1.07 0.15 4.00 6.23E-07 4.00E-05

1989 0.85 1.00 0.15 4.00 6.23E-07 4.00E-05

1990 1.18 1.37 0.19 5.00 7.79E-07 5.00E-05

1991 0.93 1.18 0.26 6.99 1.09E-06 6.99E-05

1992 0.96 1.11 0.15 4.00 6.23E-07 4.00E-05

1993 1.00 1.04 0.04 1.00 1.56E-07 9.99E-06

a. Sources: Environmental reports in the References section.

b. Source receptor location assumed to be 500 m from U-20a emplacement hole.

c. Includes gonads, breast, lung, red bone marrow, bone surfaces, thyroid, remainder, and effective whole body.

d. Assumes 2,000 hr/yr exposure and DCF of 2.2 × 10-16 Sv/Bq/s/m3 (Eckerman and Ryman 1993).

e. Assumes 2,000 hr/yr exposure and DCF of 1.32 × 10-14 Sv/Bq/s/m3 (Eckerman and Ryman 1993).

approximately 0.5 cm3/min. The two collection bottles were exchanged weekly and yielded a sample 

volume of about 3 × 105

 cm3 (Scoggins 1983).

The noble gases were separated and collected from the atmospheric sample by a series of cryogenic 

gas chromatographic techniques. Water and carbon dioxide were removed at room temperature and 

krypton and xenon were collected on charcoal at liquid nitrogen temperatures. These gases were 

transferred to a molecular sieve where they were separated from remaining gases and each other. 

The krypton and xenon were 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 samplers showed no 

statistical difference in atmospheric concentrations of 85Kr during 1983 and 1984. The average 

concentration of all stations for these years was 0.93 Bq/m3 and 1.04 Bq/m3

, respectively, which was 

established as NTS background. This background measurement shows good agreement with the 

global background of 1.0 Bq/m3 (WMO 2002).

In 1984, REECo measured a statistically significant elevation in 85Kr concentrations in Area 20 of 0.11 

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 4-16 lists the average background and the Area 20 elevated concentrations 

from 1983 through 1993. 

The source of the elevated 85Kr concentrations was attributed to the emplacement hole nearest the 

Area 20 sampler (U-20a), which is about 3,660 m (12,000 ft) to the south. To determine if 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 (ORAUT 2003) 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 
Document No. ORAUT-TKBS-0008-4 Revision No. 04 Effective Date: 02/19/2020 Page 37 of 94

 

the sampler location. The potential dose resulting from submersion in the plume of this noble gas is 

entirely external because the gas is not readily assimilated by the body. Therefore, these source 

concentrations were converted to potential organ doses using submersion DCFs from Federal 

Guidance Report No. 12 (Eckerman and Ryman 1993). 

As Table 4-16 indicates, the upper bound annual organ doses for all years are less than 1 mrem. For 

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 in the individual dosimetry record.

Uncertainty estimates for a number of 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 4-16 are less than 1 mrem and, therefore, inconsequential for employee dose 

reconstruction.

Xenon-133

Although the vast majority of the 133Xe measurements were below the lower limit of detection (i.e., 

3.7 × 10-7 Bq/m3

), elevated atmospheric concentrations of 133Xe were occasionally measured in 

Area 20. As with 85Kr, these elevated concentrations were attributed to seepage from 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 NTS locations. 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 (Fauver 1987). This section discusses the 

results of these surveys and those in 1991 and 1992 (Lyons 1992a, 1992b).

4.4.1 Underground Activities

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, 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 

2008). 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 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 

but, since 1971, has been used only as an underground research facility (ORAUT 2008). 
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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. 

At present, the U.S. Department of Defense 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 2008). 

The U1a Complex in Area 1 is an underground laboratory of horizontal tunnels about 0.5 mi long 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 2008).

4.4.2 Radon Measurements

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 seconds, occurs in the 232Th chain. The decay daughters associated with 

222Rn 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. 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 1987).

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 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, 

and 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 enables 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 1988, p. 19). The WL unit considers this factor. However, because the 

DCFs 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 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. 
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4.4.3 Underground Radon Concentrations

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 (Fauver 1987, Figure 2, p. 7). The RDCs 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 4-17).

Table 4-17. Results of experiment to determine the effects of ventilation 

conditions on RDCs in the G-Tunnel inclined drift.

a

Location

Ventilation conditions and

radon progeny integrated sample average 

(WL)

Rock mechanics drift at 0+52 Alternating 0.19,

Continuous 0.05

Average Alternating 0.13,

Continuous 0.05

a. Source: Fauver (1987, Table 3).

Lyons (1992a, 1992b) reported additional radon measurements from 1991 and 1992. Although the 

G-Tunnel complex remained inactive in 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 in 1992 in T-Tunnel, which was inactive 

that year.

Tables 4-18 and 4-19 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 0.008 and 0.003 WL, respectively, and the maximum 

concentrations in N- and P-Tunnels were 0.013 and 0.006 WL, respectively.

Table 4-18. RDCs for N-Tunnel in 1991 and 1992.

a

July–December 1991

N-Tunnel location

Rn-222 

average 

(WL)

Rn-222 

maximum 

(WL)

Rn-220 

average 

(WL)

Rn-220 

maximum 

(WL)

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
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January–June 1992

N-Tunnel location

Rn-222 

average 

(WL)

Rn-222 

maximum 

(WL)

Rn-220 

average 

(WL)

Rn-220 

maximum 

(WL)

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

Average 0.008 0.013 0.014 0.025

a. GZ = ground zero; .LOS = line of sight.

Table 4-19. RDCs for P-Tunnel in 1991 and 1992.

a

July–December 1991

P-Tunnel location

Rn-222 

average 

(WL)

Rn-222 

maximum 

(WL)

Rn-220 

average 

(WL)

Rn-220 

maximum 

(WL)

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

January–June 1992

P-Tunnel location

Rn-222 

average 

(WL)

Rn-222 

maximum 

(WL)

Rn-220 

average 

(WL)

Rn-220 

maximum 

(WL)

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

Average 0.003 0.006 0.006 0.011

a. HPD = Health Protection Department; LOS = line of sight. 

4.4.4 Underground Worker Exposure to Radon

Although measurements were periodically performed in the tunnel complexes to ensure adequate 

worker protection, neither DOE nor its predecessor agencies attempted to quantify or record 

occupational exposures to radon and its daughters. Therefore, in accordance with DCAS-TIB-011, 

Dose Conversion Factors for Radon WLM (NIOSH 2018), dose reconstructors should adjust the dose 

for all cancer organs 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 4-20 lists airborne 

222Rn RDCs based on measurement results that are favorable to claimants. The values in Table 4-20

pertain only to 222Rn and its daughter products. Based on measurement results shown in Tables 4-18 

and 4-19, 220Rn WL concentrations are assumed to be present at twice the 222Rn WL concentrations 

and should be evaluated in addition to the 222Rn exposures.
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Table 4-20. Average annual 222Rn exposures and uncertainties for internal dose 

reconstruction for miners and tunnel workers (Fauver 1987). 

Tunnel complex

RDC concentration

(WL)

Annual exposure

(WLM)a

Uncertainty

GSDb

G (before 1984) 0.13 1.56 1.52

G (1984 and later) 0.05 0.60 1.52

N 0.008 0.096 1.52

P 0.003 0.036 1.52

T 0.01 0.12 1.52

Unidentifiedc (before 1984) 0.13 1.56 1.52

Unidentifiedc (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 if underground work location is not known.

For G-Tunnel workers, annual exposures (assuming a full year of underground activity) are assumed 

to be 1.56 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 assumption was made that the RDCs in U1a would be similar to those in G-Tunnel 

(which is mined in highly fractured tuff) [1], which is favorable to claimants. 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.

Based on measurements shown in Table 4-19, for purposes of dose reconstruction, 220Rn should be 

assumed to be present at twice the concentrations of 222Rn shown in Table 4-20. For dose 

reconstruction, the GSD of the values in Table 4-20 can be estimated using Equation 4-1 

(Section 4.2.1.2.1). Under the assumption that the 50th-percentile expected annual exposures are 

those in Table 4-20 and that the 95th-percentile values are twice the values in Table 4-20, the GSD 

for all values in Table 4-20 is 1.52. 

Lastly, when a best estimate is required, and occupancy factor may be applied based on tunnel entry 

logs contained within the dosimetry data provided by the DOE. The derived occupancy factor should 

assume full occupancy at 2,040 hours per year based on the assumption of 170 hours of exposure 

per month used to derive the Annual Exposure (WLM) values provided in Table 4-20. The derived 

occupancy factor would be the ratio of the hours worked in a tunnel to the full occupancy hours (i.e., 

2,040 hours). For example, if the logs indicate that the worker was in the tunnel for 1,020 hours in a 

year, the derived occupancy factor would be 0.50. The Annual Exposure (WLM) from Table 4-20 

would then be multiplied by the occupancy factor. In some instances, the derived occupancy factor 

could be greater than 1.00. In this case, the Annual Exposure (WLM) from Table 4-20 would be 

increased by multiplication by the derived occupancy factor. In addition, the Unidentified tunnel 

exposure values from Table 4-20 should not be used since the actual tunnels worked would have 

been identified in the tunnel entry logs used to derive the occupancy factor.

Similarly, for a bounding overestimate, the assumption is made that the worker spends 2,600 hours 

per year (50 hours per week for 52 weeks) underground. This would result in a derived occupancy 

factor of 1.275.
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4.4.4.1 Gravel Gerties

As discussed in Part 2, Section 2.2.5 of this site profile (ORAUT 2008), underground bunkers were 

constructed in Area 5 for testing containment capabilities involving accidental explosion of highexplosive 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 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, they have 

not been activated for device assembly.

Because the Area 5 gravel gerties were constructed for testing 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 a significant time, it should be assumed that they would have been exposed to 

elevated levels of radon, and this radon exposure should be evaluated and included in the dose 

reconstruction.

Although radon measurements are not available for inside the NTS gravel gerties, because of the 

similarities of construction, application of the radon concentrations measured in the Pantex gravel 

gerties is assumed to be appropriate. Therefore, for a 12-month period for which it can be determined 

that a claimant worked in the NTS gravel gerties, a radon exposure of 0.072 WLM should be assigned 

(ORAUT 2015, Section 5.3.3). This exposure can 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 

2015, Section 5.3.3). In addition, when a best estimate is required, and occupancy factor may be 

applied based on entry logs contained within the dosimetry data provided by the DOE. This 

occupancy factor as well as the bounding overestimate occupancy factor should be derived in a 

similar fashion as described in Section 4.4.4 for tunnel entries.

4.4.4.2 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 used 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 about 5 L/min and counted on an EDA 

Instruments Incorporated Radon Detector (Model RD200). The Rolle method allowed a choice of 

several different analysis regimes. The regime chosen for tunnel measurements was a 5-minute 

sampling time, a 6-minute decay time, and a 5-minute counting time. Using appropriate correction 

factors, the RDC in WL can be evaluated 11 minutes after sample collection. The values obtained 

using the Rolle method were periodically checked using the Kusnetz (Fauver 1987, p. 3) method, 

which also used a single count of an air filter but allowed for a 40-to-90-minute decay time. These 

measurement techniques have good accuracy and a relative standard deviation of less than 15% 

(Fauver 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 picocuries per liter. A radon progeny integrating sampling unit (RPISU) measured the RDC in WL.
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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% 

confidence level is about 0.3 pCi/L for a 1-week sample with a relative standard deviation of 20% 

(Fauver 1987, p. 4). The RPISU is a low-volume air sampler that draws air through a 0.65-μm-poresize 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 about 0.0001 WL for a 

1-week sampling period with a relative standard deviation of 10% (Fauver 1987, p. 4). The RPISU 

technique is recommended for the measurement of RDCs by the DOE Technical Measurements 

Center, Division of Remedial Action Projects (Fauver 1987, p. 4).

4.4.5 Underground Worker Exposure to Other Radionuclides

In addition to exposure to elevated levels of radon, underground workers could have been exposed to 

other sources of airborne contamination. Because the breathing air in the underground environment 

was drawn from the outside environment, underground workers would have been exposed to the 

same inhalation intakes as aboveground workers (see Sections 4.2.1.2.2 through 4.2.1.2.6). 

4.5 INSTRUCTION TO DOSE RECONSTRUCTORS FOR ASSIGNMENT OF EXTERNAL 

AMBIENT DOSES AND ENVIRONMENTAL INTAKES AND RESULTANT DOSES

4.5.1 Assigning External Ambient Dose

Starting in April 1957 and ending in 1992, all personnel entering NTS were monitored for external 

radiation exposure (ORAUT 2012). Because unexposed control badges were processed 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 before 1993. Before the institution of universal badging in April 1957 for 

employees who were not monitored, coworker doses should be assigned based on the values in 

ORAUT (2012, Table 6-11) and no additional ambient dose should be assigned through 1992. 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 appropriately included in the 

dosimeter doses beginning in April 1957 through December 1992. After 1992, ambient external dose 

should be added to dose reconstruction for unmonitored workers in accordance with the guidelines for 

assigning dose provided below.

The assignment of a best estimate must take into account all available records pertinent to 

determination of work location. For employees who worked in multiple areas of the site, if the 

employee records and claimant interview do not provide enough information to determine specific 

work locations, values representing a site average are appropriate as a best estimate. Otherwise, the 

annual dose should be calculated according to the fraction of time spent in different areas of the site. 

The assumption should be made that the energy employee worked 50 hr/wk. In addition, 50 wk/yr

should be assumed (to reasonably account for holiday and/or vacation time), for a total of 2,500 work

hours per year as a best estimate. Partial years of employment should be scaled accordingly. In 

addition, if the claimant interview or other available information indicates definitively that fewer or 

more hours were worked than the default assumption, this information should be used accordingly. 

For example, if it is indicated that the employee spent half of the time at offsite locations where only 

natural background radiation levels existed, only 25 hours of exposure per week should be assumed. 

The calculated annual dose should be multiplied by the appropriate exposure (R)-to-organ DCF 

documented in OCAS-IG-001 for an isotropic exposure geometry. If the supporting documentation for 

all of the elements necessary for a best-estimate dose reconstruction does not exist, conservative 

assumptions should be applied in relation to work location and area conditions, erring in favor of the 

claimant.
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External ambient doses for each year are available in Table 4-15 and can be used for workers with 

continuous employment but nonexistent or incomplete monitoring records. For example, EG&G 

employees were likely to be stationed in Las Vegas and received dosimeters when they traveled to 

NTS. 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 

at least one NTS dosimeter. Often no additional records are provided for these employees by Sandia 

National Laboratories, the responsible organization for dosimetry at TTR. For REECo employees 

assigned at TTR, ambient dose (ORAUT 2013) can be assigned as appropriate. 

Assignment of dosimeters to all persons at NTS began in April 1957 and ended in December 1992 

(ORAUT 2012). During this time, readings from control badges at Mercury were used to subtract 

ambient background dose from the personnel dosimeters (ORAUT 2012). For this reason, from April 

1957 to December 1992, ambient background external radiation dose should not be included in dose 

reconstructions. Because no documentation is available before 1967 (see Table 4-15), ambient 

external doses should not be assigned under any circumstances from 1951 through 1956. After 1992, 

ambient background external dose should be assigned to all unmonitored workers in accordance with 

the guidelines for Table 4-15.

4.5.2 Assigning Internal Ambient Dose

With the exception of cases that can be worked after 1992 using the bounding assumptions in 

ORAUT-OTIB-0018 (ORAUT 2005), environmental inhalation and ingestion intakes listed in Tables 4-

7 and 4-11, respectively, shall be applied starting in 1963. The intake values in Tables 4-7 and 4-11 

are to be considered bounding and should be applied in IREP as constants. For a best estimate, 10% 

of the intakes provided in Tables 4-7 and 4-11 should be assigned. The best-estimate intakes should 

be applied in IREP as a lognormal distribution with a GSD of 3.0.

To correct for exposure to short-lived fission and activation products, the dose reconstructor should 

add annual doses that were greater than 0.001 rem from Tables 4-9 and 4-14 as 30-to-250-keV 

photons with a constant distribution. These doses shall be entered into the IREP analysis as 

constants because they can be considered bounding overestimates of the actual 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 (SRDB).

[1] Rollins, Eugene M. Oak Ridge Associated Universities (ORAU) Team. Senior Health 

Physicist, 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.
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Hicks, H. G., 1981a, Results of Calculations of External Gamma Radiation Exposure Rates from 

Fallout and the Related Radionuclide Compositions, Operation Buster-Jangle, 1951, UCRL53152 Pt. 2, University of California, Lawrence Livermore National Laboratory, Livermore, 

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Hicks, H. G., 1981b, Results of Calculations of External Gamma Radiation Exposure Rates from 

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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, 

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Laboratory, Livermore, California, July. [SRDB Ref ID: 1851]

Hicks, H. G., 1981d, Results of Calculations of External Gamma Radiation Exposure Rates from 

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 

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Hicks, H. G, 1984, Results of Calculations of External Gamma Radiation Exposure Rates from Local 

Fallout and the Related Radionuclide Compositions of Selected U.S. Pacific Events, UCRL53505, University of California, Lawrence Livermore National Laboratory, Livermore, 

California, February. [SRDB Ref ID: 1856]

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the Energy Employees Occupational Illness Compensation Program Act, Designating a Class 

of Employees from Nevada Test Site, Mercury, Nevada, U.S. Department of Health and 

Human Services, Office of the Secretary, Washington, D.C., June 26. [SRDB Ref ID: 39411]

Lyons, C. L., 1992a, Operational Air Sampling Report, January 1 Through June 30, 1992, 

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McArthur, R. D., and J. F. Kordas, 1983, Radionuclide Inventory and Distribution Program: The 

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Vegas, Nevada, December 28. [SRDB Ref ID: 8493]

McArthur, R. D., and J. F. Kordas, 1985, Nevada Test Site Radionuclide Inventory and Distribution 

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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, February. [SRDB Ref ID: 7860]

McArthur, R. D., and S. W. Mead, 1988, Nevada Test Site Radionuclide Inventory and Distribution 

Program: Report #4, Areas 18 and 20, DOE/NV/10384-22, U.S. Department of Energy, 

Nevada Operations Office, Las Vegas, Nevada, April. [SRDB Ref ID: 7854]

McArthur, R. D., and S. W. Mead, 1989, Nevada Test Site 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. 

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7865]

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and Radon Daughters in Air, Report 97, Bethesda, Maryland, November 15. [SRDB Ref ID: 

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

Compensation Program Act, DCAS-IG-003, Rev. 1, Division of Compensation Analysis and 

Support, Cincinnati, Ohio, October 5. [SRDB Ref ID: 88929]

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Radon WLM, DCAS-TIB-011, Rev. 05, Division of Compensation Analysis and Support, 

Cincinnati, Ohio, June 28. [SRDB Ref ID: 175105]

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with Air Sampling Programs, ORAUT-OTIB-0018, Rev. 01, Oak Ridge, Tennessee, August 9. 

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Retained in the Lung, ORAUT-OTIB-0049, Rev. 01 PC-2, Oak Ridge, Tennessee, November 

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the Energy Employees Occupational Illness Compensation Program Act, Designating a Class 

of Employees from Nevada Test Site, Mercury, Nevada, U.S. Department of Health and 

Human Services, Office of the Secretary, Washington, D.C., April 5. [SRDB Ref ID: 153992]

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Doses from Short-Lived Gaseous Radionuclides and Radioactive Particles Released to the 

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GLOSSARY

absorption

In external dosimetry, process in which radiation energy is imparted to material. In internal 

dosimetry, movement of material to blood regardless of mechanism.

absorption type

Categories for materials according to their rates of absorption from the respiratory tract to the 

blood, which replaced the earlier inhalation clearance classes. Defined by the International 

Commission on Radiological Protection, the absorption types are F: deposited materials that 

are readily absorbed into blood from the respiratory tract (fast solubilization), M: deposited 

materials that have intermediate rates of absorption into blood from the respiratory tract 

(moderate rate of solubilization), and S: deposited materials that are relatively insoluble in the 

respiratory tract (slow solubilization).

accuracy

The characteristics of an analysis or determination that ensures that both the bias and 

precision of the resultant quantity will remain within the specified limits.

ambient gamma radiation

Penetrating gamma radiation in the outside environment. Includes gamma radiation from 

natural cosmic and terrestrial sources as well as synthetic (manmade) sources.

background radiation

Radiation from cosmic sources, naturally occurring radioactive materials including naturally 

occurring radon, and global fallout from the testing of nuclear explosives. Background 

radiation does not include radiation from source, byproduct, or Special Nuclear Materials 

regulated by the U.S. Nuclear Regulatory Commission. The average individual exposure from 

background radiation is about 360 millirem per year.

breathing rate

Amount of air a person breathes in a specified time. In relation to health physics for workers, 

rates typically vary from light (1.2 cubic meters per hour) to heavy (1.7 cubic meters per hour) 

as defined by the International Commission on Radiological Protection.

buildup

Increase in flux or dose due to scattering in the medium.

confidence level

The interval about an estimate of a stated quantity within which the value of the quantity is 

expected to be within a specified probability. See uncertainty.

contamination

Radioactive material in undesired locations including air, soil, buildings, animals, and persons.

curie (Ci)

Traditional unit of radioactivity equal to 37 billion (3.7 × 1010) becquerels, which is 

approximately equal to the activity of 1 gram of pure 226Ra.

dose

In general, the specific amount of energy from ionizing radiation that is absorbed per unit of 

mass. Effective and equivalent doses are in units of rem or sievert; other types of dose are in 

units of rad, rep, or grays.
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dose conversion factor (DCF)

Multiplier for conversion of potential dose to the personal dose equivalent to the organ of 

interest (e.g., liver or colon). In relation to radiography, ratio of dose equivalent in tissue or 

organ to entrance kerma in air at the surface of the person being radiographed.

dose equivalent (H)

In units of rem or sievert, product of absorbed dose in tissue multiplied by a weighting factor 

and sometimes by other modifying factors to account for the potential for a biological effect 

from the absorbed dose. See dose.

dosimeter

Device that measures the quantity of received radiation, usually a holder with radiationabsorbing filters and radiation-sensitive inserts packaged to provide a record of absorbed dose 

received by an individual. See film dosimeter and thermoluminescent dosimeter.

dosimetry

Measurement and calculation of internal and external radiation doses.

element

One of the known chemical substances in which the atoms have the same number of protons. 

Elements cannot be broken down further without changing their chemical properties. 

Chemical symbols for the elements consist of either a single letter or a combination of letters, 

some of which descend from the Latin names [e.g., Au from aurum (gold), Fe from ferrum

(iron)]. This glossary indicates elements by their names. Specific isotopes appear as their 

standard chemical symbols with the number of protons and neutrons in the nucleus. For 

example, the isotope of uranium that contains 92 protons and 143 neutrons can appear as 

235U, U-235, or uranium-235. 

error

Difference between the correct, true, or conventionally accepted value and the measured or 

estimated value. Sometimes used to mean estimated uncertainty. See accuracy and 

uncertainty.

exchange period (frequency)

Period (weekly, biweekly, monthly, quarterly, etc.) for routine exchange of dosimeters.

exposure

(1) In general, the act of being exposed to ionizing radiation; see acute exposure and chronic 

exposure. (2) Measure of the ionization produced by X- and gamma-ray photons in air in units 

of roentgens.

external dose

Dose received from radiation (e.g., photons, electrons, and neutrons) that originates outside 

the body including medical screening examinations.

favorable to claimants

A bias in selecting input parameters and assumptions that would result in a greater probability 

of causation (and hence greater likelihood of compensability) than if more accurate 

parameters or assumptions were used. See probability of causation.

film

(1) In the context of external dosimetry, radiation-sensitive photographic film in a light-tight 

wrapping. See film dosimeter. (2) X-ray film.
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film dosimeter

Package of film for measurement of ionizing radiation exposure for personnel monitoring 

purposes. A film dosimeter can contain two or three films of different sensitivities, and it can 

contain one or more filters that shield parts of the film from certain types of radiation. When 

developed, the film has an image caused by radiation measurable with an optical 

densitometer. Also called film badge.

filter

Material used (1) in a dosimeter to adjust radiation response to provide an improved tissue 

equivalent or dose response and (2) in an X-ray machine to selectively absorb photons from 

the beam to reduce unnecessary exposure of individuals or to improve radiographic quality.

geometric standard deviation (GSD)

In probability theory and statistics, the geometric standard deviation describes the spread of a 

set of numbers whose preferred average is the geometric mean.

gravel gertie

At NTS, a round room with 2-foot-thick concrete walls and a staging area with two staging 

bays connected to the round room. Tests were conducted in the gravel gertie with high 

explosives and uranium devices to measure fallout.

high explosive

Chemical compound or mechanical mixture that, when subjected to heat, impact, friction, 

shock, or other initiation stimulus, undergoes a rapid chemical change resulting in large 

volumes of highly heated gases that exert pressure in the surrounding medium.

ingestion

Process of taking a substance into the body through the mouth.

intake

Radioactive material taken into the body by inhalation, absorption through the skin, injection, 

ingestion, or through wounds.

isotope

One of two or more atoms of a particular element that have the same number of protons 

(atomic number) but different numbers of neutrons in their nuclei (e.g., 234U, 235U, and 238U). 

Isotopes have very nearly the same chemical properties. See element.

kiloelectron-volt (keV)

Unit of particle energy equal to 1,000 (1 × 103

) electron-volts.

megaelectron-volt (MeV)

Unit of particle energy equal to 1 million (1 × 106

) electron-volts.

monitor

Device that detects and usually records the presence of radioactivity or radioactive 

substances.

monitoring

Periodic or continuous determination of the presence or amount of ionizing radiation or 

radioactive contamination in air, surface water, groundwater, soil, sediment, equipment 

surfaces, or personnel (for example, bioassay or alpha scans). In relation to personnel, 
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monitoring includes internal and external dosimetry including interpretation of the 

measurements.

neutron

Basic nucleic particle that is electrically neutral with mass slightly greater than that of a proton. 

There are neutrons in the nuclei of every atom heavier than normal hydrogen. See element. 

nuclide

Stable or unstable isotope of any element. Nuclide relates to the atomic mass, which is the 

sum of the number of protons and neutrons in the nucleus of an atom. A radionuclide is an 

unstable nuclide.

photon

Quantum of electromagnetic energy generally regarded as a discrete particle having zero rest 

mass, no electric charge, and an indefinitely long lifetime. The entire range of electromagnetic 

radiation that extends in frequency from 1023 cycles per second (hertz) to 0 hertz.

probability of causation

For purposes of dose reconstruction for the Energy Employees Occupational Illness 

Compensation Program Act of 2000, the percent likelihood, at the 99th percentile, that a 

worker incurred a particular cancer from occupational exposure to radiation.

progeny

Nuclides that result from decay of other nuclides. Also called decay products and formerly 

called daughter products.

radiation

Subatomic particles and electromagnetic rays (photons) with kinetic energy that interact with 

matter through various mechanisms that involve energy transfer.

radioactive

Of, caused by, or exhibiting radioactivity. 

radioactivity

Property possessed by some elements (e.g., uranium) or isotopes (e.g., 14C) of spontaneously 

emitting energetic particles (electrons or alpha particles) by the disintegration of their atomic 

nuclei. See radionuclide. 

radionuclide

Radioactive nuclide. See radioactive and nuclide. 

rem

Traditional unit of radiation dose equivalent that indicates the biological damage caused by 

radiation equivalent to that caused by 1 rad of high-penetration X-rays multiplied by a quality 

factor. The sievert is the International System unit; 1 rem equals 0.01 sievert. The word 

derives from roentgen equivalent in man; rem is also the plural.

roentgen

Unit of photon (gamma or X-ray) exposure for which the resultant ionization liberates a positive 

or negative charge equal to 2.58 × 10-4 coulombs per kilogram (or 1 electrostatic unit of 

electricity per cubic centimeter) of dry air at 0 degrees Celsius and standard atmospheric 

pressure. An exposure of 1 roentgen is approximately equivalent to an absorbed dose of 

1 rad in soft tissue for higher energy photons (generally greater than 100 kiloelectron-volts).
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scattering

Change in direction of radiation by refraction or reflection, often accompanied by a decrease in 

radiation due to absorption by the refracting or reflecting material.

scintillation counter or detector

Combination of phosphor, a photomultiplier tube, and the associated electronic circuits for 

measuring the light emissions produced in the phosphor by ionizing radiation.

seep

At NTS, uncontrolled slow release of radioactive material with little or no energy. Seeps are 

not visible and can be detected only by measuring for radiation.

shielding

Material or obstruction that absorbs ionizing radiation and tends to protect personnel or 

materials from its effects.

sievert (Sv)

International System unit for dose equivalent, which indicates the biological damage caused 

by radiation. The unit is the radiation value in gray (equal to 1 joule per kilogram) multiplied by 

a weighting factor for the type of radiation and a weighting factor for the tissue; 1 sievert 

equals 100 rem.

thermoluminescent dosimeter (TLD)

Device for measuring radiation dose that consists of a holder containing solid chips of material 

that, when heated, release the stored energy as light. The measurement of this light provides 

a measurement of absorbed dose. TLDs replaced film dosimeters at essentially all U.S. 

Department of Energy sites beginning in the 1960s.

thermoluminescent dosimeter chip

Small block or crystal of lithium fluoride in a thermoluminescent dosimeter. For example, a 

TLD-600 dosimeter contains a chip made from more than 95% 6

Li for neutron radiation 

detection, and a TLD-700 dosimeter contains a chip made from more than 99.9% 7

Li for 

photon and beta radiation detection. Also called crystals.

transuranic (TRU) elements

Elements with atomic numbers above 92 (uranium). Examples include plutonium and 

americium.

uncertainty

Standard deviation of the mean of a set of measurements. The standard error reduces to the 

standard deviation of the measurement when there is only one determination. See accuracy, 

confidence level, and error. Also called standard error.

unmonitored dose

Potential unrecorded dose that could have resulted because a worker was not monitored. 

venting

At NTS, prompt, massive, and uncontrolled releases of radioactive material characterized as 

active releases under pressure, such as when radioactive material is driven out of the ground 

by steam or gas.
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working level (WL)

Unit of concentration in air of the short-lived decay products of 222Rn (218Po, 214Pb, 214Bi, and 

214Po) and 220Rn (216Po, 212Pb, 212Bi, 212Po) defined as any combination of the short-lived 

radioactive progeny of radon or thoron in 1 liter of air, without regard to the degree of 

equilibrium, that results in the ultimate emission of 1.3 × 105 megaelectron-volts of alpha 

energy; 1 working level equals 2.083 × 10-5 joule per cubic meter. 

working level month (WLM)

Unit of exposure to radon progeny defined as exposure for 1 working month (170 working 

hours) to a potential alpha energy concentration of 1 working level; 1 working level month 

equals 1 working level times 170 hours, which is 0.00354 joule-hour per cubic meter. See 

working level.

X-ray

See X-ray radiation.

X-ray radiation

Electromagnetic radiation (photons) produced by bombardment of atoms by accelerated 

particles. X-rays are produced by various mechanisms including bremsstrahlung and electron 

shell transitions within atoms (characteristic X-rays). Once formed, there is no difference 

between X-rays and gamma rays, but gamma photons originate inside the nucleus of an atom.
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ATTACHMENT A 

AMBIENT ENVIRONMENTAL INTAKES AT THE NEVADA TEST SITE 

BASED ON AIR SAMPLING AND SOIL CONTAMINATION DATA

TABLE OF CONTENTS

SECTION TITLE PAGE

A.1 Introduction ............................................................................................................................ 61

A.2 Air Sampling Data and Intake Estimates ................................................................................ 62

A.3 Soil Contamination and Decay Corrections ............................................................................ 67

A.4 Scaling Factors for Inhalation Intake Estimates...................................................................... 67

A.5 Correction for Resuspension for Early Times After Atmospheric Tests................................... 69

A.6 Corrections for Inhalation Dose from Short-lived Fission and Activation Products.................. 71

A.7 Ingestion Pathway.................................................................................................................. 81

A.8 Instruction to Dose Reconstructors for Assignment of Environmental Intakes ........................ 90

LIST OF TABLES

TABLE TITLE PAGE

A-1 Atmospheric concentrations of 239Pu for sampled areas......................................................... 63

A-2 Annual inhalation intakes from 239Pu for sampled areas ......................................................... 65

A-3 Inventory of contaminated soil................................................................................................ 67

A-4 Radionuclide areal soil deposition.......................................................................................... 68

A-5 Radionuclide areal soil deposition decay, corrected to 1963 .................................................. 68

A-6 Abundance of radionuclides in NTS soils relative to 239Pu decay, corrected to 1963 .............. 69

A-7 Scaled inhalation intakes corrected for early resuspension in 1963........................................ 71

A-8 Far-field refractory enrichment ratios...................................................................................... 73

A-9 Organ-specific inhalation dose fission and activation product correction factors..................... 79

A-10 Inhalation dose from short-lived fission and activation products ............................................. 80

A-11 Radionuclide soil concentration by area................................................................................. 81

A-12 Area-specific and maximum annual ingestion rates................................................................ 82

A-13 Annual organ doses from ingestion of 238,239Pu and 241Am...................................................... 83

A-14 Organ-specific ingestion fission and activation product correction factors .............................. 87

A-15 Organ-specific annual ingestion doses for the 90Sr intake of 55.42 Bq/yr ............................... 88

A-16 Organ-specific doses from ingestion of fission and activation products .................................. 89
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ATTACHMENT A

AMBIENT ENVIRONMENTAL INTAKES AT THE NEVADA TEST SITE 

BASED ON AIR SAMPLING AND SOIL CONTAMINATION DATA (continued)

LIST OF FIGURES

FIGURE TITLE PAGE

A-1 Resuspension factor as a function of time after initial deposition............................................ 70

A-2 SMALL BOY – 90Sr fraction of total dose to lungs................................................................... 74

A-3 LITTLE FELLER I – 90Sr fraction of total dose to lungs........................................................... 74

A-4 TURK – 90Sr fraction of total dose to lungs............................................................................. 75

A-5 Strontium-90 fraction of total dose.......................................................................................... 76

A-6 Strontium-90 fraction of total dose.......................................................................................... 76

A-7 Strontium-90 fraction of total dose.......................................................................................... 77

A-8 Strontium-90 fraction of total dose.......................................................................................... 77

A-9 Strontium-90 fraction of total dose to the liver ........................................................................ 78

A-10 Strontium-90 fraction of total dose to the red bone marrow .................................................... 78

A-11 Strontium-90 fraction of total dose to the bone surface .......................................................... 79

A-12 Strontium-90 fraction of total ingestion dose .......................................................................... 84

A-13 Strontium-90 fraction of total ingestion dose to red bone marrow........................................... 84

A-14 Strontium-90 fraction of total ingestion dose to red bone marrow........................................... 85

A-15 Strontium-90 fraction of total ingestion dose .......................................................................... 85

A-16 Strontium-90 fraction of ingestion dose to the upper large intestine ....................................... 86

A-17 Strontium-90 fraction of total ingestion dose .......................................................................... 86
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ATTACHMENT A

AMBIENT ENVIRONMENTAL INTAKES AT THE NEVADA TEST SITE 

BASED ON AIR SAMPLING AND SOIL CONTAMINATION DATA (continued)

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 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 and 

Kordas 1983, 1985; McArthur and Mead 1987, 1988, 1989; McArthur 1991). Because the air 

monitoring data were 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 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 be 

estimated by assuming consumption of contaminated NTS soils. To ensure that inhalation and 

ingestion intakes are not underestimated, the relative abundances of long-lived radionuclides in NTS 

soils that were determined from the 1991 soil contamination data (McArthur 1991) were decaycorrected 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, 1981b, 1981c, 1981d). In addition, a 

correction factor is developed for inhalation intakes that accounts for the phenomenon of early 

resuspension (Anspaugh et al. 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 employees of the primary contractor, 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 is 

likely to represent 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 are likely to have bioassay results in the DOE records.

The following paragraphs discuss:

• Air sampling data from the NTS annual environmental reports from 1971 through 2001 and the 

calculated annual inhalation 239Pu intakes. 

• NTS soil contamination data that were gathered and reduced in the late 1980s (McArthur 

1991); soil contamination data that are decay-corrected to 1963. 

• Methods 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. 

• The method used to account for higher air concentrations due to resuspension for 1963 (the 

first year after suspension of atmospheric testing). 
Document No. ORAUT-TKBS-0008-4 Revision No. 04 Effective Date: 02/19/2020 Page 62 of 94

ATTACHMENT A

AMBIENT ENVIRONMENTAL INTAKES AT THE NEVADA TEST SITE 

BASED ON AIR SAMPLING AND SOIL CONTAMINATION DATA (continued)

• The methods 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. 

• The method used to calculate ingestion intakes from oral consumption of contaminated NTS 

soils; the method used to account for ingestion dose from short-lived fission and activation 

products. 

• 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.

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 and 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 (DOE 1990). 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. If 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 lists 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 occupationally exposed for 

2,000 hr/yr with a breathing rate of 2,400 m3

/yr. Table A-2 lists these calculated intakes.

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 

workweek 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 types S, Super S, or M, depending on which 

type delivers the maximum organ dose. 
ATTACHMENT A

AMBIENT ENVIRONMENTAL INTAKES AT THE NEVADA TEST SITE 

BASED ON AIR SAMPLING AND SOIL CONTAMINATION DATA (continued)

Document No. ORAUT-TKBS-0008-4 Revision No. 0

4 Effective Date: 02/19/2020 Page 63 of 94

Table A-1. Atmospheric concentrations of 239Pu for sampled areas (Bq/m3).a,b

Year Area 1 Area 2 Area 3 Area 4 Area 5 Area 6 Area 7 Area 9 Area 10 Area 11 Area 12

1971 1.37E-05 2.45E-05 6.11E-06 ND 4.37E-06 6.36E-06 ND 2.67E-05 7.70E-06 3.37E-06 6.03E-06

1972 5.48E-06 8.29E-06 1.36E-05 ND 5.18E-06 4.96E-05 ND 1.59E-04 1.07E-05 8.99E-06 2.91E-05

1973 3.22E-06 8.25E-06 7.62E-06 ND 3.23E-06 4.44E-06 ND 3.18E-05 1.63E-06 8.99E-06 1.59E-06

1974 2.96E-06 2.52E-06 4.66E-06 ND 1.85E-06 2.93E-06 ND 7.81E-06 2.11E-06 2.96E-06 2.15E-06

1975 1.07E-06 1.41E-06 5.40E-06 ND 1.59E-06 2.32E-06 ND 6.22E-06 2.11E-06 1.81E-06 1.22E-06

1976 1.06E-05 6.36E-06 3.85E-05 ND 5.55E-06 1.64E-05 ND 1.18E-04 7.88E-06 8.73E-06 1.03E-05

1977 9.99E-07 2.41E-06 9.44E-06 ND 1.24E-06 2.05E-06 ND 9.36E-06 1.70E-06 4.48E-06 7.77E-07

1978 1.74E-06 4.92E-06 1.04E-05 ND 2.26E-06 3.66E-06 ND 2.01E-05 1.07E-05 3.96E-06 2.29E-06

1979 2.74E-06 1.72E-05 6.11E-08 ND 7.59E-07 1.46E-06 1.81E-06 1.94E-05 1.07E-05 1.37E-06 8.88E-07

1980 8.51E-07 9.58E-06 4.14E-08 ND 1.08E-06 1.78E-06 1.67E-06 1.67E-05 ND 1.63E-06 7.03E-07

1981 9.62E-07 2.35E-06 5.18E-06 ND 9.18E-07 1.07E-06 1.22E-06 1.19E-05 ND 1.70E-06 7.40E-07

1982 2.26E-06 1.48E-06 2.31E-06 ND 7.36E-07 1.30E-06 2.74E-06 7.96E-06 ND 1.96E-06 9.99E-07

1983 1.15E-06 2.39E-06 3.42E-06 ND 8.40E-07 1.20E-06 1.22E-06 7.92E-06 ND 6.73E-06 8.51E-07

1984 7.03E-07 3.68E-06 8.44E-06 ND 1.40E-06 7.40E-07 1.41E-06 3.77E-05 ND 2.07E-06 6.29E-07

1985 5.55E-07 2.13E-06 7.88E-06 ND 1.13E-06 9.99E-07 9.99E-07 4.92E-05 ND 1.92E-06 8.51E-07

1986 5.48E-06 2.29E-06 2.02E-05 ND 1.02E-06 6.88E-06 1.78E-06 1.04E-05 ND 9.25E-07 1.18E-06

1987 2.02E-06 8.51E-07 1.04E-04 ND 6.11E-07 9.36E-07 5.92E-07 4.07E-06 ND 1.18E-06 5.92E-07

1988 3.40E-06 1.04E-06 7.92E-06 ND 1.02E-06 8.14E-07 5.55E-07 1.89E-06 ND 4.44E-07 5.92E-07

1989 3.16E-05 2.15E-06 1.99E-05 ND 1.86E-06 2.96E-06 9.99E-06 1.30E-05 ND 2.00E-05 2.52E-05

1990 3.66E-06 1.20E-06 1.28E-05 ND 1.14E-06 1.37E-06 1.50E-06 1.84E-05 1.57E-06 3.96E-06 3.85E-07

1991 1.15E-04 7.59E-07 1.42E-05 ND 1.30E-06 2.68E-06 1.20E-06 1.30E-05 1.99E-08 8.29E-06 5.29E-07

1992 5.55E-06 4.07E-06 8.36E-05 ND 2.97E-06 2.69E-06 1.44E-05 3.26E-05 8.88E-06 3.26E-06 1.41E-06

1993 7.51E-06 1.81E-06 1.47E-05 ND 2.96E-06 3.31E-06 2.89E-06 9.99E-05 6.29E-06 1.55E-05 3.11E-05

1994 ND ND ND ND ND ND ND ND ND ND ND

1995 1.50E-06 4.07E-07 8.99E-06 ND 5.66E-07 1.86E-06 4.44E-07 1.44E-05 1.39E-06 1.48E-06 4.07E-07

1996 2.78E-05 ND 1.12E-05 ND 6.07E-07 1.70E-06 1.67E-05 2.26E-05 2.13E-06 7.03E-07 1.37E-07

1997 ND 3.96E-07 ND ND ND ND ND ND 9.73E-07 4.07E-07 5.55E-08

1998 1.74E-05 ND 2.49E-06 ND 3.52E-06 4.07E-07 1.01E-06 2.72E-05 ND 3.50E-07 ND

1999 ND ND ND 2.18E-06 ND ND ND ND ND ND ND

2000 3.14E-06 ND 9.95E-06 2.18E-06 7.59E-07 1.52E-05 1.26E-06 1.05E-04 ND ND ND

2001 1.68E-05 4.62E-07 4.29E-06 6.92E-06 2.53E-06 3.57E-06 6.28E-07 1.86E-05 3.13E-07 ND ND
ATTACHMENT A

AMBIENT ENVIRONMENTAL INTAKES AT THE NEVADA TEST SITE 

BASED ON AIR SAMPLING AND SOIL CONTAMINATION DATA (continued)

Document No. ORAUT-TKBS-0008-4 Revision No. 0

4 Effective Date: 02/19/2020 Page 64 of 94

Year Area 15 Area 16 Area 18 Area 19 Area 20 Area 23 Area 25 Area 27 Area 28

Site 

averagec

Site 

maximumd

1971 ND 7.14E-06 2.96E-06 8.36E-06 ND 3.46E-06 ND 3.11E-06 1.44E-06 8.35E-06 2.67E-05

1972 ND 6.25E-06 5.37E-06 1.01E-04 ND 1.29E-05 ND 6.14E-06 1.89E-06 2.82E-05 1.59E-04

1973 ND 1.63E-06 1.41E-06 2.65E-06 ND 2.07E-06 ND 1.37E-06 1.22E-06 5.41E-06 3.18E-05

1974 ND 2.22E-06 ND 2.44E-06 ND 3.15E-06 3.18E-06 2.66E-06 4.33E-06 3.20E-06 7.81E-06

1975 ND 1.59E-06 ND 1.24E-06 ND 1.63E-06 2.11E-06 2.52E-06 2.59E-06 2.32E-06 6.22E-06

1976 ND 5.77E-06 ND 3.66E-06 ND 2.10E-05 7.44E-06 5.07E-06 2.33E-06 1.78E-05 1.18E-04

1977 ND 9.62E-07 ND 8.51E-07 ND 9.99E-07 1.22E-06 1.15E-06 8.51E-07 2.56E-06 9.44E-06

1978 ND 1.89E-06 ND 2.28E-06 ND 2.15E-06 2.58E-06 3.48E-06 1.44E-06 4.92E-06 2.01E-05

1979 1.92E-06 7.77E-07 ND 7.22E-07 ND 8.25E-07 6.66E-07 5.92E-07 8.51E-07 3.69E-06 1.94E-05

1980 2.36E-06 ND ND 3.53E-06 ND 1.31E-06 1.13E-06 7.77E-07 5.92E-07 2.73E-06 1.67E-05

1981 3.44E-06 9.62E-07 ND 8.51E-07 ND 1.06E-06 8.99E-07 9.25E-07 5.92E-07 2.17E-06 1.19E-05

1982 1.55E-06 6.29E-07 ND 8.03E-07 ND 7.77E-07 6.29E-07 ND ND 1.87E-06 7.96E-06

1983 8.25E-07 5.55E-07 ND 1.09E-06 ND 1.04E-06 1.11E-06 ND ND 2.17E-06 7.92E-06

1984 1.18E-06 1.04E-06 ND 9.25E-07 ND 1.06E-06 1.18E-06 1.78E-06 ND 4.26E-06 3.77E-05

1985 1.48E-06 7.77E-07 ND 1.30E-06 ND 8.14E-07 9.73E-07 1.15E-06 ND 4.81E-06 4.92E-05

1986 1.44E-06 1.04E-06 ND 8.14E-07 ND 2.42E-06 7.59E-07 8.51E-07 ND 3.83E-06 2.02E-05

1987 6.66E-07 5.18E-07 ND 6.29E-07 ND 5.92E-07 8.51E-07 ND ND 8.43E-06 1.04E-04

1988 8.99E-07 ND ND 5.00E-07 ND 1.30E-06 5.55E-07 5.55E-07 ND 1.53E-06 7.92E-06

1989 8.62E-07 4.07E-07 ND 3.33E-07 ND 4.33E-06 3.50E-07 1.78E-07 ND 8.87E-06 3.16E-05

1990 3.03E-06 2.16E-07 ND 1.45E-07 ND 3.13E-06 4.18E-07 1.30E-06 ND 3.39E-06 1.84E-05

1991 4.55E-06 1.07E-06 ND 3.26E-07 ND 4.92E-07 9.51E-07 2.86E-07 ND 1.03E-05 1.15E-04

1992 4.07E-05 7.03E-07 ND 3.29E-07 ND 4.70E-07 6.70E-07 5.92E-06 ND 1.30E-05 8.36E-05

1993 1.22E-05 4.81E-07 ND 2.85E-07 ND 2.31E-07 1.26E-07 4.07E-07 ND 1.25E-05 9.99E-05

1994 ND ND ND ND ND ND ND ND ND ND ND

1995 5.92E-06 5.18E-07 ND 3.70E-08 ND 2.75E-07 1.67E-07 ND ND 2.56E-06 1.44E-05

1996 8.88E-06 6.66E-08 1.30E-07 ND 1.67E-07 2.31E-07 1.55E-07 9.62E-08 ND 5.82E-06 2.78E-05

1997 1.07E-06 7.03E-09 7.40E-07 ND 7.40E-07 2.22E-08 3.74E-08 5.92E-08 ND 4.11E-07 1.07E-06

1998 2.35E-06 ND 1.14E-06 ND 1.21E-06 4.29E-07 2.32E-07 ND ND 4.81E-06 2.72E-05

1999 4.07E-07 ND 4.07E-07 ND 7.40E-07 ND 2.29E-07 ND ND 7.93E-07 2.18E-06

2000 9.56E-06 ND 3.78E-07 ND 2.43E-07 ND 4.25E-07 ND ND 1.34E-05 1.05E-04

2001 5.18E-07 3.53E-07 6.37E-07 ND 2.80E-07 7.51E-08 1.63E-07 ND ND 3.75E-06 1.86E-05 a. Source: NTS environmental reports in References section.

b. ND = no data. 

c. Values represent the arithmetic average of the area average concentrations for 1971 through 1988 and the arithmetic average of the area maximum concentrations for 1989 through 

2001.

d. Values represent the maximum of the average area concentrations for 1971 through 1988 and the maximum of the maximum area concentrations for 1989 through 2001.
ATTACHMENT A

AMBIENT ENVIRONMENTAL INTAKES AT THE NEVADA TEST SITE 

BASED ON AIR SAMPLING AND SOIL CONTAMINATION DATA (continued)

Document No. ORAUT-TKBS-0008-4 Revision No. 0

4 Effective Date: 02/19/2020 Page 65 of 94

Table A-2. Annual inhalation intakes from 239Pu for sampled areas (Bq).

a

Year Area 1 Area 2 Area 3 Area 4 Area 5 Area 6 Area 7 Area 9 Area 10 Area 11 Area 12

1971 0.0329 0.0587 0.0147 ND 0.0105 0.0153 ND 0.0640 0.0185 0.0081 0.0145

1972 0.0131 0.0199 0.0327 ND 0.0124 0.1190 ND 0.3810 0.0258 0.0216 0.0698

1973 0.0077 0.0198 0.0183 ND 0.0078 0.0107 ND 0.0763 0.0039 0.0216 0.0038

1974 0.0071 0.0060 0.0112 ND 0.0044 0.0070 ND 0.0187 0.0051 0.0071 0.0052

1975 0.0026 0.0034 0.0130 ND 0.0038 0.0056 ND 0.0149 0.0051 0.0044 0.0029

1976 0.0255 0.0153 0.0924 ND 0.0133 0.0392 ND 0.2824 0.0189 0.0210 0.0248

1977 0.0024 0.0058 0.0226 ND 0.0030 0.0049 ND 0.0225 0.0041 0.0107 0.0019

1978 0.0042 0.0118 0.0250 ND 0.0054 0.0088 ND 0.0481 0.0258 0.0095 0.0055

1979 0.0066 0.0413 0.0001 ND 0.0018 0.0035 0.0044 0.0465 0.0258 0.0033 0.0021

1980 0.0020 0.0230 0.0001 ND 0.0026 0.0043 0.0040 0.0401 ND 0.0039 0.0017

1981 0.0023 0.0056 0.0124 ND 0.0022 0.0026 0.0029 0.0286 ND 0.0041 0.0018

1982 0.0054 0.0036 0.0055 ND 0.0018 0.0031 0.0066 0.0191 ND 0.0047 0.0024

1983 0.0028 0.0057 0.0082 ND 0.0020 0.0029 0.0029 0.0190 ND 0.0162 0.0020

1984 0.0017 0.0088 0.0202 ND 0.0034 0.0018 0.0034 0.0906 ND 0.0050 0.0015

1985 0.0013 0.0051 0.0189 ND 0.0027 0.0024 0.0024 0.1181 ND 0.0046 0.0020

1986 0.0131 0.0055 0.0484 ND 0.0025 0.0165 0.0043 0.0249 ND 0.0022 0.0028

1987 0.0048 0.0020 0.2495 ND 0.0015 0.0022 0.0014 0.0098 ND 0.0028 0.0014

1988 0.0082 0.0025 0.0190 ND 0.0025 0.0020 0.0013 0.0045 ND 0.0011 0.0014

1989 0.0759 0.0052 0.0478 ND 0.0045 0.0071 0.0240 0.0311 ND 0.0480 0.0604

1990 0.0088 0.0029 0.0308 ND 0.0027 0.0033 0.0036 0.0442 0.0038 0.0095 0.0009

1991 0.2753 0.0018 0.0340 ND 0.0031 0.0064 0.0029 0.0311 0.0000 0.0199 0.0013

1992 0.0133 0.0098 0.2007 ND 0.0071 0.0065 0.0346 0.0781 0.0213 0.0078 0.0034

1993 0.0180 0.0044 0.0352 ND 0.0071 0.0079 0.0069 0.2398 0.0151 0.0373 0.0746

1994 ND ND ND ND ND ND ND ND ND ND ND

1995 0.0036 0.0010 0.0216 ND 0.0014 0.0045 0.0011 0.0346 0.0033 0.0036 0.0010

1996 0.0666 0.0000 0.0268 ND 0.0015 0.0041 0.0400 0.0542 0.0051 0.0017 0.0003

1997 ND 0.0010 ND ND ND ND ND ND 0.0023 0.0010 0.0001

1998 0.0416 ND 0.0060 ND 0.0084 0.0010 0.0024 0.0653 ND 0.0008 ND

1999 ND ND ND ND ND ND ND ND ND ND ND

2000 0.0075 ND 0.0239 0.0052 0.0018 0.0366 0.0030 0.2509 ND ND ND

2001 0.0404 0.0011 0.0103 0.0166 0.0061 0.0086 0.0015 0.0447 0.0008 ND ND
ATTACHMENT A

AMBIENT ENVIRONMENTAL INTAKES AT THE NEVADA TEST SITE 

BASED ON AIR SAMPLING AND SOIL CONTAMINATION DATA (continued)

Document No. ORAUT-TKBS-0008-4 Revision No. 0

4 Effective Date: 02/19/2020 Page 66 of 94

Year Area 15 Area 16 Area 18 Area 19 Area 20 Area 23 Area 25 Area 27 Area 28

Site 

average

Site 

maximum

1971 ND 0.0171 0.0071 0.0201 ND 0.0083 ND 0.0075 0.0035 0.0200 0.0640

1972 ND 0.0150 0.0129 0.2424 ND 0.0309 ND 0.0147 0.0045 0.0677 0.3810

1973 ND 0.0039 0.0034 0.0063 ND 0.0050 ND 0.0033 0.0029 0.0130 0.0763

1974 ND 0.0053 ND 0.0059 ND 0.0075 0.0076 0.0064 0.0104 0.0077 0.0187

1975 ND 0.0038 ND 0.0030 ND 0.0039 0.0051 0.0060 0.0062 0.0056 0.0149

1976 ND 0.0139 ND 0.0088 ND 0.0504 0.0178 0.0122 0.0056 0.0428 0.2824

1977 ND 0.0023 ND 0.0020 ND 0.0024 0.0029 0.0028 0.0020 0.0062 0.0226

1978 ND 0.0045 ND 0.0055 ND 0.0052 0.0062 0.0083 0.0035 0.0118 0.0481

1979 0.0046 0.0019 ND 0.0017 ND 0.0020 0.0016 0.0014 0.0020 0.0089 0.0465

1980 0.0057 ND ND 0.0085 ND 0.0031 0.0027 0.0019 0.0014 0.0070 0.0401

1981 0.0083 0.0023 ND 0.0020 ND 0.0025 0.0022 0.0022 0.0014 0.0052 0.0286

1982 0.0037 0.0015 ND 0.0019 ND 0.0019 0.0015 ND ND 0.0045 0.0191

1983 0.0020 0.0013 ND 0.0026 ND 0.0025 0.0027 ND ND 0.0052 0.0190

1984 0.0028 0.0025 ND 0.0022 ND 0.0025 0.0028 0.0043 ND 0.0102 0.0906

1985 0.0036 0.0019 ND 0.0031 ND 0.0020 0.0023 0.0028 ND 0.0115 0.1181

1986 0.0035 0.0025 ND 0.0020 ND 0.0058 0.0018 0.0020 ND 0.0092 0.0484

1987 0.0016 0.0012 ND 0.0015 ND 0.0014 0.0020 0.0000 ND 0.0202 0.2495

1988 0.0022 ND ND 0.0012 ND 0.0031 0.0013 0.0013 ND 0.0037 0.0190

1989 0.0021 0.0010 ND 0.0008 ND 0.0104 0.0008 0.0004 ND 0.0213 0.0759

1990 0.0073 0.0005 ND 0.0003 ND 0.0075 0.0010 0.0031 ND 0.0081 0.0442

1991 0.0109 0.0026 ND 0.0008 ND 0.0012 0.0023 0.0007 ND 0.0246 0.2753

1992 0.0977 0.0017 ND 0.0008 ND 0.0011 0.0016 0.0142 ND 0.0312 0.2007

1993 0.0293 0.0012 ND 0.0007 ND 0.0006 0.0003 0.0010 ND 0.0300 0.2398

1994 ND ND ND ND ND ND ND ND ND ND ND

1995 0.0142 0.0012 ND ND ND 0.0007 0.0004 ND ND 0.0061 0.0346

1996 0.0213 0.0002 0.0003 0.0000 0.0004 0.0006 0.0004 0.0002 ND 0.0140 0.0666

1997 0.0026 0.0000 0.0018 ND 0.0018 0.0001 0.0001 0.0001 ND 0.0010 0.0026

1998 0.0056 ND 0.0027 ND 0.0029 0.0010 0.0006 ND ND 0.0115 0.0653

1999 0.0010 ND 0.0010 ND 0.0018 ND 0.0006 ND ND 0.0019 0.0018

2000 0.0229 ND 0.0009 ND 0.0006 ND 0.0010 ND ND 0.0322 0.2509

2001 0.0012 0.0008 0.0015 ND 0.0007 0.0002 0.0004 ND ND 0.0090 0.0447

a. ND = no data.
Document No. ORAUT-TKBS-0008-4 Revision No. 04 Effective Date: 02/19/2020 Page 67 of 94

ATTACHMENT A

AMBIENT ENVIRONMENTAL INTAKES AT THE NEVADA TEST SITE 

BASED ON AIR SAMPLING AND SOIL CONTAMINATION DATA (continued)

A.3 SOIL CONTAMINATION AND DECAY CORRECTIONS

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

and Kordas 1983, 1985; McArthur and Mead 1987, 1988, 1989; McArthur 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. The results in Table A-4 are 

representative of areas of 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 A-3. Inventory of contaminated soil (Bq).a,b

Area Area (m2) Am-241 Pu-238 Pu-239,240 Co-60 Cs-137 Sr-90 Eu-152 Eu-154 Eu-155

1 6.86E+07 1.554E+11 2.41E+11 8.88E+11 4.07E+10 3.26E+11 5.55E+11 5.55E+11 3.7E+09 1.85E+10

2 5.10E+07 1.073E+11 3.18E+11 8.14E+11 4.44E+10 8.88E+11 1.7E+12 5.18E+11 ND 1.48E+10

3 8.37E+07 1.702E+11 1.15E+11 1.37E+12 3.7E+10 4.44E+11 1.22E+12 6.66E+11 3.7E+09 1.85E+10

4 4.14E+07 2.442E+11 4.81E+11 1.48E+12 5.92E+10 4.44E+11 4.81E+11 3.37E+11 ND 7.4E+09

5 7.51E+06 2.22E+10 3.7E+09 1.78E+11 2.22E+10 1.48E+10 3.33E+10 3.7E+11 7.4E+09 ND

6 8.37E+07 6.29E+10 1.22E+11 3.11E+11 7.4E+09 1.04E+11 1.3E+11 ND ND ND

7 5.00E+07 8.14E+10 2.22E+10 5.92E+11 3.7E+10 1.92E+11 3.4E+11 8.14E+11 7.4E+09 1.11E+10

8 3.60E+07 6.29E+11 2.96E+11 4.07E+12 2.11E+11 1.55E+12 9.25E+11 1.63E+11 ND 2.22E+10

9 5.18E+07 1.554E+11 8.14E+10 3.29E+12 2.59E+10 3.22E+11 4.81E+11 8.51E+11 7.4E+09 1.11E+10

10 5.18E+07 7.03E+11 7.03E+11 4.07E+12 3.59E+11 3.11E+12 2.04E+12 8.14E+10 1.11E+10 1.85E+11

11 1.04E+07 1.221E+11 1.85E+10 1.07E+12 ND 1.85E+10 1.11E+10 ND ND ND

12 1.03E+08 2.109E+11 3.15E+11 1.44E+12 4.44E+10 7.4E+11 6.29E+11 ND ND ND

15 9.14E+07 2.96E+11 2.89E+11 2.33E+12 1.11E+10 7.03E+11 8.14E+11 ND ND ND

16 3.70E+07 2.59E+10 5.55E+10 1.37E+11 3.7E+09 1.07E+11 1.37E+11 ND ND ND

17 8.13E+07 1.036E+11 1.67E+11 6.66E+11 3.7E+10 5.55E+11 7.03E+11 ND ND ND

18 7.07E+07 7.03E+11 2.07E+11 3.7E+12 2.59E+10 3.7E+11 6.29E+11 4.07E+10 3.7E+09 2.96E+10

19 3.84E+08 7.77E+11 1.18E+12 5.18E+12 4.07E+10 1.33E+12 1.15E+12 ND ND ND

20 1.61E+07 8.51E+11 1.11E+12 1.52E+12 2.92E+11 2.04E+11 1.59E+11 4.81E+11 5.92E+10 1.78E+11

25 2.33E+06 ND ND ND ND 7.4E+09 3.7E+09 1.48E+10 ND ND

26 5.18E+05 ND ND ND ND ND ND ND ND ND

30 7.77E+04 1.184E+11 1.67E+11 5.18E+11 2.96E+10 5.55E+10 4.81E+10 2.59E+10 3.7E+09 7.4E+09

a. Source: McArthur (1991).

b. ND = no data.

Because the data in Table A-3 are representative of 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 lists the decay-corrected areal soil deposition.

A.4 SCALING FACTORS FOR INHALATION INTAKE ESTIMATES

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

NTS soils, scaling factors were developed to estimate potential intakes for these radionuclides based 

on their relative abundances in comparison with 239Pu with soil contamination data (McArthur 1991) 

decay-corrected to 1963. These area-specific ratios are listed 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 potential environmental intakes of all radionuclides important to dose (Table A-6) that have 

been identified as persistent in NTS soils. For 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. 
Document No. ORAUT-TKBS-0008-4 Revision No. 04 Effective Date: 02/19/2020 Page 68 of 94

ATTACHMENT A

AMBIENT ENVIRONMENTAL INTAKES AT THE NEVADA TEST SITE 

BASED ON AIR SAMPLING AND SOIL CONTAMINATION DATA (continued)

Table A-4. Radionuclide areal soil deposition (Bq/m2).a

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 ND 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 ND 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 ND

6 7.52E+02 1.46E+03 3.72E+03 8.85E+01 1.24E+03 1.55E+03 ND ND ND

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 ND 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 ND 1.79E+03 1.07E+03 ND ND ND

12 2.06E+03 3.07E+03 1.41E+04 4.33E+02 7.22E+03 6.13E+03 ND ND ND

15 3.24E+03 3.16E+03 2.55E+04 1.21E+02 7.69E+03 8.90E+03 ND ND ND

16 6.99E+02 1.50E+03 3.70E+03 9.99E+01 2.90E+03 3.70E+03 ND ND ND

17 1.27E+03 2.05E+03 8.19E+03 4.55E+02 6.82E+03 8.64E+03 ND ND ND

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 ND ND ND

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 ND ND ND ND 3.17E+03 1.59E+03 6.35E+03 ND ND

26 ND ND ND ND ND ND ND ND ND

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

a. ND = no data.

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

a

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 NDa 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 ND 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 ND

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

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 ND 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 ND 3.41E+03 2.09E+03 ND ND ND

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

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

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

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

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 ND ND ND

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 ND ND ND ND 6.06E+03 3.09E+03 2.72E+04 ND ND

26 ND ND ND ND ND ND ND ND ND

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. ND = no data.
Document No. ORAUT-TKBS-0008-4 Revision No. 04 Effective Date: 02/19/2020 Page 69 of 94

ATTACHMENT A

AMBIENT ENVIRONMENTAL INTAKES AT THE NEVADA TEST SITE 

BASED ON AIR SAMPLING AND SOIL CONTAMINATION DATA (continued)

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

a

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 ND 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 ND 0.250

5 0.131 0.026 1.000 4.958 0.159 0.365 8.925 0.378 ND

6 0.212 0.490 1.000 0.944 0.636 0.810 ND ND ND

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 ND 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 ND 0.033 0.020 ND ND ND

12 0.153 0.272 1.000 1.220 0.978 0.848 ND ND ND

15 0.133 0.154 1.000 0.189 0.575 0.679 ND ND ND

16 0.198 0.505 1.000 1.072 1.495 1.945 ND ND ND

17 0.163 0.312 1.000 2.204 1.590 2.053 ND ND ND

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 ND ND ND

20 0.586 0.912 1.000 7.643 0.256 0.204 1.358 0.354 5.849

25 ND ND ND ND ND ND ND ND ND

26 ND ND ND ND ND ND ND ND ND

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

a. ND = no data.

A.5 CORRECTION FOR RESUSPENSION FOR EARLY TIMES AFTER ATMOSPHERIC TESTS

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

radionuclides in air has been noted to display a strong time dependence early after deposition, and 

this pathway might be important for reoccupation of contaminated property. Anspaugh et al. also 

stated that there has been no universal agreement that resuspension is an important pathway but that 

it is now generally accepted that there are a few instances in which 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 et 

al. (2002) stated that the resuspension factor model has been widely used to predict the concentration 

of resuspended radionuclides early 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 et al. 

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

periods (such as those in this attachment).

Anspaugh et al. (2002) presented several resuspension models that have been proposed but 

concluded 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 
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ATTACHMENT A

AMBIENT ENVIRONMENTAL INTAKES AT THE NEVADA TEST SITE 

BASED ON AIR SAMPLING AND SOIL CONTAMINATION DATA (continued)

1980s, Anspaugh et al. proposed a resuspension model that more accurately describes the observed 

results over the entire timespan of the expanded dataset for NTS:

(A-1) 5 0.07 9 0.003 9 1 1 (1 10 )e (6 10 )e (1 10 ) (1 10 ) m t t Sf

− − − − − ± − = × + × +× ××    

A depiction of Equation A-1 is provided in Figure A-1 from time t equals zero to 1,000 days after 

detonation. As Figure A-1 shows, the resuspension factor Sf ranges from about 10-5 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 the mass loading approach is more predictive at times long after initial 

deposition and the resuspension proposed by Anspaugh et al. (2002) is predictive of the observed 

results over the expanded dataset (including those developed in the 1980s), 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 air monitoring data are available (i.e., 1971 

through 2001; see Section A.2). When the Anspaugh et al. 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 

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 

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

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
Document No. ORAUT-TKBS-0008-4 Revision No. 04 Effective Date: 02/19/2020 Page 71 of 94

ATTACHMENT A

AMBIENT ENVIRONMENTAL INTAKES AT THE NEVADA TEST SITE 

BASED ON AIR SAMPLING AND SOIL CONTAMINATION DATA (continued)

180 because the last atmospheric tests at NTS were in July 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 et al. 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, 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 listed 

in Table A-7. If necessary, these intakes can be prorated for times less than a year for a best 

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

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

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 (144Ce, 106Ru, etc.) that have not persisted in 

NTS soils in measurable quantities. For 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 

developed by Hicks (1981e). Hicks provided 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 DCF (Bunker 1999) to determine the relative importance of each of the 

177 radionuclides to total organ dose. With the exception of 90Sr, the inhalation DCFs were 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 ICRP Publication 68 organ DCFs represent the 

50-year committed dose, the IMBA computer program was used to develop the annual organ-specific 

DCF for 241Am, which was used in the calculations to better represent the relative importance of 

241Am, which delivers dose to the affected organ over long periods. Similarly, annual organ-specific 

DCFs were developed for 90Sr to provide a more accurate time-dependent 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 chemical and 

physical separation of the radionuclides in the fireball in the 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 krypton and xenon) and their 

progeny condense in 6 to 8 minutes and begin falling to the surface (Hicks 1981e). Because the 

Hicks data were developed to estimate offsite levels of fallout and resultant dose, fractionation effects 

were simulated in these data by the removal of a fraction of the refractory nuclides from the calculated 
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ATTACHMENT A

AMBIENT ENVIRONMENTAL INTAKES AT THE NEVADA TEST SITE 

BASED ON AIR SAMPLING AND SOIL CONTAMINATION DATA (continued)

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 (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 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 were 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 are 

enriched with refractory elements.

In Table A-8, the radionuclides with ratios less than 1.0 represent the radionuclides that condense as 

refractory elements while the radionuclides with ratios greater than 1.0 are assumed to behave as 

volatile elements. Although strontium is a refractory element, for these calculations the assumption is 

that it behaves like a volatile element because its fission precursors are noble gases (i.e., volatile 

elements). 

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 (1984) provided time-dependent 

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 x = + 0.0001 0.0074 (A-2)

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, 1981b, 1981c, 1981d). 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) and 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 (1984) data for the STORAX SMALL BOY test were 
ATTACHMENT A

AMBIENT ENVIRONMENTAL INTAKES AT THE NEVADA TEST SITE 

BASED ON AIR SAMPLING AND SOIL CONTAMINATION DATA (continued)

Document No. ORAUT-TKBS-0008-4 Revision No. 0

4 Effective Date: 02/19/2020 Page 73 of 94

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

Radionuclide Ratio

Be-7 0.68

Na-24 0.68

Mn-54 0.68

Fe-55 0.68

Fe-59 0.68

Cu-64 0.68

Cu-67 0.68

W-181 1.36

W-185 1.35

W-187 1.36

W-188 1.36

U-237 0.68

U-239 0.68

U-240 0.68

Np-239 0.68

Np240m 0.69

Np-240 0.68

Am-241 0.68

Cm-242 0.68

Ge-75 1.36

Ge-77 1.36

As-77 1.36

Se-77m 1.36

Ge-78 1.36

As-78 1.36

As-79 1.37

Se-79m 1.36

Br-80 1.36

Se-81m 1.36

Se-81 1.36

Br-82 1.36

Se-83 1.36

Br-83 1.36

Kr-83m 1.37

Radionuclide Ratio

Br-84 1.36

Kr85m 1.36

Kr-87 1.36

Kr-88 1.36

Rb-88 1.36

Rb-89 1.36

Sr-89 1.36

Sr-90 1.36

Sr-91 1.19

Y-91m 1.19

Y-91 1.19

Sr-92 0.68

Y-92 0.68

Sr-93 0.68

Y-93 0.68

Y-94 0.68

Y-95 0.68

Zr-95 0.68

Nb-95 0.68

Zr-97 0.68

Nb-97m 0.68

Nb-97 0.68

Nb-98 0.68

Mo-99 0.68

Tc-99m 0.68

Mo-101 1.36

Tc-101 1.36

Mo-102 1.36

Tc-102m 1.36

Tc-102 1.36

Ru-103 1.36

Rh-103m 1.36

Rh-105m 1.36

Rh-105 1.36

Radionuclide Ratio

Ru-106 1.36

Rh-106 1.36

Rh-107 1.36

Pd-107m 1.36

Pd-109 1.36

Ag-109m 1.36

Pd-111m 1.37

Pd-111 1.36

Ag-111m 1.36

Ag-111 1.36

Pd-112 1.36

Ag-112 1.36

Ag-113 1.36

Ag-115 1.36

Cd-115m 1.36

Cd-115 1.36

In-115m 1.37

Cd-117 1.36

In-117m 1.36

In-117 1.36

Tc-101 1.36

Mo-102 1.36

Tc-102m 1.36

Tc-102 1.36

Ru-103 1.36

Rh-103m 1.36

Tc-104 1.36

Ru-105 1.36

Rh-105m 1.36

Rh-105 1.36

Ru-106 1.36

Rh-106 1.36

Rh-107 1.36

Pd-107m 1.36

Radionuclide Ratio

Pd-109 1.36

Ag-109m 1.36

Pd-111m 1.37

Pd-111 1.36

Ag-111m 1.36

Ag-111 1.36

Pd-112 1.36

Ag-112 1.36

Ag-113 1.36

Ag-115 1.36

Cd-115m 1.36

Cd-115 1.36

In-115m 1.37

Cd-117 1.36

In-117m 1.36

In-117 1.36

Cd-118 1.36

In-118 1.36

Cd-119 1.36

In-119m 1.36

In-119 1.36

Sn-121 1.36

Sn-123m 1.37

Sn-123 1.36

Sn-125 1.36

Sb-125 1.36

Sb-126 1.36

Sn-127 1.36

Sb-127 1.36

Te-127 1.36

Sn-128 1.36

Sb-128m 1.37

Sb-128 1.37

Sn-129m 1.36

Radionuclide Ratio

Sn-129 1.36

Sb-129 1.36

Te-129m 1.36

Te-129 1.36

Sb-130m 1.36

Sb-130 1.36

I-130 1.36

Sb-131 1.36

Te-131m 1.36

Te-131 1.36

I-131 1.36

Te-132 1.36

I-132 1.36

Te-133m 1.36

Te-133 1.36

I-133 1.36

Xe-133m 1.36

Xe-133 1.36

Te-134 1.36

I-134 1.36

I-135 1.36

Xe-135m 1.36

Xe-135 1.36

Cs-136 1.36

Cs-137 1.36

Ba-137m 1.36

Xe-138 1.36

Cs-138 1.36

Cs-139 1.36

Ba-139 1.36

Ba-140 1.16

La-140 1.16

Ba-141 0.90

La-141 0.90

Radionuclide Ratio

Ce-141 0.90

Ba-142 0.68

La-142 0.68

La-143 0.68

Ce-143 0.68

Pr-143 0.68

Ce-144 0.68

Pr-144 0.68

Pr-145 0.68

Ce-146 0.68

Pr-146 0.68

Pr-147 0.68

Nd-147 0.68

Nd-149 0.68

Pm-149 0.68

Pm-150 0.68

Nd-151 0.68

Pm-151 0.68

Pm-152 0.68

Sm-153 0.68

Sm-155 0.68

Eu-155 0.68

Sm-156 0.68

Eu-156 0.68

Eu-157 0.68

Eu-158 0.68

Eu-159 0.68

Gd-159 0.68

Tb-161 0.68
Document No. ORAUT-TKBS-0008-4 Revision No. 04 Effective Date: 02/19/2020 Page 74 of 94

 

ATTACHMENT A

AMBIENT ENVIRONMENTAL INTAKES AT THE NEVADA TEST SITE 

BASED ON AIR SAMPLING AND SOIL CONTAMINATION DATA (continued)

Figure A-2. SMALL BOY – 90Sr fraction of total dose to lungs. 

 

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-3. LITTLE FELLER I – 90Sr fraction of total dose to lungs. 
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ATTACHMENT A

AMBIENT ENVIRONMENTAL INTAKES AT THE NEVADA TEST SITE 

BASED ON AIR SAMPLING AND SOIL CONTAMINATION DATA (continued)

Figure A-4. TURK – 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

selected to be used to determine the fission and activation product dose correction factor. This 

assumption is 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 for which organ dose 

from environmental intakes is calculated).

Integrating Equation A-2 for SMALL BOY from 0 to 365 days and dividing the result by 365 (the value 

selected to be used to determine the fission and activation product dose correction factor. This 

assumption is 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 for which organ dose 

from environmental intakes is calculated).

That represents the integrated total dose for 1 year, 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 by which the 

lung dose from the 90Sr intake (Table A-7) could be multiplied to account for inhalation dose from 

short-lived fission and activation products. Similar integrations were performed for subsequent years 

through 1972. 

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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ATTACHMENT A

AMBIENT ENVIRONMENTAL INTAKES AT THE NEVADA TEST SITE 

BASED ON AIR SAMPLING AND SOIL CONTAMINATION DATA (continued)

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

 

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-6. Strontium-90 fraction of total dose. 

 

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)
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ATTACHMENT A

AMBIENT ENVIRONMENTAL INTAKES AT THE NEVADA TEST SITE 

BASED ON AIR SAMPLING AND SOIL CONTAMINATION DATA (continued)

Figure A-7. 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-8. Strontium-90 fraction of total dose. 

 

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)
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ATTACHMENT A

AMBIENT ENVIRONMENTAL INTAKES AT THE NEVADA TEST SITE 

BASED ON AIR SAMPLING AND SOIL CONTAMINATION DATA (continued)

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

 

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-10. Strontium-90 fraction of total dose to the red bone marrow. 

 

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
Document No. ORAUT-TKBS-0008-4 Revision No. 04 Effective Date: 02/19/2020 Page 79 of 94

 

ATTACHMENT A

AMBIENT ENVIRONMENTAL INTAKES AT THE NEVADA TEST SITE 

BASED ON AIR SAMPLING AND SOIL CONTAMINATION DATA (continued)

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

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

Organ 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

The IMBA computer program was used to determine the organ doses from the scaled intakes in 

Table A-7 for a period of 10 years. These doses were 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 product doses 

are listed in Table A-10. 
ATTACHMENT A

AMBIENT ENVIRONMENTAL INTAKES AT THE NEVADA TEST SITE 

BASED ON AIR SAMPLING AND SOIL CONTAMINATION DATA (continued)

Document No. ORAUT-TKBS-0008-4 Revision No. 0

4 Effective Date: 02/19/2020 Page 80 of 94

Table A-10. Inhalation dose from short-lived fission and activation products (rem). 

Organ 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
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ATTACHMENT A

AMBIENT ENVIRONMENTAL INTAKES AT THE NEVADA TEST SITE 

BASED ON AIR SAMPLING AND SOIL CONTAMINATION DATA (continued)

The application of the early fission and activation product correction factor through 1972 intakes is 

favorable to claimants because the Hicks (1984) data verifies that after 10 years, more than 90% of

the total 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 becomes 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 and a soil density of 1.5 g/cm3 (DOE 2003). The 

area-specific radionuclide soil concentrations are listed in Table A-11.

Table A-11. Radionuclide soil concentration by area (Bq/g).

a

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 ND 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 ND 0.259

5 0.090 0.018 0.686 3.401 0.109 0.250 6.122 0.259 ND

6 0.023 0.053 0.108 0.102 0.069 0.087 ND ND ND

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 ND 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 ND 0.099 0.060 ND ND ND

12 0.062 0.111 0.408 0.498 0.399 0.346 ND ND ND

15 0.098 0.114 0.740 0.140 0.426 0.502 ND ND ND

16 0.021 0.054 0.107 0.115 0.160 0.209 ND ND ND

17 0.039 0.074 0.238 0.524 0.378 0.488 ND ND ND

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 ND ND ND

20 1.607 2.500 2.741 20.946 0.701 0.559 3.723 0.969 16.031

25 ND ND ND ND 0.176 0.090 0.789 ND ND

26 ND ND ND ND ND ND ND ND ND

30 4.620 7.749 19.340 43.835 3.953 3.493 4.143 1.252 13.804

a. ND = no data.

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 that full-time employment was 250 d/yr, 

annual ingestion can be calculated. The area-specific annual ingestion rates are listed in Table A-12. 

For most radionuclides, Area 30 provided the highest areal deposition and resultant intakes. Area 30 

is relatively small (150 km2

) and inaccessible, and is 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 sand. 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 
Document No. ORAUT-TKBS-0008-4 Revision No. 04 Effective Date: 02/19/2020 Page 82 of 94

 

ATTACHMENT A

AMBIENT ENVIRONMENTAL INTAKES AT THE NEVADA TEST SITE 

BASED ON AIR SAMPLING AND SOIL CONTAMINATION DATA (continued)

Table A-12. Area-specific and maximum annual ingestion rates (Bq/yr).

a

Area Am-241 Pu-238

Pu239/240 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 ND 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 ND 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 ND ND ND

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 ND 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 ND ND ND

12 1.56 2.77 10.20 12.45 9.98 8.65 ND ND ND

15 2.45 2.85 18.49 3.49 10.64 12.56 ND ND ND

16 0.53 1.35 2.68 2.87 4.01 5.21 ND ND ND

17 0.97 1.85 5.94 13.09 9.44 12.19 ND ND ND

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 ND ND ND

20 40.17 62.49 68.51 523.64 17.54 13.98 93.06 24.23 400.77

25 ND ND ND ND 4.39 2.24 19.73 ND ND

26 ND ND ND ND ND ND ND ND ND

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

Max.

b 40.17 62.49 81.99 523.64 83.02 55.42 153.05 24.23 400.77

a. ND = no data.

b. Maximum value with Area 30 excluded.

excluded. The Area 30 intakes should be used only 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 

238,239Pu and 241Am greater than 0.001 rem are listed 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.

As with the inhalation intakes discussed in Section A.5, 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 developed (for the reasons discussed 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 

1962. As in Section A.5 for inhalation dose, organ-specific relationships were developed for the 90Sr 

fraction of total dose. These relationships are shown 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-dependent

relationships were determined for the first year and for the last 9 years. 
Document No. ORAUT-TKBS-0008-4 Revision No. 04 Effective Date: 02/19/2020 Page 83 of 94

 

ATTACHMENT A

AMBIENT ENVIRONMENTAL INTAKES AT THE NEVADA TEST SITE 

BASED ON AIR SAMPLING AND SOIL CONTAMINATION DATA (continued)

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

(rem). 

Year Liver Ovaries Testes RBM

Bone 

surface LLI

1963 <0.001 <0.001 <0.001 <0.001 0.002 0.001

1964 0.001 <0.001 <0.001 <0.001 0.007 0.001

1965 0.002 <0.001 <0.001 0.001 0.011 0.001

1966 0.003 <0.001 <0.001 0.001 0.016 0.001

1967 0.004 <0.001 <0.001 0.002 0.020 0.001

1968 0.005 <0.001 <0.001 0.002 0.024 0.001

1969 0.005 <0.001 <0.001 0.002 0.028 0.001

1970 0.006 <0.001 <0.001 0.002 0.032 0.001

1971 0.007 <0.001 <0.001 0.003 0.035 0.001

1972 0.008 <0.001 <0.001 0.003 0.039 0.001

1973 0.008 <0.001 <0.001 0.003 0.043 0.001

1974 0.009 <0.001 <0.001 0.003 0.046 0.001

1975 0.010 <0.001 <0.001 0.003 0.050 0.001

1976 0.010 <0.001 <0.001 0.004 0.053 0.001

1977 0.011 <0.001 <0.001 0.004 0.057 0.001

1978 0.011 <0.001 <0.001 0.004 0.060 0.001

1979 0.012 <0.001 <0.001 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 <0.001

1995 0.019 0.002 0.002 0.005 0.103 <0.001

1996 0.018 0.002 0.002 0.005 0.101 <0.001

1997 0.018 0.002 0.002 0.004 0.099 <0.001

1998 0.017 0.002 0.002 0.004 0.098 <0.001

1999 0.017 0.001 0.001 0.004 0.096 <0.001

2000 0.017 0.001 0.001 0.004 0.095 <0.001

2001 0.016 0.001 0.001 0.004 0.093 <0.001

2002 0.016 0.001 0.001 0.003 0.092 <0.001

2003 0.016 0.001 0.001 0.003 0.090 <0.001

2004 0.015 0.001 0.001 0.003 0.089 <0.001

2005 0.015 0.001 0.001 0.003 0.088 <0.001

2006 0.015 0.001 0.001 0.003 0.086 <0.001

2007 0.014 0.001 0.001 0.003 0.085 <0.001
Document No. ORAUT-TKBS-0008-4 Revision No. 04 Effective Date: 02/19/2020 Page 84 of 94

 

ATTACHMENT A

AMBIENT ENVIRONMENTAL INTAKES AT THE NEVADA TEST SITE 

BASED ON AIR SAMPLING AND SOIL CONTAMINATION DATA (continued)

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

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

 

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
Document No. ORAUT-TKBS-0008-4 Revision No. 04 Effective Date: 02/19/2020 Page 85 of 94

 

ATTACHMENT A

AMBIENT ENVIRONMENTAL INTAKES AT THE NEVADA TEST SITE 

BASED ON AIR SAMPLING AND SOIL CONTAMINATION DATA (continued)

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

 

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-15. Strontium-90 fraction of total ingestion dose. 
Document No. ORAUT-TKBS-0008-4 Revision No. 04 Effective Date: 02/19/2020 Page 86 of 94

 

ATTACHMENT A

AMBIENT ENVIRONMENTAL INTAKES AT THE NEVADA TEST SITE 

BASED ON AIR SAMPLING AND SOIL CONTAMINATION DATA (continued)

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

 

 

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-17. Strontium-90 fraction of total ingestion dose. 

 

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)
Document No. ORAUT-TKBS-0008-4 Revision No. 04 Effective Date: 02/19/2020 Page 87 of 94

 

ATTACHMENT A

AMBIENT ENVIRONMENTAL INTAKES AT THE NEVADA TEST SITE 

BASED ON AIR SAMPLING AND SOIL CONTAMINATION DATA (continued)

To account for fractionation, the refractory elements included in the STORAX SMALL BOY Hicks 

(1981c) 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 for the inhalation 

correction factors, these radionuclide- and time-dependent relative abundances are multiplied by their 

organ-specific ingestion DCFs (Bunker 1999) to determine the radionuclide-specific, relative 

importance to total dose as a function of time after detonation. Because of the large difference 

between the 50-year committed DCF and the annual DCF for 241Am, the IMBA computer program was 

used to calculate the annual ingestion DCFs, which were then used rather than the 50-year committed 

DCFs. The same method was used for 90Sr annual DCFs. 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.

The ingestion correction factors were evaluated over a 10-year period. This is because 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 if the relative contribution of 144Ce and 106Ru to total dose was less than 3% for all organs. In

addition, at the end of 10 years, more than 95% of the total ingestion dose is delivered by 90Sr and 

137Cs; their dose is 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 product 

ingestion correction factors were developed (Table A-14). As with the inhalation fission and activation

product 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 activation products. 

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

Organ 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 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).
ATTACHMENT A

AMBIENT ENVIRONMENTAL INTAKES AT THE NEVADA TEST SITE 

BASED ON AIR SAMPLING AND SOIL CONTAMINATION DATA (continued)

Document No. ORAUT-TKBS-0008-4 Revision No. 0

4 Effective Date: 02/19/2020 Page 88 of 94

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 bone 

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
ATTACHMENT A

AMBIENT ENVIRONMENTAL INTAKES AT THE NEVADA TEST SITE 

BASED ON AIR SAMPLING AND SOIL CONTAMINATION DATA (continued)

Document No. ORAUT-TKBS-0008-4 Revision No. 0

4 Effective Date: 02/19/2020 Page 89 of 94

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
Document No. ORAUT-TKBS-0008-4 Revision No. 04 Effective Date: 02/19/2020 Page 90 of 94

ATTACHMENT A

AMBIENT ENVIRONMENTAL INTAKES AT THE NEVADA TEST SITE 

BASED ON AIR SAMPLING AND SOIL CONTAMINATION DATA (continued)

A.8 INSTRUCTION TO DOSE RECONSTRUCTORS FOR ASSIGNMENT OF ENVIRONMENTAL 

INTAKES

With the exception of cases that can be worked after 1992 with the use of the bounding assumptions 

from ORAUT-OTIB-0018 (ORAUT 2005), environmental inhalation and ingestion intakes in Tables A7 and A-12, respectively, should be applied starting in 1963. In addition, for applicable years of 

employment and affected organs, the annual doses 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 reasonable overestimates of the actual intakes and doses. These doses need to be 

entered into the IREP analysis and they should be entered as 30-to-250-keV photons with a constant 

distribution. Best estimate intakes are 10% of the intakes in Tables A-7 and A-12 and if assigned, the 

resultant doses should be assigned as a log normal distribution with a GSD of 3.0.
Document No. ORAUT-TKBS-0008-4 Revision No. 04 Effective Date: 02/19/2020 Page 91 of 94

 

ATTACHMENT B 

TOTAL ANNUAL ORGAN DOSES FROM 30 YEARS OF INHALATION AND INGESTION INTAKES 

AND SOURCES OF OVERESTIMATED ORGAN DOSE

In the previous sections, methods were presented to account for organ dose from ambient 

environmental inhalation and ingestion intakes of radioactive materials in the 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 NTS soils are listed in Table B-1. To correct for exposure to short-lived 

fission and activation products, the dose reconstructor should add annual doses that were greater 

than or equal to 0.001 rem from Tables 4-9 and 4-14 as 30-to-250-keV photons with a constant 

distribution. The doses shown in table B-1 are for purposes of illustration only and should not be 

entered into the IREP analysis. These doses will be accounted for by the assignment of the 

maximum intakes in Tables 4-7 and 4-11 and repeated in Tables A-7 and A-12 for all years of 

employment.
ATTACHMENT B 

TOTAL ANNUAL ORGAN DOSES FROM 30 YEARS OF INHALATION AND INGESTION INTAKES AND SOURCES OF 

OVERESTIMATED ORGAN DOSE (continued)

Document No. ORAUT-TKBS-0008-4 Revision No. 0

4 Effective Date: 02/19/2020 Page 92 of 94

Table B-1. Total annual organ doses greater than or equal to 0.001 rem from 30 years of inhalation and ingestion intakes (rem). 

Year Radiation type Liver Ovaries Testes RBM

Bone 

surface LLI ET Lung Colon ET1 ET2 LN(ET) LN(TH)

1963 Alpha 0.001 <0.001 <0.001 <0.001 0.006 0.001 0.004 0.007 0.001 0.005 0.004 <0.001 0.001

1964 Alpha 0.003 <0.001 <0.001 0.002 0.017 0.001 0.009 0.006 0.001 0.003 0.009 0.001 0.003

1965 Alpha 0.005 <0.001 <0.001 0.002 0.026 0.001 0.009 0.005 0.001 0.002 0.009 0.001 0.004

1966 Alpha 0.006 <0.001 <0.001 0.003 0.033 0.001 0.009 0.004 0.001 0.002 0.009 0.002 0.006

1967 Alpha 0.008 <0.001 <0.001 0.003 0.040 0.001 0.009 0.004 0.001 0.002 0.009 0.003 0.007

1968 Alpha 0.009 <0.001 <0.001 0.004 0.046 0.001 0.008 0.004 0.001 0.002 0.008 0.003 0.008

1969 Alpha 0.010 <0.001 <0.001 0.004 0.052 0.001 0.008 0.004 0.001 0.002 0.008 0.004 0.009

1970 Alpha 0.011 <0.001 <0.001 0.004 0.058 0.001 0.008 0.004 0.001 0.002 0.008 0.004 0.011

1971 Alpha 0.012 <0.001 <0.001 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 <0.001 0.006 0.003 <0.001 <0.001 0.006 0.011 0.035
ATTACHMENT B 

TOTAL ANNUAL ORGAN DOSES FROM 30 YEARS OF INHALATION AND INGESTION INTAKES AND SOURCES OF 

OVERESTIMATED ORGAN DOSE (continued)

Document No. ORAUT-TKBS-0008-4 Revision No. 0

4 Effective Date: 02/19/2020 Page 93 of 94

Year Radiation type Liver Ovaries Testes RBM

Bone 

surface LLI ET Lung Colon ET1 ET2 LN(ET) LN(TH)

1994 Alpha 0.030 0.003 0.003 0.008 0.169 <0.001 0.004 0.002 <0.001 <0.001 0.004 0.011 0.035

1995 Alpha 0.030 0.003 0.003 0.008 0.166 <0.001 0.003 0.002 <0.001 <0.001 0.003 0.011 0.035

1996 Alpha 0.029 0.003 0.003 0.007 0.163 <0.001 0.002 0.001 <0.001 <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 <0.001 <0.001 0.001 0.011 0.035

1998 Alpha 0.028 0.002 0.002 0.006 0.158 <0.001 <0.001 <0.001 <0.001 <0.001 <0.001 0.010 0.034

1999 Alpha 0.027 0.002 0.002 0.006 0.155 <0.001 <0.001 <0.001 <0.001 <0.001 <0.001 0.010 0.034

2000 Alpha 0.026 0.002 0.002 0.006 0.153 <0.001 <0.001 <0.001 <0.001 <0.001 <0.001 0.010 0.033

2001 Alpha 0.026 0.002 0.002 0.006 0.151 <0.001 <0.001 <0.001 <0.001 <0.001 <0.001 0.009 0.033

2002 Alpha 0.025 0.002 0.002 0.005 0.148 <0.001 <0.001 <0.001 <0.001 <0.001 <0.001 0.009 0.032

2003 Alpha 0.025 0.002 0.002 0.005 0.146 <0.001 <0.001 <0.001 <0.001 <0.001 <0.001 0.009 0.032

2004 Alpha 0.024 0.002 0.002 0.005 0.144 <0.001 <0.001 <0.001 <0.001 <0.001 <0.001 0.008 0.031

2005 Alpha 0.024 0.002 0.002 0.005 0.142 <0.001 <0.001 <0.001 <0.001 <0.001 <0.001 0.008 0.031

2006 Alpha 0.023 0.002 0.002 0.005 0.140 <0.001 <0.001 <0.001 <0.001 <0.001 <0.001 0.008 0.030

2007 Alpha 0.023 0.002 0.002 0.005 0.138 <0.001 <0.001 <0.001 <0.001 <0.001 <0.001 0.007 0.029

2008 Alpha 0.022 0.002 0.002 0.004 0.136 <0.001 <0.001 <0.001 <0.001 <0.001 <0.001 0.007 0.029

2009 Alpha 0.022 0.002 0.002 0.004 0.134 <0.001 <0.001 <0.001 <0.001 <0.001 <0.001 0.007 0.028

2010 Alpha 0.021 0.002 0.002 0.004 0.132 <0.001 <0.001 <0.001 <0.001 <0.001 <0.001 0.007 0.027

2011 Alpha 0.021 0.002 0.002 0.004 0.130 <0.001 <0.001 <0.001 <0.001 <0.001 <0.001 0.006 0.026

2012 Alpha 0.021 0.002 0.002 0.004 0.128 <0.001 <0.001 <0.001 <0.001 <0.001 <0.001 0.006 0.026

1963 Photons E>250 keV <0.001 <0.001 <0.001 <0.001 <0.001 <0.001 <0.001 <0.001 <0.001 0.002 <0.001 <0.001 <0.001

1964 Photons E>250 keV <0.001 <0.001 <0.001 <0.001 <0.001 <0.001 <0.001 <0.001 <0.001 0.001 <0.001 <0.001 <0.001

1965 Photons E>250 keV <0.001 <0.001 <0.001 <0.001 <0.001 <0.001 <0.001 <0.001 <0.001 0.001 <0.001 <0.001 <0.001

1963 Electrons E>15 keV <0.001 <0.001 <0.001 <0.001 <0.001 <0.001 <0.001 <0.001 <0.001 0.002 <0.001 <0.001 <0.001

1964 Electrons E>15 keV <0.001 <0.001 <0.001 <0.001 <0.001 <0.001 <0.001 <0.001 <0.001 0.001 <0.001 <0.001 <0.001

1965 Electrons E>15 keV <0.001 <0.001 <0.001 <0.001 <0.001 <0.001 <0.001 <0.001 <0.001 <0.001 <0.001 <0.001 <0.001

1966 Electrons E>15 keV <0.001 <0.001 <0.001 <0.001 <0.001 <0.001 <0.001 <0.001 <0.001 <0.001 <0.001 <0.001 <0.001

1967 Electrons E>15 keV <0.001 <0.001 <0.001 <0.001 0.001 <0.001 <0.001 <0.001 <0.001 <0.001 <0.001 <0.001 <0.001

1968 Electrons E>15 keV <0.001 <0.001 <0.001 <0.001 0.001 <0.001 <0.001 <0.001 <0.001 <0.001 <0.001 <0.001 <0.001

1969 Electrons E>15 keV <0.001 <0.001 <0.001 <0.001 0.001 <0.001 <0.001 <0.001 <0.001 <0.001 <0.001 <0.001 <0.001

1970 Electrons E>15 keV <0.001 <0.001 <0.001 <0.001 0.001 <0.001 <0.001 <0.001 <0.001 <0.001 <0.001 <0.001 <0.001

1971 Electrons E>15 keV <0.001 <0.001 <0.001 <0.001 0.001 <0.001 <0.001 <0.001 <0.001 <0.001 <0.001 <0.001 <0.001

1972 Electrons E>15 keV <0.001 <0.001 <0.001 <0.001 0.002 <0.001 <0.001 <0.001 <0.001 <0.001 <0.001 <0.001 <0.001
ATTACHMENT B 

TOTAL ANNUAL ORGAN DOSES FROM 30 YEARS OF INHALATION AND INGESTION INTAKES AND SOURCES OF 

OVERESTIMATED ORGAN DOSE (continued)

Document No. ORAUT-TKBS-0008-4 Revision No. 0

4 Effective Date: 02/19/2020 Page 94 of 94

Year Radiation type Liver Ovaries Testes RBM

Bone 

surface LLI ET Lung Colon ET1 ET2 LN(ET) LN(TH)

1973 Electrons E>15 keV <0.001 <0.001 <0.001 <0.001 0.002 <0.001 <0.001 <0.001 <0.001 <0.001 <0.001 <0.001 <0.001

1974 Electrons E>15 keV <0.001 <0.001 <0.001 0.001 0.002 <0.001 <0.001 <0.001 <0.001 <0.001 <0.001 <0.001 <0.001

1975 Electrons E>15 keV <0.001 <0.001 <0.001 0.001 0.002 <0.001 <0.001 <0.001 <0.001 <0.001 <0.001 <0.001 <0.001

1976 Electrons E>15 keV <0.001 <0.001 <0.001 0.001 0.002 <0.001 <0.001 <0.001 <0.001 <0.001 <0.001 <0.001 <0.001

1977 Electrons E>15 keV <0.001 <0.001 <0.001 0.001 0.002 <0.001 <0.001 <0.001 <0.001 <0.001 <0.001 <0.001 <0.001

1978 Electrons E>15 keV <0.001 <0.001 <0.001 0.001 0.002 <0.001 <0.001 <0.001 <0.001 <0.001 <0.001 <0.001 <0.001

1979 Electrons E>15 keV <0.001 <0.001 <0.001 0.001 0.002 <0.001 <0.001 <0.001 <0.001 <0.001 <0.001 <0.001 <0.001

1980 Electrons E>15 keV <0.001 <0.001 <0.001 0.001 0.002 <0.001 <0.001 <0.001 <0.001 <0.001 <0.001 <0.001 <0.001

1981 Electrons E>15 keV <0.001 <0.001 <0.001 0.001 0.002 <0.001 <0.001 <0.001 <0.001 <0.001 <0.001 <0.001 <0.001

1982 Electrons E>15 keV <0.001 <0.001 <0.001 0.001 0.002 <0.001 <0.001 <0.001 <0.001 <0.001 <0.001 <0.001 <0.001

1983 Electrons E>15 keV <0.001 <0.001 <0.001 0.001 0.002 <0.001 <0.001 <0.001 <0.001 <0.001 <0.001 <0.001 <0.001

1984 Electrons E>15 keV <0.001 <0.001 <0.001 0.001 0.002 <0.001 <0.001 <0.001 <0.001 <0.001 <0.001 <0.001 <0.001

1985 Electrons E>15 keV <0.001 <0.001 <0.001 0.001 0.002 <0.001 <0.001 <0.001 <0.001 <0.001 <0.001 <0.001 <0.001

1986 Electrons E>15 keV <0.001 <0.001 <0.001 0.001 0.002 <0.001 <0.001 <0.001 <0.001 <0.001 <0.001 <0.001 <0.001

1987 Electrons E>15 keV <0.001 <0.001 <0.001 0.001 0.002 <0.001 <0.001 <0.001 <0.001 <0.001 <0.001 <0.001 <0.001

1988 Electrons E>15 keV <0.001 <0.001 <0.001 0.001 0.002 <0.001 <0.001 <0.001 <0.001 <0.001 <0.001 <0.001 <0.001

1989 Electrons E>15 keV <0.001 <0.001 <0.001 0.001 0.002 <0.001 <0.001 <0.001 <0.001 <0.001 <0.001 <0.001 <0.001

1990 Electrons E>15 keV <0.001 <0.001 <0.001 0.001 0.002 <0.001 <0.001 <0.001 <0.001 <0.001 <0.001 <0.001 <0.001

1991 Electrons E>15 keV <0.001 <0.001 <0.001 0.001 0.002 <0.001 <0.001 <0.001 <0.001 <0.001 <0.001 <0.001 <0.001

1992 Electrons E>15 keV <0.001 <0.001 <0.001 0.001 0.002 <0.001 <0.001 <0.001 <0.001 <0.001 <0.001 <0.001 <0.001

1993 Electrons E>15 keV <0.001 <0.001 <0.001 <0.001 0.002 <0.001 <0.001 <0.001 <0.001 <0.001 <0.001 <0.001 <0.001

1994 Electrons E>15 keV <0.001 <0.001 <0.001 <0.001 0.002 <0.001 <0.001 <0.001 <0.001 <0.001 <0.001 <0.001 <0.001

1995 Electrons E>15 keV <0.001 <0.001 <0.001 <0.001 0.002 <0.001 <0.001 <0.001 <0.001 <0.001 <0.001 <0.001 <0.001

1996 Electrons E>15 keV <0.001 <0.001 <0.001 <0.001 0.001 <0.001 <0.001 <0.001 <0.001 <0.001 <0.001 <0.001 <0.001

1997 Electrons E>15 keV <0.001 <0.001 <0.001 <0.001 0.001 <0.001 <0.001 <0.001 <0.001 <0.001 <0.001 <0.001 <0.001

1998 Electrons E>15 keV <0.001 <0.001 <0.001 <0.001 0.001 <0.001 <0.001 <0.001 <0.001 <0.001 <0.001 <0.001 <0.001

1999 Electrons E>15 keV <0.001 <0.001 <0.001 <0.001 0.001 <0.001 <0.001 <0.001 <0.001 <0.001 <0.001 <0.001 <0.001

2000 Electrons E>15 keV <0.001 <0.001 <0.001 <0.001 0.001 <0.001 <0.001 <0.001 <0.001 <0.001 <0.001 <0.001 <0.001