ORAU TEAM

Dose Reconstruction

Project for NIOSH

Oak Ridge Associated Universities I Dade Moeller I MJW Technical Services

Page 1 of 116

DOE Review Release 08/27/2012

Document Title:

Nevada Test Site – Occupational Internal 

Dose

Document Number: ORAUT-TKBS-0008-5

Revision: 03

Effective Date: 08/17/2012

Type of Document: TBD

Supersedes: Revision 02

Subject Expert(s): Eugene M. Rollins

Approval: Signature on File Approval Date: 08/07/2012

Eugene M. Rollins, Document Owner

Concurrence: Signature on File Concurrence Date: 08/03/2012

John M. Byrne, Objective 1 Manager

Concurrence: Signature on File Concurrence Date: 08/02/2012

Edward F. Maher, Objective 3 Manager

Concurrence: Vickie S. Short Signature on File for Concurrence Date: 08/02/2012

Kate Kimpan, Project Director

Approval: Signature on File Approval Date: 08/17/2012

James W. Neton, Associate Director for Science

New Total Rewrite Revision Page Change

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

REVISION AND DISCARD / DESTROY ALL COPIES OF THE PRIOR REVISION.
Document No. ORAUT-TKBS-0008-5 Revision No. 03 Effective Date: 08/17/2012 Page 2 of 116

PUBLICATION RECORD

EFFECTIVE 

DATE

REVISION 

NUMBER DESCRIPTION

09/30/2004 00 Technical Basis Document for the Nevada Test Site – Occupational 

Internal Dose. First approved issue. Incorporates formal internal 

and NIOSH review comments. Initiated by Eugene M. Rollins.

07/20/2006 00 PC-1 Approved page change initiated to incorporate definitions and 

directions for dose reconstruction for non-presumptive cancers that 

are excluded from the 1951 through 1962 Special Exposure Cohort. 

Text was added or modified on pages 8-10 in Section 5.1. Approved 

issue of Rev 00 PC-1. No sections were deleted. 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 07/17/2006

Eugene M. Rollins, TBD Team Leader

Signature on File 07/17/2006

John M. Byrne, Task 3 Manager

Signature on File 07/18/2006

Edward F. Maher, Task 5 Manager

Signature on File 07/19/2006

Kate Kimpan, Project Director

Signature on File 07/20/2006

James W. Neton, Associate Director for Science

05/28/2008 01 Approved Revision 01 to incorporate additional information regarding 

bioassay procedures used and expanded MDAs. Addresses 

comments from the Advisory Board’s contractor, and includes 

guidance to the dose reconstructor in Section 5.6. Adds Attributions 

and Annotations Section 5.7. Incorporates formal internal and 

NIOSH review comments. Adds Sections 5.6.4 and 5.6.5. Updates 

boilerplate language in Section 5-1 and adds reference to OCAS-IG003. 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. Text was added or 

modified on pages 11 and 12 in Section 5.1.2, and pages 51, and 52 

in Sections 5.3 and 5.4. Added Sections 5.1.2 and 5.3. Incorporates 

formal internal and NIOSH review comment. Training required: As 

determined by the Objective Manager. Initiated by Eugene M. 

Rollins. 
Document No. ORAUT-TKBS-0008-5 Revision No. 03 Effective Date: 08/17/2012 Page 3 of 116

08/17/2012 03 Revision initiated to revise ratios of 238Pu and 241Am to that of 239Pu to 

reflect the scaling factors in Table 4-6 of Revision 02 of ORAUTTKBS-0008-4, Nevada Test Site – Occupational Environmental Dose

technical basis document. Revised text in Sections 5.2.2.3 and 

5.6.4.2 to reflect these ratios. Revised text in Section 5.6.3 to be 

consistent with text in Section 5.6.2 about methods available and 

appropriate for assignment during SEC periods. Updated 

appropriate references. Incorporates formal internal and NIOSH 

review comments. Constitutes a total rewrite of the document. 

Training required: As determined by the Objective Manager. 

Initiated by Eugene M. Rollins. 
Document No. ORAUT-TKBS-0008-5 Revision No. 03 Effective Date: 08/17/2012 Page 4 of 116

TABLE OF CONTENTS

SECTION TITLE PAGE

Acronyms and Abbreviations ............................................................................................................... 8

5.1 Introduction ............................................................................................................................ 11

5.1.1 Purpose...................................................................................................................... 12

5.1.2 Special Exposure Cohort Petition Information for the Nevada Test Site...................... 12

5.1.3 Scope......................................................................................................................... 13

5.1.4 Historic Monitoring Perspective .................................................................................. 14

5.1.5 Radionuclides of Concern........................................................................................... 15

5.1.6 Bioassay Program Description.................................................................................... 15

5.2 In Vitro Minimum Detectable Activities and Counting Methods............................................... 18

5.2.1 In Vitro Urine and Fecal Analysis................................................................................ 19

5.2.2 In Vitro Methods for Individual Radionuclides ............................................................. 19

5.2.2.1 In Vitro Bioassay for Iodine......................................................................... 20

5.2.2.2 In Vitro Bioassay for Americium.................................................................. 20

5.2.2.3 In Vitro Bioassay for Plutonium................................................................... 21

5.2.2.4 In Vitro Bioassay for Tritium........................................................................ 22

5.2.2.5 In Vitro Bioassay for Uranium ..................................................................... 23

5.2.2.6 In Vitro Analysis for Strontium..................................................................... 24

5.2.2.7 In Vitro Analysis for Thorium....................................................................... 25

5.2.2.8 In Vitro Analysis for Radium........................................................................ 25

5.2.2.9 In Vitro Gross Fission Product Analysis ...................................................... 25

5.2.2.10 In Vitro Analysis for Gamma Emitters ......................................................... 27

5.2.3 Sample Collection and Correcting for Urinalysis Volume ............................................ 28

5.2.4 Fecal Sample Collection and Analysis ........................................................................ 30

5.3 In Vivo MDAS, Counting Methods, and Reporting Practices .................................................. 30

5.3.1 Whole-Body Counting................................................................................................. 31

5.3.1.1 Effect of Cesium on Whole-Body Counting ................................................. 35

5.3.2 Chest Counting........................................................................................................... 35

5.3.3 Thyroid Counting ........................................................................................................ 36

5.3.3.1 Iodine ......................................................................................................... 37

5.3.4 Wound Counting......................................................................................................... 38

5.4 Personal Air Sampling Data ................................................................................................... 38

5.4.1 Drilling Operations...................................................................................................... 39

5.4.2 Tunnel Operations...................................................................................................... 42

5.4.3 Examples of Air Sampling Results.............................................................................. 43

5.5 Interferences and Uncertainty ................................................................................................ 46

5.5.1 Contamination of Samples.......................................................................................... 46

5.5.2 Uncertainty ................................................................................................................. 46

5.5.3 Less-Than Values....................................................................................................... 46

5.6 Dose Reconstruction Recommendations ............................................................................... 47

5.6.1 Evaluation of Potential Internal Worker Exposure...................................................... 47

5.6.2 Workers with Pre-1963 Employment........................................................................... 47

5.6.3 Workers with Employment From 1963 through 1992 .................................................. 48

5.6.4 Approach to Internal Dose Reconstruction for NTS .................................................... 49
Document No. ORAUT-TKBS-0008-5 Revision No. 03 Effective Date: 08/17/2012 Page 5 of 116

5.6.4.1 Single Positive Result................................................................................. 49

5.6.4.2 For Individuals with Urine Bioassay Results (Positive or <MDA)................. 50

5.6.4.3 For Individuals with Whole-Body Count Results.......................................... 51

5.6.4.4 For Individuals with Lung Count Results..................................................... 51

5.6.5 Examples of the Assignment of Internal Dose for Selected Job Titles......................... 51

5.6.6 Unusual or Unique Working Environments ................................................................. 52

5.6.6.1 Potential for Iodine Exposure...................................................................... 52

5.6.6.2 Cleanup after Safety Tests ......................................................................... 52

5.6.6.3 Nuclear Reactor Development Station........................................................ 53

5.6.6.4 Incidents..................................................................................................... 53

5.7 Attributions and Annotations .................................................................................................. 53

References ........................................................................................................................................ 54

Glossary ............................................................................................................................................ 63

ATTACHMENT A, OCCUPATIONAL INTERNAL DOSE FOR MONITORED 

WORKERS.......................................................................................................... 67
Document No. ORAUT-TKBS-0008-5 Revision No. 03 Effective Date: 08/17/2012 Page 6 of 116

LIST OF TABLES

TABLE TITLE PAGE

5-1 Types of bioassay monitoring and radionuclides .................................................................... 16

5-2 2003 target MDAs for in vitro bioassay sample analysis......................................................... 19

5-3 MDAs for the HNS shadow shield .......................................................................................... 31

5-4 Frequency of nuclide appearance and concentration in subjects counted before January 

1967.......................................................................................................................................32

5-5 Frequency of nuclide appearance and concentration in subjects counted from January to 

August 1967...........................................................................................................................32

5-6 Preliminary calibration sources for 1977 through 1981........................................................... 33

5-7 WBC sensitivities ...................................................................................................................33

5-8 Relative abundance of radioiodines ....................................................................................... 37

5-9 Air sample results from 1957 surface plutonium dispersal test ............................................... 44

5-10 Additional air sampling data, Area 9 balloon test site ............................................................. 44

5-11 Historic air sampling limits...................................................................................................... 45

A-1 Codes for analyte...................................................................................................................69

A-2 Codes for body parts.............................................................................................................. 69

A-3 Codes for radionuclides and sample types............................................................................. 71

A-4 Codes for sample types ......................................................................................................... 71

A-5 Codes for units.......................................................................................................................72

A-6 Special terminology................................................................................................................ 72

A-7 Limits of detection for urine and fecal analysis ....................................................................... 74

A-8 Radiological data reporting..................................................................................................... 78

A-9 Whole-body counting MDAs................................................................................................... 80

A-10 1993 MDAs for chest counting ............................................................................................... 83

A-11 MDAs for thyroid counts......................................................................................................... 83

A-12 Drill-back resuspension and mine back containment loss radionuclides for identification 

versus time after test..............................................................................................................86

A-13 Drill-back resuspension, reentry/mine back resuspension, and decontamination facility, 

isotopes of concern for dose versus time after test................................................................. 86

A-14 Decontamination facility, isotopes for identification versus time after test............................... 90

A-15 Atmospheric weapons test areas, isotopes for identification and of concern for dose............. 90

A-16 Low-level waste site (Area 3) isotopes for identification and of concern for dose ................... 91

A-17 Low-level waste site (Area-5), isotopes for identification and of concern for dose .................. 91

A-18 Radiation instrument calibration facilities, isotopes for identification and of concern for 

dose.......................................................................................................................................92

A-19 Radiation instrument calibration facilities, isotopes for identification and of concern for 

dose.......................................................................................................................................93

A-20 Radiochemistry and counting laboratories, isotopes for identification and of concern for 

dose.......................................................................................................................................93

A-21 Radionuclides of concern for dose, summary list ................................................................... 94

A-22 Current nuclides of concern for NTS locations ....................................................................... 94

A-23 Fission products associated with above ground testing.......................................................... 96

A-24 Activation products associated with above ground testing...................................................... 96

A-25 Actinide elements associated with above ground testing........................................................ 96

A-26 Releases from underground tests .......................................................................................... 98

A-27 Historical NTS respiratory protection action levels................................................................ 112

A-28 NTS historical contamination limits....................................................................................... 113

A-29 Inhalation classes for radionuclides found at NTS................................................................ 114

A-30 Fission products up to 1 year old as identified in 1954 documentation ................................. 115
Document No. ORAUT-TKBS-0008-5 Revision No. 03 Effective Date: 08/17/2012 Page 7 of 116

A-31 Other common radionuclides not normally a part of fission products that might be 

present................................................................................................................................. 116

LIST OF FIGURES

FIGURE TITLE PAGE

A-1 NaI detector efficiency, counts per photon disintegration ....................................................... 79

A-2 NaI detector efficiency, cpm per photon nCi........................................................................... 79
Document No. ORAUT-TKBS-0008-5 Revision No. 03 Effective Date: 08/17/2012 Page 8 of 116

ACRONYMS AND ABBREVIATIONS

BEEF Big Explosive Experimental Facility

BN Bechtel Nevada

Bq becquerel

CA contamination area

CAM continuous air monitor

cfm cubic feet per minute

CFR Code of Federal Regulations

Ci curie

cm centimeter

cpm counts per minute

cps counts per second

CWT chest wall thickness

DAC derived air concentration

DL decision level

DOD U.S. Department of Defense

DOE U.S. Department of Energy

DOL U.S. Department of Labor

DPP drift protection plug

dpm disintegrations per minute

DR dose reconstructor

DU depleted uranium

EE Energy Employee

EEOICPA Energy Employees Occupational Illness Compensation Program Act

EG&G Edgerton, Germeshausen, and Greer Corporation

FSN Fenix & Scisson of Nevada

FWHM full-width, half-maximum

g gram

gal gallon

GeLi germanium-lithium 

GFP gross fission product

gK 0.000118 g K-40

GSP gas seal plug

H&N Holmes & Narver

HEPA high-efficiency particulate air

HEU highly enriched uranium

HNS Helgeson Nuclear Services

hr hour

ICRP International Commission on Radiation Protection

IMBA Integrated Modules for Bioassay Analysis

in. inch

IREP Interactive RadioEpidemiological Program

keV kiloelectron-volt, 1,000 electron-volts

kg kilogram
Document No. ORAUT-TKBS-0008-5 Revision No. 03 Effective Date: 08/17/2012 Page 9 of 116

kt kiloton

L liter

LANL Los Alamos National Laboratory

LASL Los Alamos Scientific Laboratory

LLD Lower limit of detection

LLNL Lawrence Livermore National Laboratory

LOS line of sight

LRL Lawrence Radiation Laboratory

L X-ray low-energy X-ray

m meter

MCi megacurie

mCi millicurie

MDA minimum detectable activity or amount

MeV megaelectron-volt, 1 million electron-volts

Mg milligram

Min minute

mL milliliter

mm millimeter

MPBB maximum permissible body burden

MPC maximum permissible concentration

MPCa maximum permissible concentration in air

MPCUa maximum permissible concentration in air for an unknown alpha emitter

mR milliroentgen

mrad millirad

mrem millirem

mS millisiemen

MSA Mine Safety Appliances Company

NBS National Bureau of Standards

nCi nanocurie

NCRP National Council on Radiation Protection and Measurements

NIOSH National Institute for Occupational Safety and Health

NRDS Nuclear Reactor Development Station

NTS Nevada Test Site

NV DOE Nevada Operations Office

OBP overburden plug

ORAU Oak Ridge Associated Universities

Pan Am Pan American Airways

pCi picocurie

POC probability of causation

PPE personnel protection equipment

POC probability of causation

PPG Pacific Proving Grounds

radex radiation exclusion

RadSafe radiological safety

RAMATROL Radioactive Material Control

RAS retrospective air sampler

RCT Radiological Control Technician
Document No. ORAUT-TKBS-0008-5 Revision No. 03 Effective Date: 08/17/2012 Page 10 of 116

REECo Reynolds Electrical & Engineering Company

RSN Raytheon Services Nevada

SEC Special Exposure Cohort

SMS Site Monitoring Services

SNL Sandia National Laboratories

SRDB Ref ID Site Research Database Reference Identification (number)

TBD technical basis document

TLD thermoluminescent dosimeter

TRU transuranic

TTR Tonopah Test Range

U.S.C. United States Code

WBC whole-body count or counter

WEF Waste Examination Facility

WLM working level month

WSI Wackenhut Services, Inc.

yr year

μCi microcurie

μg microgram

μm micrometer

μS microsiemen

§ section or sections
Document No. ORAUT-TKBS-0008-5 Revision No. 03 Effective Date: 08/17/2012 Page 11 of 116

5.1 INTRODUCTION

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

facility encompassed by EEOICPA.

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

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

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

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

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

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

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

determined in accordance with the POC [probability of causation1

] guidelines established under 

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

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

referred to in 42 U.S.C. § 7384n(b) to nuclear weapons work (NIOSH 2010a).

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

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

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

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

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

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

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

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

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

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

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

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

occupationally derived (NIOSH 2010a):

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

conventional structures

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

reasons

 1 The U.S. Department of Labor (DOL) is ultimately responsible under the EEOICPA for determining the POC.
Document No. ORAUT-TKBS-0008-5 Revision No. 03 Effective Date: 08/17/2012 Page 12 of 116

5.1.1 Purpose

The purpose of this document is to describe the Nevada Test Site (NTS) internal dosimetry systems 

and practices. The Oak Ridge Associated Universities (ORAU) Team will use this information as 

needed to evaluate potential internal occupational doses for EEOICPA claims. This document will be 

used in conjunction with other sections of the site profile.

5.1.2 Special Exposure Cohort Petition Information for the Nevada Test Site

The status of Special Exposure Cohort (SEC) petitions for NTS is:

Classes Added to the SEC

• DOE employees or DOE contractor or subcontractor employees who worked at the Nevada 

Test Site from January 27, 1951 through December 31, 1962 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 (Leavitt 2006a).

• 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 included in the SEC (Sebelius 2010).

NIOSH has determined, and the Secretary of Health and Human Services has concurred, that in the 

absence of bioassay results for the worker, internal doses cannot be reconstructed between 1951 and 

1962 inclusive for an Energy Employee (EE). 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. Any bioassay results in the DOE files for NTS workers before 

1963 should be assumed to be valid [HASL-300 (Harley 1976) procedures were used 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.

NIOSH has determined, and the Secretary of Health and Human Services has concurred that 

NIOSH lacks sufficient information that would allow it to adequately estimate internal exposures 

during the period 1963 through 1992 (Sebelius 2010). NIOSH believes that the cessation of nuclear 

testing, coupled with the implementation of the 1993 NTS internal technical basis document 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 technical basis document (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 proposed 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 employed at NTS during the period from 1951 through December 31, 

1992, but who do not qualify for inclusion in the SEC, can be performed using these data as 

appropriate.
Document No. ORAUT-TKBS-0008-5 Revision No. 03 Effective Date: 08/17/2012 Page 13 of 116

5.1.3 Scope

NTS is unique in the DOE Complex because it has not hosted ongoing production of materials as 

other DOE facilities have. NTS was, and is, an outdoor testing and research facility rather than a 

manufacturing or processing site. When NTS began atmospheric testing in 1951, its radiological 

safety programs were assumed to be consistent with programs at Los Alamos National Laboratory 

(LANL; previously known as Los Alamos Scientific Laboratory, LASL) and Lawrence Livermore 

National Laboratory (LLNL; previously known as Lawrence Radiation Laboratory, LRL). For each test 

the laboratory responsible for the device was the responsible organization during the test, providing 

the scientists and technicians associated with the device. NTS personnel provided support to these 

teams.

Reynolds Electrical & Engineering Company (REECo), the NTS general contractor from 1952 to 1995, 

became responsible for onsite radiological safety activities in 1955 (Smith 2003). LANL and the 

military were the primary parties responsible for radiological safety oversight before 1955 (Smith 

2003).

LANL conducted bioassay procedures from 1955 to 1958 (Smith 2003). In 1958, REECo began 

onsite bioassay (Smith 2003). By 1961, REECo was conducting routine bioassay for 3

H, 239Pu, gross 

fission products (GFP), and gamma emitters (Author unknown undated a). Bechtel Nevada (BN), was 

the site contractor from 1995 to 2006, and as such had responsibility for the bioassay program (BN 

1998, 2000; McMahan and Ogurek 2003). In 2001, BN disbanded the onsite analytical laboratory and 

contracted the services to an outside laboratory (Smith 2003). The current contractor is National 

Security Technologies, LLC (DOE 2007).

In 1967, an onsite whole-body counting capability was established (Teasdale 1985). Prior capability 

existed at other locations, but this information is not well documented. Records include lung, chest, 

thyroid, and wound counting results in support of specific tests and projects. The whole-body 

counting capability at NTS was maintained until 1999 when BN decommissioned the system.

Initially NTS adopted the contamination and internal limits published in the National Bureau of 

Standards Handbook 52 issued March 20, 1953 (REECo 1961). Authorized external limits could vary 

by test and job assignment based on the type, purpose, and size of the test. For specific tests the 

information about authorized external limits for certain job titles would be found in the Radiation Safety 

Plan. The Site Research Database (SRDB) includes examples of these plans.

The primary internal dose considerations during the nuclear underground testing era at NTS included 

tritium in some tunnel environments, venting [fission products (primarily iodines and xenon/krypton)], 

and seepage (usually xenon/krypton) released over time (Author unknown ca. 2004). Iodine-131 was 

a radionuclide of concern up to 10 days after a test; 133I and 135I were dose concerns up to 1 day after 

a test (Allen et al. 1993). Xenon, an external dose concern that is discussed in ORAUT (2008), could 

have been present in significant quantities several days after a test.

The information in this TBD characterizes NTS. For some claims the NTS worker was at non-NTS 

locations (e.g., Pacific, Mississippi, Colorado, New Mexico, Alaska, or other Nevada locations, such 

as the Tonopah Test Range, Central Nevada Test Area, and the Project Shoal site near Fallon) during 

testing. Reconstruction of doses for NTS workers who were present at tests at those locations may in 

some cases be developed using the information found in this TBD, if documentation for the test 

identifies REECo as responsible for Radiation Safety.

NTS contractors include BN; Lockheed Martin Nevada Technologies; Johnson Controls Nevada; and 

Wackenhut Services (WSI). Former NTS contractors included Holmes & Narver (H&N), 1956 to 1990; 
Document No. ORAUT-TKBS-0008-5 Revision No. 03 Effective Date: 08/17/2012 Page 14 of 116

Fenix & Scisson of Nevada (FSN), 1963 to 1990; Edgerton, Germeshausen, and Greer Corporation 

(EG&G), 1951 to 1995; REECo, 1953 to 1995; and Raytheon Services Nevada (RSN), 1990 to 1995.

This TBD contains supporting documentation to assist in the evaluation of occupational internal doses 

from the activities performed by the above contractors using the methodology in the Internal Dose 

Reconstruction Implementation Guideline (NIOSH 2002). NIOSH considers the available data and 

methods for performing internal dose reconstruction to be adequate for estimating with sufficient 

accuracy the internal radiation doses at NTS from 1993 to the present.

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

justification, or clarification of the associated information, are presented in Section 5.7.

5.1.4 Historical Monitoring Perspective

At NTS, the primary emphasis was external dosimetry, particularly during atmospheric testing. This 

emphasis was based on animal data showing that the internal dose received during the life span of 

the animal due to inhalation was small in comparison to the external dose. The rule-of-thumb for 

atmospheric testing was that external exposure was the controlling factor. If the external dose was 

controlled, the internal dose would be low. The following paragraphs describe the NTS historical 

perspective for internal dose (Author unknown undated a):

“During the atmospheric testing era at the Nevada Test Site, early animal data showed 

that the internal dose due to inhalation was small in comparison to the external 

exposure. For example, in WT-396, “Biological Injury from Particle Inhalation,” [Smith, 

Boddy, and Goldman 1952] reached the conclusion, that total internal dose due to 

the emission of beta particles was less than one percent of the external dose.”

UCLA [the University of California at Los Angeles] studied the question of internal 

exposure and reported their results in WT-1172, “Evaluation of the Acute Inhalation 

Hazard from Radioactive Fall-out Materials by Analysis of Results from Field 

Operations and Controlled Inhalation Studies in the Laboratory” [Taplin, Meredith, and 

Kade 1958]. The UCLA scientists concluded that ”from consideration of physical 

factors alone (such as strength and type of detonation, particle-size distribution, decay 

rates, meteorological conditions, air-borne radioactivity levels, and percentages of 

radioactivity from 0.1 to 5.0 micron size range), the acute external beta-gamma 

radiation hazard is at least 1000 times greater than that from inhalation.“ The 

rule-of-thumb for atmospheric testing was that the external exposure was the 

controlling factor. If you controlled for the external dose, the internal doses would be 

low.

Tritium does not follow this rule-of-thumb, but tritium exposure was confined to the NTS 

tunnel environments. The Tunnel workers in this environment received quarterly 

bioassay tests. During the underground testing era, there was an active bioassay 

program. Personnel that had a potential for internal exposures were placed in the 

bioassay monitoring program.

Black and Potter (1986) stated that during the early period of nuclear weapons testing, external 

radiation exposure was more hazardous than internal exposure when local fallout (i.e., within a few 

hundred miles of a test) caused both. This was affirmed many times by advisory groups and 

published data. Black and Potter cited examples from published literature to illustrate this point.

Dose reconstructions performed for military personnel on the site during atmospheric testing 

determined that the rules-of-thumb were not valid for individuals in the line of the fallout or present in 
Document No. ORAUT-TKBS-0008-5 Revision No. 03 Effective Date: 08/17/2012 Page 15 of 116

an area where recent or previous tests had been performed and resuspension of residual 

contamination was a possibility. Depending on work activities, internal exposure from resuspension of 

radioactive material can be an issue in some areas years after aboveground testing ended (NRC

2003). ORAUT (2010a) addresses environmental doses and discusses potential intake from 

resuspension. NTS is a large site with much infrastructure. The vast majority of workers had no 

direct contact with radioactive materials (e.g., emplacing devices, retrieval of instrumentation and 

samples after an event, etc.). Environmental levels discussed in ORAUT (2010a) characterize the 

radiological exposure for the majority of the NTS workforce.

5.1.5 Radionuclides of Concern

Technical Basis for Internal Dosimetry at NTS (Allen et al. 1993) reflects radiological protection 

practices from about 1970 to the end of nuclear weapons testing in 1992, and is the best available 

source of internal dosimetry information for the nuclear weapons testing era (Arent and Smith 2004). 

Subsequent technical basis documents by BN (BN 2000; McMahan and Ogurek 2003) cover the 

needs of NTS after 1992 and are not comprehensive on the support of nuclear weapons testing.

NTS operations and facilities had different sets of radionuclides of concern at different times and 

locations on the site. Radionuclides were identified for NTS locations and timeframes, including 

atmospheric, underground, and nuclear reactor/rocket development tests, and legacy contamination. 

Radionuclides of concern for internal dosimetry can be identified by locations, test category, or facility. 

This information is in Attachment A, Section A.4. These radionuclide-of-concern lists are applicable to 

individuals identified with the specific operation or test, AND who had a potential for exposure based 

on their job categories. Dose reconstructors (DRs) can use these lists if a realistic assessment of the 

worker’s dose is necessary to evaluate potential exposure.

Iodine, krypton, xenon, and tritium were important radionuclides following nuclear tests. Krypton and 

xenon were primarily external hazards, whereas 131I and 3

H were internal hazards (Glasstone 1971).

Radionuclides that have resulted in recorded doses above established limits at NTS are 3

H, 131I, 239Pu, 

and 241Am (Arent and Smith 2004).

At NTS, the primary elements of dosimetric concern are plutonium (238Pu, 239Pu, 240Pu, 241Pu, 242Pu), 

uranium (233U, 234U, 235U, 236U, 238U), americium (241Am, 243Am), curium (244Cm), strontium (85Sr, 90Sr, 

90Sr90Y), cesium (137Cs), tritium (3

H), radium (226Ra), and thorium (228Th, 232Th) (Allen et al. 1993). 

Iodine was of concern after full criticality experiments (i.e., atmospheric testing and venting). 

Inhalation is assumed to be the most frequent mode of intake, so the lungs are organs of concern in 

all cases. Cesium and tritium irradiate the whole body relatively uniformly and do not contribute to 

one organ or tissue preferentially. Associated internal dose with atmospheric testing has not been 

developed further with the establishment of the SEC for the Pacific Proving Grounds (PPG) and NTS 

(Leavitt 2006a,b).

5.1.6 Bioassay Program Description

As noted in the Historical Perspective (Author unknown undated b), the dosimetry emphasis was on 

external monitoring. Early bioassay screens (e.g., nasal swabs, respirator swipes, and urine samples) 

were performed in the event that contamination was found or suspected. A positive nasal swab 

initiated the collection of a urine sample. As the bioassay program matured, urine samples were 

collected in a routine, random screening process for workers identified by their supervisors as having 

the potential for internal exposure based on their assigned job tasks. Routine bioassays at NTS 

included quarterly urine samples, annual whole-body count (WBC), and new/termination WBCs (Allen 

et al. 1993). Table 5-1 lists the types of bioassay monitoring and radionuclides discussed in the 

REECo/BN Bioassay Worksheet. These should be considered the routine types of bioassays 

conducted at NTS and can be seen in the case documentation.
Document No. ORAUT-TKBS-0008-5 Revision No. 03 Effective Date: 08/17/2012 Page 16 of 116

Table 5-1. Types of bioassay monitoring and radionuclides.

Monitoring Radionuclides

Urine H-3 (activity),a Pu-238, Pu-239,a gamma emitters,a GFP (gross beta),a Sr-90

Fecal Pu-238, Pu-239

WBC Gamma emittersb

Lung count U-238, Pu-238, Pu-239, Am-241

Thyroid Radioiodines

Nasal swabs Pu-239

Wound counting Pu-238, Pu-239, beta, gamma

Wound swabs H-3, gross alpha, gross beta, gross gamma

a. Routine analysis by REECo starting in 1961. NOTE: REECo started analyzing nasal swabs (Pu-239) and 

urine samples (Pu-239, GFP) in 1958; LANL conducted nasal and urine sampling for Pu-239 in 1955 and 

1956.

b. Routine analysis by Pan American Airways (Pan Am) from 1967 to 1973, and by REECo at NTS in 1973. 

Before 1967, several analyses were done by Helgeson Nuclear Services (HNS).

By instrument survey, air sampling, and surface swipe analysis procedures, health physics personnel 

identified and isolated work sites where personnel might have encountered internal dose hazards. 

Appropriate measures were taken to prevent workers at such sites from inhaling or ingesting 

radioactive contaminants. Urine samples were obtained to monitor the effectiveness of these 

measures. The data resulting from the analysis of such samples were not ordinarily used for 

estimation of internal dose. If internal contaminants were detected in sufficient quantities to indicate 

that dosimetry was possible, special procedures were initiated. If gamma emitters were involved, in 

vivo measurement procedures such as whole body (or specific organ) counting were preferred. If the 

internal contaminant was tritium, it was possible to make an initial estimate of the dose commitment 

from a single urine sample result. Dosimetry of any of the alpha or pure beta emitters was attempted 

on the basis of a carefully collected and analyzed series of urine samples (REECo ca. 1980s).

By 1993, the common in vitro radiobioassay conducted at NTS included the following (Allen et al. 

1993):

• Tritium in urine – 5 mL of distillate from a 30-mL sample were quantified by liquid scintillation 

beta counting (70 minutes).

• Strontium in urine – the strontium precipitated as a carbonate from a 900-mL sample was 

determined by beta counting (100 minutes).

• GFPs in urine – a 250-mL sample was determined by gross beta counting the oxide from a 

basic oxalate (100 minutes).

• Uranium in urine – an aliquot of a 1.5-L sample was converted to fluoride and analyzed 

fluorometrically.

• Plutonium and americium in urine – a 2-L sample processed by the phosphate precipitate 

method, purified by anion exchange, followed by electrodeposition and counted by alpha 

spectroscopy (passivated implanted planar silicon detector system, 1,000 minutes).

• Gamma in urine – 50 mL were counted with a germanium-lithium (GeLi) gamma spectroscopy 

system for determination of gamma-emitting radionuclide activities.

• Gamma in feces – 500 g were counted with a GeLi gamma spectroscopy system for 

determination of gamma-emitting radionuclide activities. Fecal ash samples were analyzed by 

alpha spectroscopy for radium, uranium, plutonium, and americium.
Document No. ORAUT-TKBS-0008-5 Revision No. 03 Effective Date: 08/17/2012 Page 17 of 116

Allen et al. (1993) lists routine bioassay programs for specific operations. General bioassay programs 

for drill-backs, mine-backs, atmospheric testing, and nuclear rocket operations were as follows:

• Drill-backs – Collection of urine samples after each drill-back; gamma spectroscopy for gross 

gamma-emitting radioactivity and for radionuclide identification; annual WBC.

• Mine-backs – Collection of urine samples based on air sampling results; gamma spectroscopy 

and tritium analysis; annual WBC.

• Atmospheric safety tests – Quarterly urine; gamma spectroscopy and analysis for plutonium, 

and americium; annual WBC.

• Atmospheric weapon tests – Quarterly urine; gamma spectroscopy and analysis for GFP, 

plutonium, and americium. Annual WBC; some personnel involved in atmospheric weapons 

tests also received periodic lung counts. Involvement means being part of the reentry team or 

directly involved with the testing activity in the restricted area.

• Nuclear rocket development tests – Quarterly urine; gamma spectroscopy and analysis for 

GFPs, plutonium, and tritium. Annual WBC.

Workers with the potential for an intake [e.g., Miners, Drillers, Radiological Control Technicians 

(RCTs; also called Radiation Safety Monitors] participated in the routine bioassay program. The 

results of the urine sample analysis were used to determine if further bioassay measurements such as 

thyroid, lung, or whole-body counting were indicated. The urine sample results were also used to 

determine if processing of the sample for specific beta/alpha particle emitters or fecal analysis was 

necessary. BN (1998) includes a list of routine urine and fecal analysis frequencies by radionuclide:

• Weekly – 3

H

• Monthly – 131I, 239Pu

• Quarterly – 60Co, 90Sr, 137Cs, 226Ra, 234U, 238Pu (fecal), 239Pu (urine and fecal), 241Am

Nonroutine bioassay types included job-specific and occurrence response. Examples of nonroutine 

bioassays include the following analyses:

• Tritium or gamma scan

• GFP (beta)

• Specific radionuclides

WBCs, lung counts, thyroid counts, and biological sampling were performed as soon as practicable 

after a suspected intake (Allen et al. 1993). Specific examples include:

• Lung counts after a suspected intake of thorium, uranium, or a transuranic (TRU) element

• WBCs for detecting most gamma-emitting fission and activation products

• Thyroid counts for suspected radioiodine uptakes

• Urine bioassay for detection of pure beta emitters such as 89Sr and 90Sr

• Urine and feces sampling and lung and whole-body counting to detect and assess intakes of 

actinides
Document No. ORAUT-TKBS-0008-5 Revision No. 03 Effective Date: 08/17/2012 Page 18 of 116

Attachment A, Section A.4, summarizes facility descriptions and their specific routine bioassay 

programs (Allen et al. 1993). Bioassay samples might have been taken to document that the worker 

was not internally contaminated as well as to confirm a suspected intake (Author unknown undated c).

At present, a large amount of the radiological work at NTS is through projects with a specified (often 

short) duration. Routine bioassay monitoring is usually part of a project. A baseline sample is 

collected as necessary, and postwork samples are collected at the conclusion of the project. If a 

project requiring bioassay monitoring is likely to last longer than 6 months, bioassay sampling on a 

quarterly to semiannual frequency for the duration of the project is recommended. An exception to 

this could be a 3

H project with high exposure potential and a relatively high frequency of required 

monitoring (e.g., weekly to monthly), because typical field monitoring instruments do not detect 3

H 

(McMahan and Ogurek 2003). While this was the bioassay program developed and documented by 

the contractor, internal monitoring was not performed for most workers (see Section 5.1.3). For many 

job categories, internal dose based on the site environmental inventory will provide an overestimate of 

potential internal exposure (i.e., environmental dose can be evaluated as both internal and external).

5.2 IN VITRO MINIMUM DETECTABLE ACTIVITIES AND COUNTING METHODS

Decision Levels and Minimum Detectable Activities

From McMahan and Ogurek (2003), the Decision Level (DL) is calculated as:

1.65 1 3 s b s

b

s

T R T T

DL

KT

  ×× + +     = (5-1)

where:

Rb = background count rate, Cb/Tb; Cb = background counts

Ts = sample count time

Tb = background count time

K = calibration factor in appropriate units such as counts in seconds per unit activity

According to McMahan and Ogurek (2003), the DL equation is modified for unpaired blank and 

sample counting times, and was developed for an alpha probability of a Type I error (false positive) 

equal to 0.05. The DL is applied to an individual sample to determine if the sample count rate is 

different from the count rate of an appropriate blank.

The minimum detectable activity or amount (MDA) is an a priori value used to evaluate the 

laboratory’s ability to detect an analyte in a sample. McMahan and Ogurek (2003) defines the MDA 

as “the smallest amount (activity or mass) of an analyte in a sample that will be detected with a 

probability, beta, of nondetection (Type II error) while accepting a probability, alpha (Type I error), of 

erroneously deciding that a positive (nonzero) quantity of analyte is present in an appropriate blank 

sample. The MDA is computed using the same value of alpha as used for the DL. The MDA depends 

on both alpha and beta. Measurement results are compared to the DL, not the MDA; the MDA is 

used to determine whether a program has adequate detection capability. The MDA will be greater 

than or equal to the DL.” The MDA corresponding to the above DL, with a beta probability of a Type II 

(false negative) error equal to 0.05, is:

3 3.29 1 s b s

b

s

T R T T

MDA

KT

    + ×× +         = (5-2)
Document No. ORAUT-TKBS-0008-5 Revision No. 03 Effective Date: 08/17/2012 Page 19 of 116

In Allen et al. (1993), the MDA equation was reported as:

MDA = [3 + (4.65(Cb)

1/2)] ÷ ETVR (5-3)

where:

Cb = total counts collected in count time (T)

E = counting efficiency

T = count time

V = sample volume

R = fractional chemical recovery

Lower limit of detection (LLD) is defined in Allen et al. (1993) as a value selected above the MDA to 

reduce the probability of reporting false positive results. Detection limit is a general term related to the 

smallest amount of material detectable as a function of the measurement method and instrument 

background.

The tables in Attachment A, Section A.2.1, list the bioassay codes used in the records.

5.2.1 In Vitro Urine and Fecal Analysis

MDA Values for Urinalysis and Fecal Analysis

Table 5-2 lists current values of MDA for in vitro analyses of routine urine and fecal samples 

(McMahan and Ogurek 2003). Because no specific beginning dates were found, the document 

publication date should be used as the effective date. Section 5.2.2 and Attachment A, Table A-7

contain historic MDA values.

5.2.2 In Vitro Methods for Individual Radionuclides

Allen et al. (1993) states that periodic urine samples were collected and analyzed for 238Pu, 239Pu, 

elemental uranium, 234U, 235U, 238U, 241Am, 89Sr, 90Sr, 3

H, and GFPs. In vitro radiobioassays included:

• Tritium in urine by liquid scintillation beta counting

• Strontium in urine by beta counting

• GFPs in urine by beta counting

• Uranium in urine by fluorometric analysis

• Plutonium and americium in urine by alpha spectroscopy

• Radium, uranium, plutonium, and americium in feces by alpha spectroscopy

• Gamma emitters in urine and feces by gamma spectroscopy

Table 5-2. 2003 target MDAs for in vitro bioassay sample analysis.

Parameter and 

analysis method Sample typea and container

Minimum detectable 

amountb Reporting units

Pu-238, Pu-239/240 Urine, 4-L or 500-mL plastic bottlec 0.006c pCi/sample

Feces, 1-L plastic container 0.03 pCi/sample

Am-241 Urine, 4-L or 500-mL plastic bottlec 0.006c pCi/sample

Feces, 1-L plastic container 0.03 pCi/sample

Th-230/232d Urine, 4-L or 500-mL plastic bottle 0.02 pCi/sample

Feces, 1-L plastic container 0.05 pCi/sample

U-234, U-235, U-238 Urine, 4-L or 500-mL plastic bottle 0.04 pCi/sample

Feces, 1-L plastic container 0.04 pCi/sample

Cm-244d Urine, 4-L or 500-mL plastic bottle 0.008 pCi/sample

Ra-226d Urine, 4-L or 500-mL plastic bottle 0.1 pCi/sample

Sr-90 Urine, 4-L or 500-mL plastic bottle 1 pCi/L
Document No. ORAUT-TKBS-0008-5 Revision No. 03 Effective Date: 08/17/2012 Page 20 of 116

Parameter and 

analysis method Sample typea and container

Minimum detectable 

amountb Reporting units

Gamma spectroscopy Urine, 500-mL plastic bottle 100 (Cs-137) pCi/L

H-3 Urine, 500-mL plastic bottle 0.005 μCi/L

Gross alpha/beta Cotton swabs, smears Alpha–10

Beta–100

pCi/sample

Source: McMahan and Ogurek (2003).

a. Sample collection was 24 hours.

b. With the exception of Ra-226, Sr-90, Cs-137, H-3, and gross alpha/beta, upward adjustments for larger sample sizes 

can be allowed.

c. Due to limits of the alpha spectroscopy methodology employed for these parameters, these MDAs, while preferred, 

should be considered goals toward which the contracted laboratory will work to achieve on each set of samples using 

reasonable processing parameters such as analyzing the entire 24-hour void sample, obtaining reasonable recoveries, 

extending count times, and reasonable background levels.

d. Used in calibration activities

The following paragraphs discuss in vitro methods for specific radionuclides. Attachment A, Table A-7 

lists the MDAs for these radionuclides. NOTE: The actual MDA publication date is listed in this 

section whereas the date ranges for the MDAs in Table A-7 are extrapolated to cover missing periods. 

The suggested priority for DRs with regard to MDAs and other detection thresholds is to use (1) the 

limit on dose report if available, (2) Table A-7 values, or (3) an appropriate published value from 

another DOE site or an applicable value referenced in the literature.

5.2.2.1 In Vitro Bioassay for Iodine

In 1961, the laboratory limit of sensitivity for 131I was listed as 10 pCi/sample (Author unknown 

undated a). In 1993, the 131I LLD was listed as 100 pCi/L urine (Allen et al. 1993). BN (1998) lists the 

131I MDA as 99 pCi/L. Iodine was not listed as a routine bioassay in 1993 and the special bioassay 

program consisted of air monitoring. Bioassay records indicate if the air samples were collected using 

a charcoal canister or a filter.

5.2.2.2 In Vitro Bioassay for Americium

The McClelland (1955) method for detecting americium in urine was based on the coprecipitation of 

americium with bismuth phosphate from a nitric acid solution of urine salts at a pH of 1.7. The 

bismuth phosphate was dissolved in 6N HCl and the americium was coprecipitated a second time with 

lanthanum fluoride. The precipitate was slurried onto a stainless-steel plate and counted with a lowbackground proportional alpha counter. Quantities of the order of 2 dpm/24-hr sample or 6 × 10-19 g

of americium could be determined by this method. The tolerance for americium in urine used at LASL 

was 7 dpm/24-hr sample. Samples were rechecked if the count was 2 dpm/24-hr sample or higher 

(McClelland 1955).

In 1958, LASL used the same procedure for americium in urine and stated that thorium, plutonium, 

curium, actinium, and neptunium were carried through this determination. Quantities on the order of 

0.5 dpm of americium were detected by this method (McClelland 1958; Dummer 1958). The LASL 

MDA value for americium was 9 × 10-1 pCi/24-hr sample (ORAUT 2009).

REECo documentation (REECo 1969–1987a) discussed an americium method with steps for 

separation by precipitation/oxidation, purification by anion exchange column, electrodeposition of 

241Am243Am on a stainless-steel disc, and detection by alpha spectrometry during the period from 

1981 through 1983. The procedure stated that 241Am was analyzed using 243Am as a tracer and 

separated from sample matrices by precipitations such as Y(OH)3, YF3, and Y2(C2O4)3. Americium241 was separated from rare earth elements by oxidation. Final purification was accomplished using 

an anion exchange column. Americium-241/243 was electrodeposited on a stainless-steel disc and 

the activity determined by alpha spectrometry. From 1982 to 1987, the detection limit was 2 × 10-11
Document No. ORAUT-TKBS-0008-5 Revision No. 03 Effective Date: 08/17/2012 Page 21 of 116

μCi/mL for 241Am in urine (REECo 1968–1987a). The routine bioassay program was not designed for 

the control of 243Am. NTS did not specifically analyze 243Am in urine. If an employee has a history of 

work at the low-level waste sites, Waste Examination Facility (WEF), or the radiochemistry laboratory, 

DRs should evaluate 243Am intake through WBC and/or lung counting results if an intake is suspected. 

Allen et al. (1993) lists no urine MDA for 243Am; it does list MDA values for WBCs and lung counts for 

243Am.

In 1993, the LLD for 241Am was listed as 0.03 pCi/L for urine and 0.03 pCi/g for feces, with a note that 

americium cannot be chemically differentiated from californium and curium (Allen et al. 1993). The 

routine bioassay was a quarterly urine sample and the special bioassay consisted of an additional 

urine sample and a lung count. Allen et al. (1993) stated that americium in urine was detected with an 

alpha counter (assumed to be an alpha spectrometer) with a “typical” MDA of 0.05 pCi/L. However, 

the documentation does not provide a definition for “typical.” BN (1998) stated the listed MDA values 

are typical of MDAs achieved at NTS. Individual MDAs, which were calculated per sample, varied 

depending on the sample matrix for in vitro measurements or person for in vivo measurements, the 

length of the count, and the background of the specific instrument used for counting. Attachment A, 

Table A-7 lists limits of detection for urine and fecal analysis. Table 5-2 lists the current MDAs.

5.2.2.3 In Vitro Bioassay for Plutonium

LASL used the following procedure for plutonium in urine in 1954.

The urine sample was ashed with nitric acid; plutonium was coprecipitated with bismuth phosphate, 

dissolved in hydrochloric acid, and then coprecipitated with lanthanum fluoride. The lanthanum 

fluoride precipitate was slurried on a stainless-steel plate and counted for alpha activity with a lowbackground proportional counter. Quantities of the order of 2 dpm or 2 × 10-11 g of plutonium could be 

determined by this method. The tolerance for plutonium in urine in the LASL “Official Monitoring 

Handbook” was 7 dpm/24-hr sample. Samples were rechecked if the results were 2 dpm/24-hr 

sample or higher (McClelland 1955). A tolerance level of 1 dpm/24-hr sample was used for plutonium 

in urine for Project 56, a plutonium dispersal experiment conducted in Area 11 (AEC 1956).

The 1958 LASL procedure for plutonium in urine was as follows. The urine sample was wet-ashed 

with nitric acid. After reduction to plutonium (III) with hydroxylamine hydrochloride, the plutonium was 

coprecipitated from acid solution with lanthanum fluoride, oxidized to plutonium (IV) with sodium 

nitrite, and separated from the residual urine salts and the lanthanum by extraction with thenoyl

trifluoroacetone. The extracted plutonium was oxidized to plutonium (VI) by hypochlorite and 

electrodeposited on a stainless-steel disc with a 38.5-mm2 plating area to give a thin, uniformly 

distributed film of plutonium. The discs were exposed to nuclear-track alpha plates and the number of 

tracks from the alpha particles was counted visually with a microscope. The electroplated discs were 

counted electronically if the results were needed immediately. The procedure separated plutonium 

from uranium, actinium and its progeny americium, curium, and thorium. Plutonium recoveries of 85 

+5% were routinely obtained. The detection limit at the 99% confidence level was approximately 

0.05 dpm/sample (3.6 × 10-13 g of plutonium). Samples showing 0.2 dpm/sample or more 

(1.5 × 10-12 g of plutonium) were rescheduled (McClelland 1958; Dummer 1958). The LASL MDA 

values for plutonium (total alpha) were 0.4 pCi/24-hr sample (biphosphate/alpha counting) until 

January 1957 and 0.03 pCi/24-hr sample (aluminum nitrate/neutron track analysis) after January 1957 

(ORAUT 2009).

In 1961, urinalysis for plutonium included sample preparation, anion exchange, electrodeposition, 

autoradiograph, and alpha track counting (REECo 1961). The laboratory capability (limit of sensitivity) 

was 0.005 dpm/sample (Author unknown undated a). Further documentation of the limit of sensitivity 

for 239Pu is described in Geiger and Whittaker (1961). In the evaluation of the procedure, 61 urine 

samples spiked with 0.027 pCi of 239Pu; 79 blank urine samples were analyzed. The average 
Document No. ORAUT-TKBS-0008-5 Revision No. 03 Effective Date: 08/17/2012 Page 22 of 116

recovery was 84 +18% (90% CL). The average blank was equivalent to 0.001 pCi of 239Pu. The 

sensitivity was arbitrarily stated as 0.005 pCi, which was five times the blank. An aliquot of 250 mL 

was used in the procedure; therefore, 0.02 pCi of 239Pu in a 24-hour sample could be easily detected. 

This is approximately 20% of the investigation limit and represents less than 5% of a maximum 

permissible body burden (MPBB) of 239Pu. The exposure time for this method was listed as 10,000 

minutes (Geiger and Whittaker 1961).

In 1968, alpha spectrometry was used to count 239Pu in urine samples (REECo 1968–1987b) and the 

permissible bone burden for 239Pu was listed as 0.04 μCi (Author unknown undated a). From 1982 to 

1987, the 239Pu MDA was 5 × 10-11 μCi/mL and the 238Pu MDA was 2 × 10-10 μCi/mL (REECo 1968–

1987a).

Allen et al. (1993) stated that NTS used alpha spectrometry, which could not differentiate between 

239Pu and 240Pu, to analyze plutonium urine and fecal samples. Urine samples were analyzed by 

gamma spectroscopy, which could differentiate between the two radionuclides. Urine alpha 

spectroscopy had an MDA of 0.01 pCi/L, and the fecal sample MDA was 0.004 pCi/g. The gamma 

spectroscopy method had an MDA of 50 pCi/L for 239Pu. A “typical” MDA for a 1,500-mL urine sample 

counted for 1,000 minutes was 0.028 pCi/L. The 238

Pu LLD was 0.01 pCi/L urine. Routine bioassay 

consisted of a quarterly urinalysis, and the special bioassay consisted of a combination of urine and 

fecal analysis plus a lung count. Attachment A, Table A-7 lists limits of detection for urine and fecal 

analysis. Table 5-2 lists current MDAs for plutonium radionuclides.

Plutonium at NTS consists of a mix of 238Pu, 239Pu, 240Pu, and 241Pu. In addition, 241Am would be 

present from the decay of 241Pu. Therefore, when evaluating missed or confirmed doses from 

plutonium, DRs should include the other radionuclides associated with 241Pu. In the 1980s and early 

1990s, extensive surveys were conducted at NTS to determine the inventory of radionuclides in the 

soils as a result of atmospheric testing that occurred before 1963. ORAUT (2010a) summarizes the 

results of these surveys. These data can be used to estimate the quantities of 241Am and 238Pu in 

relation to 239+240Pu. A comparison of the total soil inventories over the 19 area surveys indicates that 

the ratio of 241Am to 239Pu is 0.59 and the ratio of 238Pu to 239Pu is 0.91 (see Table 4-6 of ORAUT 

2010a). Therefore, when estimating total radionuclide intakes associated with intakes of 239Pu, DRs 

should use these ratios to determine the accompanying intakes of 238Pu and 241Am. DRs should 

estimate these intakes when calculating potential missed dose and dose from confirmed intakes of 

239Pu when positive bioassay data for 238Pu and 241Am are not available.

5.2.2.4 In Vitro Bioassay for Tritium

The 1954 LASL procedure for tritium in urine was as follows. The sample was prepared for counting 

in a vacuum line. Urine was dropped onto metallic calcium, and hydrogen and tritium were evolved. 

The gas flowing into the evacuated system was passed through liquid nitrogen-cooled traps to remove 

unreacted water and condensable gases. The gas was allowed to flow into a tube similar to a GeigerMuller tube until a pressure of 15 cm of mercury was attained. Ethylene and argon were added to 

give a total pressure of 22 cm of mercury. The beta activity was counted with a scaling circuit having 

an input sensitivity of 0.25 volt. A tube similarly filled with inert hydrogen was counted simultaneously 

to determine the environmental background. The background count was subtracted from the sample 

count to obtain the true sample count. The method had an efficiency of approximately 40% and a 

precision of +5% in the range of 1 to 250 μCi/L of tritium. Samples with higher concentrations could 

be determined with appropriate dilutions. The tolerance for tritium in urine at LASL was 250 μCi/L. 

Ten days was used as the biological half-life of tritium. This half-life could be decreased by increasing 

the fluid intake of the worker (McClelland 1955). The LASL MDA value for tritium was 1 μCi/L 

(ORAUT 2009). In 1958, the procedure was the same except the tolerance for tritium in urine used 

was 85 μCi/L (McClelland 1958).
Document No. ORAUT-TKBS-0008-5 Revision No. 03 Effective Date: 08/17/2012 Page 23 of 116

Records from 1966 to 1968 show that tritium analysis was conducted with a liquid scintillation 

spectrometer that had an efficiency of 18% and an average background of 16 cpm (Author unknown 

undated a). In 1971, the urine sample tritium “alert level” was listed as 1 × 10-3 μCi/cm3 (Author 

unknown undated a). The detection limit for 3

H in urine was listed as 1 × 10-6 μCi/mL from 1982 to 

1987 (REECo 1968–1987a).

Tritium monitoring started in 1958 with an MDA of 5 μCi/L used for urine samples (Arent and Smith 

2004). Allen et al. (1993) reported an MDA for tritium of 300 pCi/L using liquid scintillation counting to 

detect the weak beta emitted. Routine bioassay is a quarterly urine sample and the special bioassay 

is an additional urine sample. A “typical” MDA value of 470 pCi/L is listed for a 70-minute count. 

Table 5-2 lists the current MDAs. Attachment A, Table A-7 lists limits of detection for urine and fecal 

analysis. On NTS record forms, tritium monitoring can be called "ACTIVITY” or “ACT," 

“EVERGREEN,” "MINT," or "T".

Tritium was reported separately in the dosimetry records. DRs should assign a tritium missed dose to 

the covered employee for years when the employee was monitored for tritium exposure and no tritium 

dose was reported or when the reported tritium dose was less than the calculated annual potential 

missed dose based on the MDAs listed in Table A-7. After 1977, the tritium bioassay MDA is such 

that if all of the monitoring results are below the MDA, the dose is less than 0.001 rem/yr and the dose 

does not need to be in the dose reconstruction. Individuals assigned to areas with the potential for 

tritium exposure were on the routine tritium bioassay program. In addition, a select number of WSI 

security personnel were on the routine tritium bioassay because they could have been assigned to 

any location on NTS (Rollins 2004). The job title (ORAUT 2008), incident summary (Section A.4), and 

facility radionuclides of dose concern (Section A.2) discussions could assist the DR in determining the 

assignment of a tritium dose.

5.2.2.5 In Vitro Bioassay for Uranium

Environmental concentrations of uranium are highly dependent on geographic location (MJW 2003). 

The activity of uranium in urine samples will vary a great deal between individuals who work in 

different areas and, in a given area, will vary primarily as a function of the individual’s primary water 

source. Well water typically contains more uranium than water from public supplies. In addition, an 

individual’s level of uranium excreted can vary significantly from day to day. Because the sources of 

uranium on the site are typically depleted uranium (DU) or highly enriched uranium (HEU), the ratio of 

238U to 234U can be used as an indicator. DU has a 238U:234U ratio range of about 3 to 10; HEU has a 

ratio of about 0.1 or smaller. Natural uranium in urine has a ratio of about 1, with 234U activity often 

somewhat larger than that of 238U (as much as a factor of 2 or 3 is not unusual). The DR should note 

that these ratios apply to solid-state material under laboratory conditions. In the field, water interferes 

with the chemistry and the natural 238U:234U ratios can vary because, when uranium enters ground 

water, there might be differences in the leaching rate between the two radionuclides.

The exception to DU and HEU being the source of uranium for exposures is at the WEF. Uranium at 

the WEF is likely to be the 233U radionuclide, which can be present in some of the TRU waste being 

characterized at the facility. The alpha particle detected for identifying 233U is similar in energy to the 

234U alpha particle and alpha spectrometry is not capable of resolving the alpha particles from the two 

radionuclides. Therefore, 233U detected in a sample is reported as 234U, and a sample from personnel 

working at the WEF with an unexpectedly low 238U:234U ratio (e.g., less than 0.3) potentially indicates 

an occupational intake.

In the case of positive 235U results without a 234U positive result in the sample, the positive 235U result 

is considered to be a false positive because the 234U is always a few times greater in activity 

concentration than the 235U throughout the range of enrichment from DU to HEU. The only possible 

exception to this is if the uranium has been enriched above 6% 235U weight percent from atomic vapor 
Document No. ORAUT-TKBS-0008-5 Revision No. 03 Effective Date: 08/17/2012 Page 24 of 116

laser isotope separation method (Rich et. al. 1988, as cited in McMahan and Ogurek 2003). This type 

of enriched uranium currently is not known to exist at NTS.

The LASL 1954 fluorophotometric method for uranium in urine was based on the intense yellow-green 

fluorescence (the principal line of which is reported to be at 555 μm) produced by traces of uranium 

fused in sodium fluoride. It was sensitive to concentrations of uranium from 10-5 to 5 × 10-10 g per 

0.25 g of sodium fluoride, with a precision of +10%. The tolerance for normal uranium in urine at 

LASL was 100 μg/L (McClelland 1955). In 1958, the same method was sensitive to concentrations of 

uranium from 10-10 to 5 × 10-11 g per 0.25 g of sodium fluoride, with a precision of +10% (McClelland 

1958).

An ion exchange method for uranium alpha activity in urine was published by LASL in 1958. The 

uranium was concentrated from urine by coprecipitation with alkaline earth phosphates. The 

precipitate was dissolved in 8N hydrochloric acid and the complex uranium chloride anion was 

separated by passing the solutions through an anion exchange column. The uranium was eluted with 

1N hydrochloric acid, which was evaporated and taken up in nitric acid, and plated directly on 

stainless-steel counting discs. The alpha activity was determined by counting in a low-background 

proportional counter. The recovery was approximately 90% in the range from 25 to 50 dpm/L. A set 

of nine analyses could be completed in 3 days with the equipment available at LASL. Rechecks were 

requested if the sample showed more than 50 dpm/L (McClelland 1958).

LASL also had an extraction method for uranium alpha activity for 1958. The urine was ashed with 

nitric acid and the salts were dissolved and made approximately 1N with nitric acid. The uranium was 

extracted from the acid solution of the salts with di-n-butyl ortho-phosphoric acid in carbon 

tetrachloride. The phosphoric acid was evaporated on platinum plates, fused, and alpha-counted with 

a low-background proportional counter. The method had an accuracy of 84 +14% with 1 to 10 dpm/L 

of enriched uranium. At higher concentrations the recovery approximated 100%. Approximately 25 

analyses could be completed in 1 day by one worker with the equipment at LASL (McClelland 1958).

In 1961, REECo laboratory operations published a sensitivity limit for uranium of 0.03 dpm/sample 

(REECo 1961). HEU analysis involved coprecipitation of the uranium with alkaline earth phosphates 

from an ammonium hydroxide solution, filtration, destruction of organic material by wet-ashing, 

purification by extraction and anion exchange, electrodeposition, and alpha counting with a gas flow 

proportional detector (Geiger and Whittaker 1961).

Methods for total uranium by fluorometry and alpha spectrometry and for 235U by anion exchange, 

purification, electrodeposition, and alpha spectrometry were listed in REECo documents published 

from 1981 to 1983 (REECo 1968–1987a).

As described in Allen et al. (1993), 235U and 238U were measured in urine by alpha spectroscopy 

(MDAs of 0.01 and 0.02 pCi/L, respectively). “Typical” MDAs were listed as 0.1 pCi/L. These 

radionuclides were also measured by fecal sample analysis (MDA of 0.008 pCi/g). The routine 

bioassay was a quarterly urine sample and the special bioassay was an additional 24-hour urine 

sample and a prompt lung count. Elemental uranium was determined by a fluorometer, and the MDA 

was listed as 5 μg/L for chemical analysis with a “method” MDA of 0.2 pCi/L. Attachment A, Table A-7 

lists historical detection thresholds for various periods. Table 5-2 lists current MDAs for uranium 

isotopes.

5.2.2.6 In Vitro Analysis for Strontium

Strontium (90Sr) is present in the environment as a result of atmospheric weapons testing. There has 

been no strontium background study at NTS. Excretion due to environmental exposure is likely to be 

small (<MDA), but is highly variable due to varying diets, and might be as large as the current MDA. 
Document No. ORAUT-TKBS-0008-5 Revision No. 03 Effective Date: 08/17/2012 Page 25 of 116

However, because the environmental component is not established, any detectable radiostrontium 

should be assumed to result from an occupational intake.

In 1961, REECo laboratory operations published a sensitivity limit for 90Sr in urine of 25 pCi/sample 

(detected with an accuracy of +10% at a 90% confidence level, counter background of 17 cpm, and a 

60-minute counting period) (Author unknown undated a). The analysis involved the following steps: 

sample preparation, solvent extraction, time delay for 90Y buildup, solvent extraction of 90Y, and beta 

count (REECo 1961). Note that this method did not account for any 89Sr present at intake.

In 1993, gas-flow proportional beta counting was used to quantify 90Sr in urine at NTS. The method 

had an MDA of 0.8 pCi/L, but did not differentiate between 89Sr and 90Sr. Strontium-90 decays to 

produce 90Y, which is a beta emitter. Strontium was extracted from 90Y and other contaminants. The 

first count included betas from 89Sr plus 90Sr. Then the 90Y was allowed to grow in for 2 weeks and 

chemically separated and counted. The second count included only 90Y. Strontium-90 was calculated 

from the 90Y value, and then subtracted from the first count to obtain the 89Sr value.

Allen et al. (1993) stated that “typical” MDAs for a 1,000-mL sample counted for 100 minutes were 

around 0.8 pCi/L. The routine bioassay for 90Sr was a quarterly urine sample, and the special 

bioassay was a prompt urine sample. Attachment A, Table A-7 lists limits of detection for urine and 

fecal analysis. Table 5-2 lists the current MDA.

5.2.2.7 In Vitro Analysis for Thorium

The McClelland (1955) method for 230Th in urine was as follows: The urine sample was ashed with 

nitric acid; the thorium was coprecipitated with bismuth phosphate, dissolved in hydrochloric acid, and 

then coprecipitated with lanthanum fluoride. The lanthanum fluoride precipitate was slurried on a 

stainless-steel plate and counted for alpha activity with a low-background proportional counter. 

Natural thorium, plutonium, some americium, curium, actinium, and neptunium were carried through 

this determination. Quantities of the order of 0.5 dpm of 230Th could be determined with this method. 

In 1993, the 230Th fecal MDA was 0.01 pCi/g (Allen et al. 1993). Attachment A, Table A-7 lists limits of 

detection for urine and fecal analysis. Table 5-2 lists the current MDA.

5.2.2.8 In Vitro Analysis for Radium

Radium in urine was coprecipitated with barium as the sulfate. Polonium was removed by deposition 

on silver foil. The precipitate was slurried onto a stainless-steel plate and the alpha activity counted 

using a methane flow proportional counter. Uranium in amounts usually found in urine did not 

interfere. The recovery was approximately 96-97%. Approximately 10-13 g (0.88 dpm) of radium could 

be detected with this method if the sample was allowed to come to full equilibrium. It has been 

estimated that about 0.01% of the radium in the body [or about 10-11 g (22 dpm) at tolerance] was 

excreted daily in urine 6 months after exposure and that 0.0005% [or 5 × 10-13 g (1 dpm) at tolerance] 

was eliminated daily by long-standing chronic radium poisoning cases. Normal radium excretion of 

unexposed humans is on the order of 0.2 to 0.4 dpm/day at radioactive equilibrium. A 24-hr sample 

was usually collected for radium analysis (McClelland 1958). Allen et al. (1993) states that the 

radiobioassay of urine for 226Ra had an MDA of 300 pCi/L, and the fecal sample MDA was from 0.04 

to 0.4 pCi/g. Attachment A, Table A-7 lists limits of detection for urine and fecal analysis. Table 5-2 

lists the current MDA.

5.2.2.9 In Vitro Gross Fission Product Analysis

For interpreting results from the fission product urinalysis, strontium, barium, europium, zirconium, 

and niobium radionuclides concentrate primarily in the bone, with 90Sr providing the largest dose rate. 

Cerium, lanthanum, and promethium concentrate primarily in the liver, with some concentration in the 
Document No. ORAUT-TKBS-0008-5 Revision No. 03 Effective Date: 08/17/2012 Page 26 of 116

bone, with 144Ce providing the largest dose rate. Cesium and ruthenium are assumed to be uniformly 

distributed in the whole body.

The McClelland (1958) method for gross beta activity in urine was as follows. The nuclides of 90Sr- 90Y, 140Ba-La, 144Ce-Pr, 89Sr, and GFPs were determined as alkaline phosphate precipitates and 

counted directly. The nuclide usually could be identified from the exposure history and decay 

characteristics. The recovery was 80 +5%. If indicated, a gross gamma count was performed on 500 

mL of urine using liquid scintillation techniques. McClelland (1958) states that if the gross beta (90Sr- 90Y, 140Ba-La) result is higher than 200 dpm/L, exposure should be suspected and investigated. The 

LASL MDA value for fission products was 50 to 100 dpm/L (ORAUT 2009).

In 1961, REECo analyzed urine samples for GFP before performing specific analyses (Geiger and 

Whittaker 1961). Strontium, barium, lanthanide rare earths, and other fission products were 

coprecipitated with alkaline earth phosphate from an alkaline solution. The precipitate was assayed 

without further chemical treatment for total beta emitters to obviate, in many cases, detailed 

radiochemical determination. If insignificant quantities of radioactive material were detected, no 

additional analyses were performed. Analyses for specific radionuclides could be requested if the 

gross beta activity exceeded the following control limits:

• Exposure to fission product mixture less than 3 months old – control limit was 1.0 pCi/mL.

• Exposure to fission product mixture greater than 3 months or age unknown – control limit was 

0.1 pCi/mL.

• Anything above 20% of the control limit should have been recorded on the person’s bioassay 

data card. When this occurred, the DR would find a note on the bioassay record indicating 

what fission products were identified during the follow-up analysis.

REECo (1961) stated that this procedure (i.e., coprecipitating fission products on alkaline earth 

phosphates from an alkaline solution and counting the sample package for gross beta activity) did not 

recover cesium, but exposure to cesium without exposure to other fission products was not probable 

in most NTS areas. The GFP procedure provided a screening technique that resulted in considerable 

savings of time and money, according to REECo records.

GFP analysis was generally reserved for quarterly urine samples. From 1970 to 1979, the procedure 

was designed to provide a urine screening or monitoring method by reclaiming radionuclides of 

concern such as 89Sr, 90Sr90Y, 95Zr-Nb, 140Ba-La, and 141Ce. Radionuclides not reclaimed included 

137Cs and the iodines. From 1981 to 1987, the quarterly urine analysis was a screening method for 

measuring gross beta concentrations of radionuclides (i.e., presence or absence of beta-emitting 

radionuclides) after the chemical removal of potassium. Throughout the period, samples were 

forwarded to the laboratory for processing and planchets containing GFP activity were submitted for 

beta measurements (REECo 1970–1987). On rare occasions, strontium was analyzed separately. In 

addition, the gamma energy peak for 137Cs was analyzed in gamma spectroscopy for some samples 

on an occasional basis (DeMarre 2005). From 1982 to 1987, the detection limit for GFP (beta) in 

urine was 1 × 10-10 μCi/mL (REECo 1968–1987a).

Allen et al. (1993) states that the term gross fission product was used rather than gross beta. This 

analysis actually was used for all beta-emitting radionuclides except the alkali group (e.g., 134Cs, 

137Cs, and 40K). The intent was to eliminate naturally occurring 40K from the sample because the 

concentration of potassium in urine varies widely. Detection for beta activity was done in a gas-flow 

proportional counter, and the MDA was listed as 3.0 pCi/L (250-mL sample counted for 100 minutes). 

Attachment A, Table A-7 lists limits of detection for urine and fecal analysis.
Document No. ORAUT-TKBS-0008-5 Revision No. 03 Effective Date: 08/17/2012 Page 27 of 116

Cesium-137 is a product of fallout from weapons testing. As with 90Sr, background levels will 

generally result in small doses. Cesium sampling is performed quarterly. Dose Reconstructors will 

need to consider 137Cs as occupational exposure in the absence of evidence in the worker’s record 

indicating the 137Cs was from another source. Cesium-137 is a radionuclide of concern at NTS 

because of the nature of the work performed and is part of the environmental source term. It is 

appropriate to address 137Cs levels by evaluating the intake using the Integrated Modules for Bioassay 

Analysis (IMBA) computer program

DRs should select fresh or aged fission products based on what is most favorable to claimants for the 

cancer diagnosed. The bioassay records do not indicate fresh or aged. Attachment A, Table A-30 

identifies common radionuclides that are a significant part of fission products up to 1 year old. DRs 

should check the allowable limits for fission products identified as a dose concern. They should 

review the records provided by DOE for additional data, such as access logs and RCT logbooks, and 

review test information to determine if they need to consider any fission products (Table A-30) unique 

to the test based on the cancer location. They might need to request additional records to determine 

aged or fresh fission products.

NOTE: If the worker had a WBC during any year and had a detectable fission product urinalysis 

result, determine intakes of 137Cs, 106Ru, or other gamma-emitting fission products from the WBC. 

Whole-body counting began at offsite locations in 1964 (Church 1967). Onsite capability started in 

1967 (Teasdale 1985). See Section 5.3 for additional in vivo information and references.

5.2.2.10 In Vitro Analysis for Gamma Emitters

In 1961, certain samples were examined with a gamma spectrometer to identify gamma-emitting 

components (REECo 1961). Quantitative estimates were made in some cases. Special chemical 

separations schemes were devised based on the results of the initial pulse height analysis. Almost 

any gamma component reported could be identified by a combination of pulse height analysis, 

chemical purification, and beta measurements. From 1982 to 1987, the detection limit for gamma 

emitters in urine was 5 × 10-8 μCi/mL (REECo 1968–1987a).

Specific analysis for 137Cs in urine in 1961 involved sample preparation, cesium-phosphotungstate 

precipitation, and a specific gamma count (REECo 1961). NTS used gamma pulse height analyses to 

evaluate samples unless there was interference from higher energy gamma emitters, in which case 

NTS separated cesium as cesium-phosphotungstate followed by pulse height analysis. The 

sensitivity was listed as about 10 pCi per sample (REECo 1961).

From 1968 to 1982, gross gamma analyses were conducted for urine samples. Samples were 

counted in a 9-in. by 9-in. sodium-iodide (NaI) well crystal. If counting rates exceeded an established 

level, the samples were placed in one of the 512-channel analyzer systems and a gamma spectrum 

was obtained. Samples counted in the well crystal had a crude spectrum available from a 

multichannel analyzer that accumulated data while the gross count rate was measured. While 

analysts could identify some radionuclides directly from this spectrum, summation effects in the well 

crystal severely restricted resolution of the more complex spectra; these samples had to be analyzed 

with the 512-channel system. The gross gamma counters were calibrated with a 137Cs standard. The 

equivalent branching ratio was used in the conversion factor (disintegrations per count) and this was 

applied to the counting data regardless of the nuclides present. Gross gamma data from the well 

crystal were discriminated to represent the integral count rate from 100 keV to 4 MeV (from 1968 to 

1979). The high-energy discrimination was necessary to limit the increase in count rate above 4 MeV 

caused by high-energy cosmic particles. In 1981 and 1982, gross gamma data were discriminated to 

represent the integral count rate from 100 keV to 2 MeV. During this period, the pulse height 

analyzers used to obtain gamma spectra were calibrated as 1 keV/channel with a zero energy 

threshold. The energy range of 50 keV to 2 MeV included emissions of most of the nuclides 
Document No. ORAUT-TKBS-0008-5 Revision No. 03 Effective Date: 08/17/2012 Page 28 of 116

encountered. Beginning in 1983, a GeLi detector system or intrinsic detector system was used to 

perform gamma spectra analyses for urine samples. The pulse height analyzers used to obtain 

gamma spectra were calibrated at 1 keV/channel with a zero energy threshold. The energy range of 

50 keV to 2 MeV included emissions of most of the nuclides encountered (REECo 1968–1987b).

REECo procedures stated that, if gamma emitters were involved, in vivo measurement procedures 

such as whole-body (or specific organ) counting were preferred (REECo ca. 1980s). For routine 

urinalysis, the gamma scan was used as a screening tool. There was no additional sample 

preparation. If the scan result was less than or equal to 5,000 cpm, a spectral analysis via gamma 

spectroscopy occurred. If the scan result was less than 5,000 cpm, the analysis reported a 137Cs 

equivalent concentration. Each positive radioiodine result was noted and compared to a curve 

relating alert level activity concentrations to the time elapsed between exposure and urine sample 

void time. A result at or near the alert level indicated that thyroid counting of the individual was to be 

considered. DRs might need to evaluate gamma results above the detection limit based on the 

radionuclide of concern identified with the work area or activities performed and should assign such 

results to the radionuclide providing the most dose to the internal organ associated with the cancer 

type. If the DR cannot determine the work area or activities performed, assume the radionuclide 

providing the most dose to the internal organ associated with the cancer type.

In 1993, all urine samples were counted by gamma spectroscopy to determine the concentration of 

gamma-emitting radionuclides in the energy range from 40 to 2,000 keV (Allen et al. 1993). There 

was no special sample treatment except transfer to a 500-mL bottle. Counting was done in a fixed 

position with any of several coaxial high-purity germanium detectors and multichannel analyzer 

gamma spectroscopy systems. The MDAs varied according to the analyte and the specific detector 

system used. The specific MDA for 137Cs in urine was listed as 100 pCi/L (Allen et al. 1993). 

Attachment A, Table A-7, lists limits of detection for urine and fecal analysis. Table 5-2 lists the 

current MDA value.

5.2.3 Sample Collection and Correcting for Urinalysis Volume

The 1961 REECo procedures stated that urine samples were usually received in a kit of four bottles 

(Geiger and Whittaker 1961). When the solution in the four bottles was combined and mixed 

thoroughly, the sample represented an “equivalent 24-hour sample.” The purpose of this procedure 

was to provide a uniform method for obtaining homogenous aliquots for the determination of 

plutonium, uranium, and fission products. The collection start and stop dates were recorded. When 

the interval was less than 1 day, the collection start and stop times were recorded. If the interval was 

more than 1 day, total hours were recorded. The quantity collected was also recorded.

Hartzell (1962) listed the quantities of urine samples required for routine laboratory analysis. For 

activity, a single void of at least 25 to 50 mL was required. For GFP, strontium, uranium, and 

plutonium, a 24-hour sample of at least 250 mL was required. To collect the 24-hour sample, a kit of 

three bottles (1,000 mL total) was issued from Building CP-2 (Area 6).

REECo procedures discussed how urine samples were received and processed from 1970 to 1987 

(REECo 1970–1987). Those and later site procedures contained specific instructions and minimum 

samples sizes for the following radionuclides (REECo 1969–1987; Allen et al. 1993; BN 1998).

• GFP: 250 mL (1969–1993)

• Plutonium-239 (urine)

o 1 L (February 1969–June 1985)
Document No. ORAUT-TKBS-0008-5 Revision No. 03 Effective Date: 08/17/2012 Page 29 of 116

o 900 mL (July 1985–July 1987)

o 1.5-L (1990–1991) optimum sample size

o 2 L (1993)

• Pu-239 (fecal): 1 g ash (1993–1998); also analyzed for radium, uranium, and americium 

(1993)

• Americium: 2 L (1993)

• Uranium: 1.5 L (1993)

• Strontium-89/90: 900 mL (1969–1993)

• Tritium: 20 mL (1990–1991); 5 mL of distillate from a 30-mL sample (1993)

• Gross alpha/beta: 250 mL (1990–1991)

• Gamma (urine): 500 mL desired, 50-mL minimum volume (1985–1993)

• Gamma (fecal): 500 g (1993)

In 1975, the employee’s last name, initials, and social security number were recorded with each 

bioassay sample submitted. By 1985, the person’s full first name and middle initial were required. 

Due to security concerns, only the event or program code associated with the project was entered on 

the laboratory analysis request form; the actual drill hole or tunnel number was not recorded. This 

could make it more difficult for DRs to associate an employee with a specific test. By 1990, the 

sample receiving/control and tracking procedures focused more on chain-of-custody forms and 

computer data entry functions.

As discussed in Allen et al. (1993), types of indirect bioassay samples include total 24-hour urine 

collection (including collection at work) and spot urine samples (single void). In general, urine data 

were normalized to total 24-hour excretion. If a 24-hour sample had a volume of less than 500 mL, 

NTS did not analyze it because this volume is too small to represent a true 24-hour collection. In such 

a case, the worker was required to provide another sample. Provided the sample was collected 

properly, a total or simulated 24-hour urine sample result was used without further normalization. A 

proper 12-hour result was normalized by doubling the result. BN (1998) updated the discussion as 

follows:

For most radionuclides, urine samples representative of excretion over a 24-hour 

period should be collected for quantitative intake and dose assessment. The 

concentration of materials in urine varies throughout the day, with the first void of the 

morning generally being the most concentrated. If an aliquot from a 24-hour sample is 

used for analysis, it is important that the sample be thoroughly mixed before removing 

the aliquot. For hard-to-detect nuclides, such as Pu, the analysis of the entire 24-hour 

sample is generally necessary to achieve the desired detection levels.

NTS collected plutonium samples for 24 hours. In effect, the collection procedure normalized them to 

24-hour samples. Screenings for all other radionuclides were normalized to 24-hour samples in the 

bioassay records (Arent and Smith 2004).
Document No. ORAUT-TKBS-0008-5 Revision No. 03 Effective Date: 08/17/2012 Page 30 of 116

Allen et al. (1993) stated that NTS ascribed a time interval to each collected urine sample. The 

volume of the sample was measured and the interval was calculated assuming excretion of a constant 

volume of urine per day (1.4 L/day is the conventional value). Samples frequently spanned 

nonintegral periods (e.g., from Day 1.1 to Day 1.7), which gives an interval of 0.6 days ending 1.7 

days after the intake. In such cases, the excretion function could normally be extrapolated to the 

samples. The excretion function was adapted to the sample by reviewing the expected excretion for 

the interval from Day 1.1 to Day 1.7.

An alternative method used at NTS was to normalize bioassay results to integral 1-day values. For 

example, the result from Day 1.1 to Day 1.7 was adjusted to give the results from Day 1 to Day 2. 

This method enabled comparison of expected and empirical results. The disadvantage with this 

method was that the normalization procedure could have introduced additional error. However, for 

most cases at NTS, the exact time of intake was not well known and the normalization error was 

rather insignificant. Therefore, NTS used the normalization method because it permitted the graphical 

evaluation of data.

Because special bioassays were initiated in response to known or suspected intake conditions, the 

intake-to-bioassay period was assumed to be relatively short. At NTS, the urine sample was 

assumed to have the concentration expected at 24 hours after intake (i.e., the “instantaneous” urinary 

excretion function was used) (Allen et al. 1993).

5.2.4 Fecal Sample Collection and Analysis

NTS performed fecal sample analyses as a special bioassay, usually if there were other indications 

that internal contamination had occurred. Allen et al. (1993) indicates that 5-g ash samples were 

analyzed by alpha spectroscopy for radium, uranium, plutonium, and americium. BN (1998) clarified 

the collection and analysis volumes: The MDA assumes that a 24-hour sample was collected, half 

was ashed (resulting in 5 g of material), and 1 g of the material was measured. Alpha spectroscopy 

began in the late 1970s as a method of measurement. In the 1990s, some environmental projects to 

remediate plutonium-contaminated sites required periodic fecal bioassay.

Allen et al. (1993) states that the material measured in a fecal sample is the sum of excretion from the 

systemic body, translocation from the lungs, and unabsorbed ingested material accumulated over a 

certain period. This period could be difficult to specify because there can be considerable and 

variable lag time in the gastrointestinal tract. To minimize this problem, fecal samples should be 

collected over a period that is long in comparison to the gastrointestinal tract lag time (e.g., a week), 

and the data should be evaluated with an accumulated feces excretion model. This is particularly 

important in the first week after an intake. A single isolated fecal sample should be assigned the 

period between voids, and if this is unknown, a period of 1 day should be used.

Fecal samples might have been collected after an incident (Allen et al. 1993). Because fecal 

excretion patterns vary considerably, it was unreasonable to specify a minimal acceptable mass on a 

single sample, but 100-g wet mass was a good target value. Fecal samples were promptly screened 

for radioactivity and additional samples collected if radioactivity was detected. Incidents with 

significant potential for an intake warranted the collection of the first 3 to 7 days of fecal excretion. 

Because special bioassays were initiated in response to known or suspected intake conditions, 

intakes to bioassay periods were assumed to be relatively short. At NTS, the feces sample was 

assumed to consist of the material excreted in the first 24 hours after the incident (Allen et al. 1993).

5.3 IN VIVO MDAS, COUNTING METHODS, AND REPORTING PRACTICES

NTS had an in vivo bioassay program that included whole-body, thyroid, lung, and wound counting. 

Glasstone (1971) mentions a stationary WBC and scanners at the Southwestern Radiological Health 
Document No. ORAUT-TKBS-0008-5 Revision No. 03 Effective Date: 08/17/2012 Page 31 of 116

Laboratory, which was used for U.S. Environmental Protection Agency/Public Health Service offsite 

monitoring. The only significant time this facility was used for NTS workers was during the YUBA 

incident (June 1963), but the information obtained was not used. Workers involved in that incident 

were sent to Donner Laboratory in California for whole-body counting.

In 1964, LASL identified whole-body counting for radionuclides that deliver whole-body doses. There 

is no description of where these counts took place; the first reference to an onsite capability is 1967 

(Teasdale 1985). This TBD assumes that previous counts occurred at Los Alamos. It is also 

assumed that the University of California-Berkeley and LLNL performed counts at their facilities for 

their employees who spent time at NTS. Any records that pertain to employees of NTS contractors 

have been retrieved from other locations for inclusion in the record archives located in Las Vegas, NV. 

Other reports and memoranda indicated that before 1967 whole-body counting was performed by 

contracted portable units on three occasions in a 2-year span. No information was provided indicating 

the type of job categories measured or if this was initial routine surveillance.

In general, the WBC was used to measure amounts of individual gamma emitters in the body. The 

scanner measured radionuclides in specific parts of the body (e.g., 131I in the thyroid). A mobile 

scanner was available for use in the field; however, under most circumstances the field instrument 

that was available was used.

In 1967, a shadow shield-type counter was installed by Pan Am at the Nuclear Rocket Development 

Station as part of the Nuclear Space Propulsion Program (Teasdale 1985) for performing WBCs. This 

counter was transferred to NTS REECo dosimetry in 1974. Beginning in 1975, routine counting was 

performed for drillers, miners, and radiation monitoring personnel. In 1977, NTS considered adding 

lung counting for low-energy X-ray detection. However, it was not cost-effective to upgrade the 

existing facility. Construction of a new facility began in March 1979, and the facility operated from 

February 1981 until shutdown in 1999. See Attachment A, Section A.2.1 and Table A-9 for wholebody counting MDAs. 

5.3.1 Whole-Body Counting

Helgeson Nuclear Services (HNS) performed the first WBCs at NTS. It used a shadow shield that 

was transported in a semitrailer (HNS 1967). The shadow shield employed an 8- by 4-in. NaI crystal 

with a 7.6% resolution for 137Cs. The crystal was placed 10.5” above a moving bed. The count time 

was 8 minutes (480 seconds). HNS estimated the sensitivity at two confidence levels: 50% and 99%. 

Table 5-3 summarizes these activities.

Table 5-3. MDAs for the HNS shadow shield.

Isotope Background

Photons MDA–nCi

Energy (keV)

Intensity, Iγ

photons/decay 50% 99%

Cs-137 125 662 0.851 1.0 3.0

Zr-95 95 724 0.81 0.7 2.1

757 1.00

Co-58 57 811 0.994 0.8 2.3

Zn-65 54 1,116 0.506 1.5 4.3

Co-60 35 1,173 1.00 0.4 1.3

1,332 1.00

K-40* 71 1,461 0.107 7.1 21.4

(8.3 gK) (24.9 gK)

Source: HNS (1967).

* 1 gK = 0.000118 g of K-40 = 8.58138E-10 Ci = 0.858138 nCi/gK.
Document No. ORAUT-TKBS-0008-5 Revision No. 03 Effective Date: 08/17/2012 Page 32 of 116

The 99% MDA value was recommended. For the range from 662 keV to 1,461 keV, the maximum 

MDA is 2.55 nCi × Iγ. For nuclides not listed in Table 5-3 that have primary gamma energies 

≥100keV, the MDA can be determined for the primary photon with an intensity, Iγ, by 3 nCi/Iγ. Table 

A-9 lists other radionuclides that DRs might need to address and includes MDA estimates that were 

calculated based on this formula.

There were more than 300 WBCs performed before 1967 on this portable system. The calibration for 

this contracted system was performed by distributing 200 point sources (chips) through a standard 

man-sized Masonite phantom. Scatter factors and efficiency constants were determined by counting 

65Zn, 40K, 137Cs, 106Ru, 60Co, 54Mn, and mock 131I. Detection consisted of a 3-in. by 3-in. NaI(Tl) crystal 

on the first visit and an 8-in. by 4-in. NaI(Tl) crystal on the remaining visits. The standard counting 

time was 10 minutes. Table 5-4 lists the results of these counts.

Table 5-4. Frequency (number of workers) of nuclide 

appearance and concentration in subjects counted before 

January 1967.

Nuclide

Body burden (μCi)

≤0.02 ≤0.05 ≤0.1 ≤0.5

Zr-Nb-95 98 43 16 6

Ta-182 102 17 3 1

Ru-103, -106 34 1 - -

Ba-La-140 26 2 - -

I-131 2 - - -

Source: Church (1967).

In January 1967, Pan Am acquired and installed the shadow shield WBC. The detection system 

consisted of a 3-in. by 3-in. NaI(Tl) crystal in conjunction with a Nuclear Data 512-channel analyzer. 

A count time of 20 to 40 minutes was used. Calibration of this system consisted of points at 0.36, 

0.66, 1.12, and 1.33 MeV corresponding to photons emitted from mock 131I, 137Cs, 65Zn, and 60Co. 

The standards discussed above were placed in a standard man-sized Masonite phantom for counting. 

The detection efficiency for nuclides appearing between these points was extrapolated from the 

results. No efficiencies were reported, but research at the time indicated that, depending on body 

weight (120- to 185-pound range), the count rate could change by a factor of 1.57 (Church 1967). 

Table 5-5 lists results of the initial WBCs performed by this system.

Table 5-5. Frequency (number of workers) of nuclide appearance 

and concentration in subjects counted from January to August 1967.

Nuclide

MPBB and organ 

of concerna

Body burden (μCi)

≤0.02 ≤0.05 ≤0.1 ≤0.5

I-133 0.3 / thyroid 3 3

Te-I-132 0.3 / thyroid 3 1

I-131 0.7 / thyroid 4 1

Ru-103, -106 (b) / kidney 27 25 11 3

Zr-Nb-95 20 / total body 36 24 7 4

Ba-La-140 4.0 / bone 3 1

Ta-182 7.0 / liver 1 -

Source: Church (1967).

a. The MPBB and organ appear in handwritten notes on the original table. 

Documentation for the quantities and organs of concern has not been found.

b. The handwritten notes indicate 20 μCi for Ru-103 and 3 μCi for Ru-106.

The January 1977 through January 1981 REECo standard procedures for chemical and radiological 

analysis describe the WBC program (REECo 1969–1987). During the preliminary calibration of 

WBCs, the phantom consisted of polyethylene blocks stacked in a 3 by 5 by 5 arrangement with the 
Document No. ORAUT-TKBS-0008-5 Revision No. 03 Effective Date: 08/17/2012 Page 33 of 116

source placed in one of the blocks in the center column. The source stack was moved over an x,y

grid marked on the bed. By varying the position of the source over the x,y,z coordinates, it was 

possible to simulate a source distributed throughout a phantom that would have been 5.9 in. thick, 

22.4 in. wide, and 74.6 in. long. The average density of the phantom was 0.698 g/cm3 (REECo 1969–

1987). Table 5-6 lists the sources used for this calibration.

Table 5-6. Preliminary calibration sources for 1977 

through 1981.

Radionuclide

Energy 

(keV)

Branching ratio 

gamma/disintegration

Cr-51 320 0.09

Mn-54 835 1.00

Zn-65 1,115 0.49

Zn-65 511 0.34

Y-88 898 0.91

Y-88 1,836 1.00

Source: REECo (1969–1987).

The efficiencies for counting gamma rays from this phantom are described by the equation:

Efficiency(counts/disintegration) = 0.0089e-0.000175E (5-4)

where E is energy in keV (REECo 1969–1987).

Sensitivity was defined as activity that is detectable with a relative 2-sigma error of +20%. 

Background counts were 1,000 seconds. Table 5-7 lists the sensitivities for potassium and cesium at 

a background count of 45 cps.

Table 5-7. WBC sensitivities.

Radionuclide

For 20-minute 

count (μCi)

For 40-minute 

count (μCi)

K-40a 0.104 0.091

Cs-137b 0.011 0.010

Source: REECo (1969–1987).

a. Potassium activity in standard man is about 0.12 μCi.

b. The MPBB of Cs-137 in standard man is 30 μCi.

Potassium-40 activity in the body was routinely calculated and normalized to the 70-kg standard man. 

The results were hand-recorded on the WBC printout. Abnormal results, defined as those in excess 

of 200 g potassium, standard man equivalent, were reviewed by a senior health physicist to determine 

the need for further action (REECo 1978, p. 22).

REECo standard procedures for personnel dosimetry and records indicate that for 1978 and 1979, 

whole-body counting used a 20.3-in.-diameter by 10.1-cm-thick NaI crystal housed in a shallow shield

(REECo 1978, p. 22; REECo 1979, p. 25). (The detection energy range was 100 to 2,500 keV.) The 

system was connected to a Nuclear Data Analyzer, Model 2400, and the data printed through a 

teletypewriter. The monitored individual lay supine on the table during the counting procedure, which 

took 1,300 to 1,400 seconds. These counts were performed in street clothing; however, if the count 

level in the 40K channels appeared to be excessive, a recount was performed after street clothes were 

exchanged for a paper suit. In addition, the background of the bed was rechecked. There is no 

available documentation for 1980; it is assumed the 1979 equipment and procedures were used until 

1981.

By 1981, a 29.2-cm-diameter by 10.1-cm-thick NaI(Tl) detector connected to a Canberra Series 30 

multichannel analyzer was used (REECo 1981, pp. 7, 20). The use of paper suits became standard 
Document No. ORAUT-TKBS-0008-5 Revision No. 03 Effective Date: 08/17/2012 Page 34 of 116

practice and the count time changed to 2,000 seconds. To provide lower background counts, a “vault” 

was constructed of pre-World War II steel plates covered with graded shielding consisting of 3 mm of 

lead, 0.6 mm of tin, and 0.25 mm of zinc. Radon daughter products in the air were reduced by a highenergy particulate air (HEPA) filtration system (Teasdale 1985).

From 1983 to 1987, two types of detection equipment were identified (REECo 1983, pp. 19, 21; 

REECo 1987, pp. 16–17): 

1. A 29.2-cm by 10-cm-thick NaI(Tl) crystal used to scan for radionuclides with gamma energies 

in excess of 100 keV. The WBC was calibrated using a Bomar water phantom constructed of 

polyethylene that weighs 70 kg when filled with water. Four monoenergetic gamma-ray 

emitters were used as dose material for the solutions. An energy response curve was 

generated for the results using energies of 51Cr, 54Mn, 65Zn, 137Cs, and the high-energy peak of 

88Y.

2. Phoswich counting systems were used to detect X-rays and gamma rays of less than 100 keV. 

These detectors consisted of a 12.7-cm-diameter by 5-cm-thick CsI(Na) crystal that was 

optically coupled to a 12.7-cm-diameter by 3-mm-thick NaI(Tl) crystal at the incident energy 

end. The crystals were hermetically sealed in a stainless-steel housing with a 0.25-mm 

beryllium entrance window. Electronics were modified to capture the signal from both parts of 

the Phoswich detector. Once-a-year calibration used a Bomab water phantom constructed of 

polyethylene that weighed 70 kg when filled with water and National Bureau of Standards 

(NBS)-traceable solutions of four monoenergetic gamma ray emitters. Phoswich detector 

system calibration used a tissue- and bone-equivalent torso phantom designed by LLNL. 

Tissue-equivalent plastic that contained 241Am, 238/239Pu, and DU was used to construct the 

lungs and liver. Curves were generated for the phantom composition of muscle, bone, and 

varying thicknesses of fat and muscle.

Employees were issued paper suits and assumed a supine position on the counting table within the 

vault. The detector moved from one end of the bed to the other over a 20-minute period (REECo 

1983, pp. 19–20; REECo 1987, p. 16). There is no available documentation from 1988 to 1993; it is 

assumed the 1987 equipment and procedures were in use until 1993.

The system used in 1993 was calibrated for whole-body counting and used for whole-body, thyroid, 

and wound counting (Allen et al. 1993). The 70% relative efficiency detector was used routinely in 

this period with a 27% relative efficiency detector as a backup. The 70% relative efficiency detector is 

referred to as an “XtRa-extended range,” closed-end, coaxial detector with a diameter of 72.4 mm and 

length of 69.5 mm. The backup detector is a p-type, closed-end, coaxial detector with a diameter of 

61.3 mm and length of 56 mm. Both detectors measured photons with energies in the calibration 

range of 50 keV to 2.5 MeV. Workers were counted while sitting in a reclining chair. The chair back 

was about 30 degrees from vertical so the back and the seat yielded an arc with the detector of 50 

cm, providing a full view of the body trunk in relation to the detector. 

WBCs were performed (Teasdale 1985, Allen et al. 1993):

• For new employees who were likely to be included in the routine bioassay program

• For a current employee who changed to a job classification that required a routine bioassay 

program

• Annually for employees who were currently on the routine WBC and bioassay program

• For terminating employees
Document No. ORAUT-TKBS-0008-5 Revision No. 03 Effective Date: 08/17/2012 Page 35 of 116

• For employees who had a suspected intake of radioactive material, particularly gammaemitting fission and activation products.

As identified in Allen et al. (1993), the MDA was calculated as a function of the gross spectral counts 

and the full-width, half-maximum (FWHM) energy and was consistent with American National 

Standards Institute standard N13.30 (HPS 1996). The MDA (assuming a 5% probability of either a 

type 1 or type 2 error) is defined as:

( )

0.5

1 2

1

ΣC W C C CW i Y

MDA

T EY k

+

−

  + ×     = ×××

(5-5)

where:

Yi = number of counts in gross spectrum

Y = photon yield, 1.0 for “unknown” peaks

C = centroid energy

W = window function (area window 2), w = 0.64 * FWHM

T1 = count collection time

E = detection efficiency at centroid energy of interest

C1 = reject MDA constant, user variable, set at 2.71

C2 = reject MDA sigma, user variable, set at 4.66

k = unit conversion factor

The suggested priority for DRs with regard to MDAs and other detection thresholds is to use (1) the 

limit on dose report if available, (2) Attachment A, Table A-9 values, or (3) an appropriate published 

value from another DOE site or an applicable value referenced in the literature (NIOSH 2002).

5.3.1.1 Effect of Cesium on Whole-Body Counting

Fallout affected everyone in North America; body burdens of 137Cs measurable in WBCs were 

common in the 1960s and 1970s. National Council on Radiation Protection and Measurements 

(NCRP) Report 94 (1988) provides mean body burdens of 137Cs for the United States for the years 

most likely to produce interference with occupational WBC results. Because of the nature of the work 

and potential exposure at NTS, unless there is evidence that the 137Cs result is not associated with 

employment at NTS, the result should be evaluated as occupational with no adjustment for the values.

5.3.2 Chest Counting

It has not been determined if lung counting was a viable bioassay method before construction of the 

WBC facility from pre-World War II steel. However, safety reports from the atmospheric testing era 

mention chest and lung counting.

The first reference to lung or chest counting was found in a REECo standard procedure for personnel 

dosimetry and records dated July 1983 (REECo 1983, pp. 19–20). The procedure stated that for lung 

counts, the subject assumed a supine position, crystals were placed over the anterior lung region, and 

the counting time was 40 minutes. Lawrence Livermore Laboratory software to process lung counting 

data is mentioned in a 1981 REECo standard procedure for laboratory instruments and equipment 

(REECo 1968–1987a, p. 67).

Allen et al. (1993) identified lung counts with nonroutine bioassay for the suspected intake of thorium 

from the use of special nuclear materials, uranium, or any of the TRU elements, especially californium 

and curium. Lung counts for low-energy photon emitters were performed with four Canberra Low 
Document No. ORAUT-TKBS-0008-5 Revision No. 03 Effective Date: 08/17/2012 Page 36 of 116

Energy Germanium detectors mounted on two adjustable support arms. Each detector had a window 

thickness of 0.5 mm and an active diameter of 50.5 mm, yielding a total active detection area of about 

8,000 mm2 for the array. The detector array measured photons from the lung area within the 

calibration energy range of 15 to 400 keV. With the individual in a reclining chair, one two-detector 

array was placed on each side of the chest. The placement minimized the CWT interference while 

maintaining proximity to the lung and bronchial region, thereby optimizing the number of photons 

measured.

Data were analyzed using Canberra Industries software (ABACOS Plus) from physical data entered 

for the individual being counted. The software used a modified peak analysis technique that 

determined the areas of photo peaks in the spectrum after subtracting the underlying continuum 

background. It used two consecutive methods for determining peak location – a library-driven peak 

search, using a radionuclide library file, followed by a sliding-peak analysis to locate spectral peaks 

not included in the library. The routine time for a lung count was 1,000 seconds (16.67 minutes). 

Attachment A, Table A-10 lists the MDAs for chest lung counts.

The method used for lung counting at NTS in the early 1990s was based on published biometric 

relations for CWT measurement relations to weight and height (Allen et al. 1993). The chosen 

relation was:

CWT = 1.973(W/H) – 2.0038 (5-6)

where:

CWT = chest wall thickness (cm)

W = weight (lb)

H = height (in.)

The Livermore torso phantom was fabricated to provide calibration information for TRU nuclides in the 

lungs of individuals with a range of body statures and chest wall tissue compositions. The phantom 

set consists of the basic torso and three sets of chest plates or overlays. The purpose of the chest 

plate sets is to simulate tissue compositions of 100% muscle, 50% muscle to 50% adipose, and 13% 

muscle to 87% adipose. The phantom covers a CWT range of about 15 cm without chest plates to 

about 40 cm. with the thickest overlays. There is some set-to-set variation.

Direct measurements of plutonium in the body, typically via lung counts, are capable of detecting only 

very large intakes or extended chronic exposures. The site no longer performs lung counting

(McMahan and Ogurek 2003).

5.3.3 Thyroid Counting

During weapons testing at NTS, exposure to radioiodines was a major concern for the first 100 hours 

after detonation. Thyroid counts were a concern for workers involved in drill-backs or reentry/minebacks in this 100-hour timeframe. DRs can determine potential exposure within 100 hours of a test by 

reviewing employee access records in the files provided by DOE for the claim. Radioiodines were 

produced directly from fission and as daughter products of other fission nuclides. This was 

particularly important for individuals involved in cloud sampling after an aboveground test or 

individuals on the ground after the fallout pattern.

REECo (1961) states that a determination of 131I in human and animal thyroid was made by direct 

gamma counting. In the field, this entailed taking a background count at 25 cm from the thigh of the 

technician or the worker and then taking a count of the worker thyroid at 25 cm. The DR might find 

this referred to in employee records as a “prompt count.”
Document No. ORAUT-TKBS-0008-5 Revision No. 03 Effective Date: 08/17/2012 Page 37 of 116

REECo procedures from 1968, 1972, and 1975 described a thyroid count performed after a urinalysis 

that indicated an individual had been exposed to radioiodine (REECo 1968–1987b). A 2-in. NaI(Tl) 

crystal detector was placed next to the individual’s larynx and a 5-minute gross gamma count was 

performed. Background was based on measurement of the machine operator’s thigh; in later years 

the thigh of the individual being counted was used. Nasal swabs were sometimes obtained in an 

effort to determine if the nose and mouth were contributing gamma activity. Results were handcalculated using appropriate correction factors and decay constants. A record of each measurement 

was maintained as the individual returned for follow-up data acquisition. In 1993, thyroid counting 

was performed with the whole-body counting system by positioning the detector over the appropriate 

area and counting for 40 minutes (Allen et al. 1993). The following paragraphs contain information on 

the ratios of the iodines in the period immediately after a test. DRs should use this for individuals 

directly involved in incidents such as Baneberry or reentry activities as indicated in the records 

provided by DOE (i.e. indication of potential for exposure based on bioassay or decontamination 

results for the worker).

5.3.3.1 Iodine

An NTS internal report, Iodine, 1960–1963 (REECo undated), includes method development 

information and post-test drilling air sampling results, isotopic ratios, thyroid doses by radionuclide, 

iodine thyroid dose related to external gamma dose, an in vitro blood testing protocol, percent 

contribution to dose among radionuclides, relative thyroid doses from inhalation of curie-per-cubicmeter concentrations of each radionuclide, and a comparison of external whole-body dose and 

internal dose to the thyroid.

After a nuclear detonation, several iodine radionuclides are produced; those with mass numbers 131 

through 135 are of significance from the standpoint of exposure to personnel. Table 5-8, derived from 

data cited in Kathren (1964), provides the relative activity normalized to that of 131I, at times ranging 

from 1 hour to 1 month after fission of 235U. Hence, if the time of detonation and the activity of one 

iodine radionuclide are known, the activity of the other iodine radionuclides can be determined. DRs 

should consult access records in the information provided by DOE to determine likely exposure time 

after the test.

Table 5-8. Relative abundance of radioiodines.

Nuclide 1 hr 2 hr 4 hr 10 hr 1 day 2 day 4 day 1 wk 2 wk 1 mo

I-131 1.0 1.0 1.0 1.0 1.0 1.0 1.0 1.0 1.0 1.0

I-132 10.1 5.9 4.5 4.2 3.4 1.9 1.7 1.4 0.4 --

I-133 22.2 14.8 14.2 13.4 9.5 6.2 1.1 -- -- --

I-134 344 142 68.1 0.9 -- -- -- -- -- --

I-135 84.4 59.1 43.6 16.6 4.5 0.3 -- -- -- --

The relative activity from each of these iodine radionuclides at early times after fission is a complex 

function of the ingrowth and decay of the radionuclides and their fission chain precursors, dependent

to some extent on whether produced by thermal or fast fission and by fission of what nuclide. At times 

ranging to about 4 hours after fission, 134I is by far the predominant radioiodine activity, accounting for 

almost all activity from the five radionuclides. The peak activity from this radionuclide occurs about 

three-fourths of an hour after detonation, after which ingrowth of activity is no longer dominant and

radioactive decay of this short-lived species (T1/2 = 53 minutes) becomes predominant. The peak 

activity from this nuclide is 26 MCi per kiloton of explosive yield, compared to 0.11 MCi from 131I, 

which occurs about 5 hours after detonation (Holland 1964). A similar but less rapid effect is apparent 

for the other iodine radionuclides in relation to 131I. At times less than about 1 hour after detonation, 

the relative activity from the higher iodines is much larger than at later times, primarily as a result of 

the rapid ingrowth of 131I, whose absolute activity peaks about 5 hours after detonation, thereafter 
Document No. ORAUT-TKBS-0008-5 Revision No. 03 Effective Date: 08/17/2012 Page 38 of 116

declining approximately according to its characteristic 8.05-day half-life. Kathren (1964) contains a 

graphic representation of the relative activity from the five iodine radionuclides of significance.

A calculation based on the maximum iodine concentration for any event (i.e. Baneberry venting 

December 1970) provides an indication of the potential dose during a release. A concentration of 

1.85 × 10-12 μCi/cm3 was measured at Camp 12 on December 24, 1970, resulting in a calculated 

value of 3.1 × 10-12 μCi/cm3 on December 18, 1970, the day of the venting. Assuming a breathing 

rate of 20 L/min and an exposure of 2 hours during the evacuation of the site, an intake of 

7.44 × 10-6 μCi is reasonable and results in a dose of less than 1 mrem to the thyroid.

5.3.4 Wound Counting

Wound counting was performed (Allen et al. 1993) with the whole-body counting system by 

positioning the detector over the appropriate area 4 in. from the wound and counting for 40 minutes. 

The whole-body counting system was calibrated for these measurements by placing the germanium 

coaxial detector 4 in. away and perpendicular to the thyroid phantom or the simulated wound. A vial 

containing the same concentrated multiradionuclide solution used in the WBC calibration was placed 

in the thyroid phantom or at a location to simulate a wound and counted for 40 minutes.

BN (2000) discusses wound activity levels and states that, when contamination is detected in the area 

of a wound, measurements should be made with a wound monitor. McMahan and Ogurek (2003)

states that a wound counter is maintained on the site as an investigative tool for situations in which 

immediate indications of potential intakes through wounds might be needed.

This background information is provided for historic purposes only; the programs used to determine 

internal and external dose for the NIOSH project cannot convert wound dose to dose to a specific 

organ.

5.4 PERSONAL AIR SAMPLING DATA

NTS air sampling records are workplace rather than intake monitoring records, and NTS does not 

maintain them in the same collection as bioassay records. Because air sampling records are 

associated with each facility, they are stored as facility records. Boxes containing workplace 

monitoring records can be retrieved from storage. However, locating a specific set of air sampling 

records would be very time-consuming and should be considered only as a last resort. Correlation 

between air sample concentrations in given rooms or work locations and a specific person would be 

difficult.

Radiological safety documentation describes historical air sampling. REECo prepared onsite 

radiation safety reports for each operation starting in 1957 (e.g., REECo 1957, 1958); these reports 

outline the support REECo provided to the testing groups [LANL, LLNL, and U.S. Department of 

Defense (DOD)/Defense Nuclear Agency (DNA)/Sandia National Laboratories (SNL)]. Additional 

documentation back to 1951 prepared by the laboratories or the Atomic Energy Commission is in a 

narrative form. The REECo reports briefly outline radiological support functions and describe specific 

support for each test until the end of the testing era in 1992. Topics discussed in the reports include:

• Test name, sponsor, location, type, date, and time.

• Number and location of remote telemetry stations and a summary of results.

• Details on initial reentry surveys, reentry and mine-back (for tunnel test), and post-test drilling. 

In the early 1980s, brief activity and operations logs were included.
Document No. ORAUT-TKBS-0008-5 Revision No. 03 Effective Date: 08/17/2012 Page 39 of 116

• A final summary statement (e.g., No whole-body external or internal exposures exceeded the 

established control limits and no toxic gases or explosive mixtures were detected during the 

post-test drilling period.)

• A list of the radiation and toxic gas detection instruments in use.

Allen et al. (1993) stated that, in general, air monitoring devices were positioned to provide samples 

representative of the worker’s breathing zone. However, some work conditions, such as mining and 

drilling, might have required location of the air monitor intake in the area of highest expected 

concentration “to ensure that airborne radioactivity possibly breathed by the worker is not overlooked.” 

Workplace air monitoring was required in occupied areas that had the potential to exceed 10% of any 

derived air concentration (DAC) value listed in DOE Order 5480.11 (DOE 1988).

The principal workplace air monitoring device used at NTS was the retrospective air sampler (RAS) 

(Author unknown undated a). Retrospective air samplers (RASs) were used at the decontamination 

facility (Decon Pad) during post-test drilling, tunnel reentry, and other special operations. A typical 

RAS consisted of a portable or permanently installed air sampling system such as a vacuum pump 

with an attached meter to measure the sample flow rate or sample volumes. The ambient air sample 

was normally pulled through a 4-in. or 7-in. cellulose filter and a 4-in. activated charcoal cartridge at 

the rate of about 1 to 3 ft3

/min. The volume of air per unit time was dependent on the size of the 

pump, the number of samples collected by that pump at one time, and how loaded the samples were 

with material that resisted flow. Depending on the location of the sampler and the work being 

monitored, the filters were replaced at a frequency from every shift to every week. RASs provided a 

history of work environment airborne contamination levels to ensure that workers were not exposed to 

unexpected levels of airborne radioactive material.

Air samples collected from most work areas were counted at the REECo laboratory in Mercury, 

Nevada (Allen et al. 1993). Air samples from LANL work sites were counted by personnel at the 

LANL Health and Safety Laboratory in Mercury. Results from both laboratories were reported in 

microcuries per milliliter. Air samples from the Decon Pad (Area 6) were counted at the facility in a 

thin-window detector scaler system for beta- and alpha-emitting radioactivity; results were reported in 

microcuries per milliliter or in DAC fractions. Charcoal canisters were not used with Decon Pad air 

samples.

Another sampler used at NTS was the continuous air monitor (CAM), which consisted of a vacuum 

source, collection media, and a detector for alpha, beta, or beta-gamma radiation. When radiation 

was detected in excess of preset limits, audible and visual alarms were triggered in the affected areas 

to alert personnel of the release. CAMs were used at the Treatability Test Facility and on drill rig 

floors during LANL drill-backs. CAMs detected radioiodines, tritium, plutonium, or other radionuclides 

of concern (Allen et al. 1993).

5.4.1 Drilling Operations

A 1959 memorandum titled “Operational Guides for Above-Ground Drill Sites into Ground Zero Areas” 

(Wilcox 1959) stated that if surveys indicated significant beta-gamma concentrations, a 1-hour Staplex 

(8-in. by 10-in. glass-fiber filter) sample and nasal swabs were collected from each person possibly 

exposed. In addition, REECo published Air Sampling for Radioactive Particulate Material, which 

included the following air sampling information (Geiger 1959):

Information relative to hazards associated with internally deposited radioactive 

materials has been increasing continuously since 1953 when NBS Handbook 52 was 

published. In June 1959, NBS Handbook 69, containing the new recommendations of 
Document No. ORAUT-TKBS-0008-5 Revision No. 03 Effective Date: 08/17/2012 Page 40 of 116

the NCRP, was published. Handbook 69 supersedes Handbook 52 as the guide for 

evaluating and controlling internal radiation hazards at NTS.

The recommended maximum permissible concentration in air for an unknown alpha 

emitter, (MPCU)a, for continuous occupational exposure is 4 E-07 uCi/m3

, and for a 40-

hour work week the recommended (MPC)a (excluding 210Pb, 227Ac, 228Ra, and 241Pu) for 

continuous exposure is 1 E-04 uCi/m3

. For a 40-hour work week this has been 

increased by a factor of three to 3 E-04 uCi/m3

.

For control purposes, the (REECo) Rad-Safe Division will determine concentration of 

airborne alpha emitters down to 10% of (MPC) for 239Pu, i.e., down to 2 E-07 uCi/m3

, 

and concentration of beta emitters down to 10% of the (MPCU)a value, i.e., down to 3 

E-05 uCi/m3

.

In order to measure (MPC)a concentrations within ± 10%, 90% of the time, a sample 

volume of 44 m3 of air must be pulled through the filter. For the large filter (Staplex) 

samples the total volume should be 66 m3

. With these volumes collected on a glass 

fiber filter, the counting errors with the large area proportional detector for various 

concentrations of 239Pu would be as follows:

Air concentration 

uCi/m3

Relationship to 

(MPC)a for Pu-239

Counting error at 90% 

confidence level

2 E-06 1 ± 10%

1 E-06 0.5 ± 17%

5 E-07 0.25 ± 22%

2 E-07 0.10 ± 54%

Some typical sampling times for various samples are as follows:

Sampling equipment Flow rate Typical sampling time

Staplex 8” × 10” filter 40 cfm 1 hr

Staplex

4”-diameter filter

20 cfm 1 hr 20 min

Filter Queen 4 cfm 6 hrs 30 min

Gelman “Caddy” 3 cfm 8 hrs 40 min

Gelman Diaphragm 0.37 m3/hr 5 days

The above sampling times, which were calculated based on 239Pu, will provide plenty of 

sample for assay of long half-life beta emitters.

Continuous air sampling is preferred over “grab” type sampling. If the operation is 

likely to be such that air sampling is necessary only a few hours, the Staplex machine 

should be used. If the need for air sampling is for one day or more, either a Filter 

Queen or “Caddy” air sampler should be used; however, if neither of these air samplers 

is available, the Staplex may be used as a “second choice.” If the history of the work 

location indicates that respiratory protection will probably be unnecessary for a 5-day 

period or more, the low volume Gelman Diaphragm pump is the preferred choice.

Radon and thoron may occasionally be a problem in tunnel areas. The most 

expeditious manner for ascertaining whether or not it is a problem is as follows: Collect 

a 5-minute air sample with a Staplex 8” × 10” filter using an air sampler pulling 

approximately 35 cfm. Immediately after the completion of the 5-minute sampling time, 

monitor the surface of the filter with an Eberline PAC-G survey meter. A reading of 

25,000 cpm on a PAC-G that has been calibrated for 2” geometry corresponds with the 
Document No. ORAUT-TKBS-0008-5 Revision No. 03 Effective Date: 08/17/2012 Page 41 of 116

(MPC)a for a 40-hour work week. Anything above this will normally require respiratory 

protection. Sometimes the monitor needs to know if the alpha activity detected on a 

filter is radon and thoron or long half-life emitters without waiting for decay. A rough 

indication may be obtained as follows: Place the filter in a Glassine envelope. 

Compare the reading obtained with PAC-G with the filter in the envelope with that 

obtained with the filter outside the envelope. Radon and thoron alpha will be reduced 

to approximately 30%. If less than 30% of the alpha particles penetrate one thickness 

of the envelope, long half-life alpha emitters may be present in significant 

concentration. If no activity is detected through the envelope, the majority of the 

activity is long half-life alpha emitters.

If the long half-life alpha emitter is known to be 235U instead of 239Pu, more liberal 

(MPC)a values can be used and correspondingly shorter air sampling times may be 

adequate. For example, the (MPC)a for a 40-hour work week recommended for 235U is 

1 E-04 uCi/m3 as compared with 2 E-06 uCi/m3 for 239Pu. In like manner, the (MPC)a

for short half-life fission products may be more liberal. The (MPCU)a for an unknown 

beta emitter is 3 E-03 uCi/m3 for a 40-hour work week if none of the following beta 

emitters are present in significant concentrations when compared with the (MPC)a

values: 90Sr, 129I, 210Pb, 227Ac, 228Ra, 230Pa, 241Pu, and 249Bk. Since 90Sr is the only 

fission product in this group of beta emitters the more liberal (MPCU)a value is more 

appropriate in the case of a normal mixture of short half-life fission products.

REECo (1976) noted differences between LANL and LLNL in performing drill-back operations 

conducted in 1966 and 1967. LANL used a pressurized recirculation containment system, whereas 

LLNL used a vent line system designed to trap radioactive effluent released from the drill casing. In 

the vent line system, air from around the drill casing was pulled through the vent line into a large 

container of activated charcoal by a positive displacement pump. Radioactive effluent, especially 

radioiodine, was absorbed on the charcoal and allowed to decay under controlled conditions. ORAUT 

(2008) contains additional information about post-test drilling.

An array of four air-sampling heads, one on each corner of the drill rig platform, provided 

continuous monitoring of the air on the platform. This array was activated at the start of drilling and 

operated continuously until closure of an abandonment valve. The results of this sampling gave an 

indication of the exposure, if any, experienced by workers on the platform during the drilling operation 

(Johnson et al. 1966). Documentation in 1963 first noted this type of air sampling on the drill platform, 

which continued through the end of the testing period in 1992.

In the appendix (McClendon and Eubank 1964) titled “A Typical Detailed Safety Support Plan for 

Underground Nuclear Tests,” REECo provided air sampling procedures on the drill platform. This 

report covered Operation Storax, which occurred in 1962 and 1963. The appendix stated that a lowvolume air sampler equipped with a Whatman-41 prefilter and a carbon cartridge was positioned at 

breathing level on each drill platform at the beginning of the drilling operation. The prefilters were 

changed when they measured 300 mR/hr or, in any case, at the end of 8 hours. In addition, a highvolume air sampler (similarly equipped and positioned) was used on each drill platform for 5 minutes

of each half-hour. Air sampling started at the beginning of the drilling operation and continued until 

the drill hole was capped. All air sample prefilters were checked periodically for alpha radiation. 

During the first 30 hours after the event, the charcoal cartridge and the prefilter from each air sampler 

were screened individually on a gross gamma counting system to determine if any samples contained 

significant activity. The detection limit of the system (a 3-in. by 3-in. thallium-coated NaI crystal) was 

approximately 500 pCi (1-minute count). When gross gamma screening indicated significant activity, 

sufficient multichannel and gross-gamma decay data were collected to provide decay curves and 

positive identification of significant radionuclides, isotopic ratios, etc. Sample results were reported as 

microcuries per cubic meter at mid-sampling time for each nuclide detected (other than radioxenons).
Document No. ORAUT-TKBS-0008-5 Revision No. 03 Effective Date: 08/17/2012 Page 42 of 116

5.4.2 Tunnel Operations

In 1961, REECo (1961) published specific air sampling guidelines for B and E Tunnel operations as 

follows:

• B Tunnel: A 5-minute air sample was collected with a Staplex (8” × 10” glass fiber filter) 

approximately one, three, and five hours after the beginning of the shift. The sample was 

measured for alpha activity immediately and one hour after collection (radon-thoron check). 

These samples were not transmitted to the lab.

A Gelman diaphragm pump air sampler near the work areas was checked for operation each 

shift and the filter was exchanged each Friday.

• E Tunnel: A 5-minute air sample was collected with a Staplex (8” × 10” glass fiber filter) each 

shift. The sample was measured for alpha activity immediately and one hour after collection 

(radon- thoron check). These samples were not transmitted to the lab.

The Filter Queen sampler was checked for operation each shift and the filter was exchanged 

each Friday.

In both tunnels, if significant beta-gamma concentrations were indicated by portable instrument survey 

or radon-thoron checks, a 1-hour Staplex (8-in. by 10-in. glass-fiber filters) sample was obtained and 

nasal swabs were collected from each person possibly exposed. In 1962, B and E Tunnel-specific air 

sampling guidelines were revised such that 30-minute air samples were collected rather than 5-minute

samples. In addition, after the 1-hour alpha check, the samples were transmitted to the Area 12 

Change House counting facility (Wilcox 1962).

Characterization of the tunnel environments occurred on a routine basis. Reports titled Current 

Tunnel Survey, Radiological Conditions and Reentry Requirements (Bicker 1975) are available for the 

period from 1969 to 1980. The reports provide information for each tunnel by individual drift locations. 

They include exposure rates, note if airborne radiation is present (yes/no), and specify personnel 

protection equipment (PPE) for workers and visitors. The cover letters for these reports state that air 

samples were collected at representative vents before reentry (going in a tunnel that has been 

inactive for a period, not a post-test reentry).

There are detailed tunnel air sampling reports covering the period from 1991 to 1993. These reports 

give ambient air sampling results and, if a test occurred during the reporting period, air sampling 

results obtained during post-test reentry including isotopics. Tunnel air sampling is described in Lyons 

(1992), which stated, “Of the NTS operations air sampling programs in the tunnel complexes, the 

initial mining and event reentry and recovery operations represent the only real airborne radioactive 

inhalation potentials to personnel.” Weekly Drierite samples for tritium analysis were collected in the 

active tunnel complexes to document any changes in normal background levels or reentry drifts as 

they were advanced toward ground zero areas. Underground water sources were considered primary 

transporters of tritium from old event areas.

Fission product beta/gamma emitters are the radionuclides of greatest interest for NTS operations 

(Lyons 1992). An air sample was collected in any work area where the potential for airborne 

contamination might have existed. These areas were identified using past air sample results from 

similar operations or from the type of work performed in the facility. The REECo laboratory in Mercury 

analyzed most air samples collected on a routine basis using gamma spectra analysis to determine 

beta/gamma radionuclide concentrations. The laboratory also analyzed Drierite samples. Gross 

alpha and beta/gamma concentrations were calculated at facilities where routine detailed radionuclide 

analysis was not required. It was REECo policy to perform air sampling in active work areas 
Document No. ORAUT-TKBS-0008-5 Revision No. 03 Effective Date: 08/17/2012 Page 43 of 116

containing contaminated debris or in areas where mining and drilling operations had the potential to 

encounter radioactive materials in fissures, cavities, or water. The sample program continued until 

the job was complete. The gamma spectra results for routine air samples were reviewed by REECo 

health physicists. Results indicating the presence of fission products were investigated and 

engineering and administrative controls were evaluated to prevent generation of airborne material 

during subsequent operations.

Routine air samples consisted of a Whatman 41 cellulose fiber prefilter and a Mine Safety Appliances 

Company (MSA)-activated charcoal cartridge housed in a 4-in.-diameter sample head. Several types 

of calibrated pumps with airflow calibrated liquid petroleum gas meters determined total air volume 

samples. Air sample data were reported in microcuries per milliliter. Tritium water vapor samples 

were drawn from ambient air in areas where elevated tritium levels were suspected or to collect 

background information for future reference. The collection system consisted of a 10/20 mesh of 

Drierite housed in a 10.75-in. by 1.75-in.-diameter Lucite column. Samples were drawn through a 

0.188-in. aperture at flow rates of 3 to 4 L/min as measured with a calibrated rotometer. Collection 

times ranged from 3 to 6 hours to attain an 80% saturation level. These data were also reported in 

microcuries per milliliter (Lyons 1992).

For iodine air sampling, bioassay records indicate if the samples were collected with a charcoal 

canister or a filter (Arent and Smith 2004).

5.4.3 Examples of Air Sampling Results

The Report of the Test Manager on 56 Project – NTS details air sampling activities during support of 

plutonium dispersal experiments in Area 11 (AEC 1956). Personnel from Sandia, LANL, and EG&G 

were involved in Project 56 field operations. REECo established air sampling stations for routine 

collection of airborne material on a filter medium at each location where personnel worked above 

specified contamination levels. Stations were located at the camera trailer, the field decontamination 

station, the field control point, the J-12 count shack on Yucca Lake, and the Rad-Safety building (Area 

6, CP-2). A tolerance level of 8.78 dpm/m3 was used for air concentrations at these locations. At 

certain times, air samples were collected from the rooms of the field decontamination station and 

around ground zero sites. REECo maintained an access log for personnel who entered the 

contamination area (CA). The sampling found air concentrations as high as 72,000 dpm/m3 in rooms 

where personnel decontamination occurred. Sample recovery air sampling results as high as 

14,600 dpm/m3 were measured and surface contamination levels greater than 2 × 106 cpm were 

recorded at ground zero. Maximum personnel contamination levels (outer coveralls) were measured 

above 20,000 cpm alpha.

Operation Plumbbob On-Site Rad-Safety Report (REECo 1957) stated that 1945 air samples were 

collected for the test series conducted from April 24 to October 7, 1957, which included a series of 

balloon, tower, surface, rocket, shaft, and tunnel tests. REECo collected air samples for long-lived 

alpha activity at designated locations (e.g., Warehouse 6, CP-2, Area 2, Gate 385, Area 13, Well 5B). 

In addition, REECo provided air sampling support to Project 57, which was a surface plutonium 

dispersal test (a safety test with zero yield conducted on April 24, 1957 in Area 13). In addition to the 

environmental fallout trays, REECo assisted Sandia Corporation in collecting samples from impactor 

stations on the day of the test. Continuous air samples were taken around the Decontamination 

Building and in the ground zero area and analyzed in the Rad-Safe Laboratory. Daily radiological 

surveys were made inside the Decontamination Building and in the surrounding “clean” areas using 

alpha survey instruments, filter paper swipes, and air samplers. An initial post-detonation survey of 

the ground zero area indicated extensive alpha contamination, but no significant beta-gamma activity. 

Table 5-9 lists results of the air samples obtained on April 30, 1957, at locations 500 and 1,000 ft from 

ground zero with a 15-knot wind blowing from the southeast.
Document No. ORAUT-TKBS-0008-5 Revision No. 03 Effective Date: 08/17/2012 Page 44 of 116

Table 5-9. Air sample results from 1957 

surface plutonium dispersal test.

Location

Long-lived alpha (dpm/m3)

500 ft 1,000 ft

North 498 686

East 200 12.7

South 217 268

West 304 183

REECo collected air samples on the north side of the Decontamination Building from April 24 to April 

30, 1957. Decontamination Building air samples had an activity of less than 32 dpm/m3

. Daily 

surveys and swipes in the Decontamination Building indicated no significant alpha contamination.

REECo (1957) contains a description of the decontamination of an Area 9 balloon test site and 

specific air sampling results from decontamination activities that began on September 3, 1957. 

Radioactive fallout from the Area 2 Smokey test (August 31, 1957) contaminated the Area 9 balloon 

site, where preparations were in progress for a subsequent test. Decontamination in the immediate 

vicinity of the balloon site ground zero was necessary to protect the workers. REECo developed a 

decontamination plan including radiological controls that included the use of heavy equipment to 

remove soil and replace it with clean fill. Personnel included two equipment operators, six truck 

drivers, seven laborers, two foremen, three radiation monitors, and a superintendent. Air samples 

were collected before, during, and after each phase of the decontamination to determine airborne 

activity. The highest concentrations of airborne contamination existed during loader operations 

(5.3 × 104 dpm/m3 beta-gamma at a sampling station approximately 100 ft downwind; the activity was 

extrapolated to collection time). Table 5-10 lists additional air sampling data.

Table 5-10. Additional air sampling data, Area 9 balloon test site.

Condition of area

Beta plus gamma at 

collection time (dpm/m3)

Before start of decontamination, no one in area 1.64E+04

During scraper operation before wetting down 1.33E+04

During scraper operation after wetting down 0.92E+04

During loader operation, very dusty 5.3E+04a

After decontamination, no one in area 1.32E+04

a. Respirators required above 2.2E+04 dpm/m3 at NTS.

Operation Hardtack Phase II On-Site Rad-Safe Report (REECo 1958) listed air sampling reporting 

levels for airborne particulates as 2 × 10-6 μCi/m3 (alpha) and 1 × 10-3 μCi/m3 (beta). Airborne 

particulate material was normally collected on fiber filters using Staplex or Filter Queen sampling 

devices. An annular impactor for alpha in dust was used when immediate analysis was required.

Radiological Safety for Underground Nuclear Explosions (Wilcox and Coogan 1960) described mining 

and drilling operations. REECo performed mining, which consisted mainly of high-explosive blasting, 

removal of broken rock, and reshoring of the reopened tunnel, at the request of LRL. Radioactive 

debris was dumped with the mine tailings. The REECo report stated that nonradioactive material was 

sufficient in the “dump” to prevent accumulation of significant radiation levels. Debris beta-gamma 

ratios during the period (1 to 7 months after the detonation) were variable from 2:1 to 10:1.

Wilcox and Coogan (1960) stated that the liberal application of water and natural water seepage 

reduced airborne radioactivity in the mining work area. Where natural seepage was not enough to 

keep areas moist, airborne radioactive particulate concentrations were occasionally as high as 

4.5 × 10-2 μCi/m3 beta, but airborne concentrations never exceeded 1 × 10-3 μCi/m3 beta in naturally 

moist portions of the tunnel. Natural emitters were excluded by a 5-day decay period or by annular 

impactor sampling. The report noted the presence of high concentrations of airborne radioactive 
Document No. ORAUT-TKBS-0008-5 Revision No. 03 Effective Date: 08/17/2012 Page 45 of 116

material and that accompanying high levels of radon-thoron daughter products made measurements 

difficult.

According to Wilcox and Coogan (1960), core drilling in ground zero areas was performed by the E. J. 

Longyear Drilling Company and REECo at the direction of LRL and LASL. Drilling operations in the 

LRL tunnels used water; drilling in LASL shafts was dry, and a large vacuum system removed cored 

material. In dry drilling operations, filtering the vacuum system exhaust reduced air activity, but 

activity levels occasionally exceeded MPC values, at which point workers wore respiratory protection 

devices. The vacuum system exhaust was downwind from the drillers’ work area when possible. The 

handling of “dried” core samples often resulted in significant airborne contaminants. Air hoods and 

respirators were used to prevent worker exposure to internal radiation.

Re-entry Problems Associated with Radiation from Underground Nuclear Detonations (Brown 1963) 

discussed radiological conditions for atmospheric versus underground testing. This document stated 

that inhalation exposure problems characterized the entire post-test work environment, but that 

radiation monitors controlled most of the hazards. It also stated that there were no external or internal 

exposures during the previous fiscal year (1962).

Hazards to Personnel Re-Entering NTS following Nuclear Reactor Tests (NRDL 1968) noted a 

concern of inhalation of particulate by reentry personnel. The document reported the ejection of 

coarse (greater than 12 μm) and fine (less than 1 μm) particulate to several thousand feet.

Table 5-11 lists examples of historical limits for air samples for years from 1950 to 1987 (REECo 

1985; Author unknown undated c; Geiger and Whittaker 1961). Threshold levels include tolerance, 

MPC, reporting, exposure, detection, alert, and sensitivity levels. Attachment A, Section A.4.4 

discusses respiratory protection and contamination control procedures.

Table 5-11. Historic air sampling limits (Author unknown undated a). 

Year Limit Note

1957 MPC in air above background of mixture of unknown 

radionuclides was 1 E-09 μCi/mL (beta and gamma) 

and 5 E-12 μCi/mL (alpha)

When radionuclides were known, MPC was 3 

times above values based on exposures of 

40-hour week

1958 Air tolerance for uranium was 50 μg/m3 or 

approximately 70 dpm/m3

LANL action level

1958 Laboratory reporting levels for airborne radionuclides 

were:

2 E-06 μCi/m3 (alpha)

1 E-03 μCi/m3 (beta)

1959 MPC in air at NTS above background were:

2 E-12 μCi/mL (alpha)

3 E-9 μCi/mL (beta and gamma)

1961 Sensitivity for I-131 concentration in air was 

approximately 10 pCi/sample

I-131 concentration in air was determined by 

collection on activated charcoal and specific 

gamma counting.

1964 Minimum detection limit for I-131 was <0.00000001E08 μCi/m3

1964 Air sample alert level was 1 E-05 μCi/m3 gross alpha 

and beta.

If gross beta in sample exceeded this level, 

analysis for actinium (Ac-227) was made. 

Radon and thoron samples were collected in 

Area 15 shaft down to 1,000 ft; levels were 

below MPC of 3 E10-8 μCi/cm3 for radon. 

Thoron levels were negligible.

1965 MPC for Pu-239 (in air) was 2 E-12 μCi/cm3 for 40-

hour week
Document No. ORAUT-TKBS-0008-5 Revision No. 03 Effective Date: 08/17/2012 Page 46 of 116

Year Limit Note

1966–

1967

Alert level in air:

1 E-14 μCi/cm3 (alpha)

1 E-11 μCi/cm3 (beta)

Alert levels based on MPC of unknown 

radionuclides in air over 168 hours were 

compared with sample activity results. 

1968 Kr-85 could be measured down to 2 E-07 μCi/cm3

Xe-133 and -135 could be measured down to 8 pCi/m3.

Kr and Xe could be measured with 1-L ion 

chamber used when only one noble gas or 

mixtures of known composition were present.

1970 Alert level for alpha detection in air samples from 

unknown radionuclides was 1 pCi/m3. Detection limit 

in air samples was 0.0019 pCi/m3.

1971 Alert level for radioactivity in air was 15 pCi/m3

1985 Kr-85 LLD in air was 4 pCi/m3

Xenon LLD in air was 8 pCi/m3

5.5 INTERFERENCES AND UNCERTAINTY

5.5.1 Contamination of Samples

Because levels in samples of activity significant in excreta, especially urine, were generally below 

detectability on workplace personnel detectors, contamination of samples from the worker’s hands or 

clothing is a possibility. Hanford found a decrease in detectable plutonium bioassay results after 

switching to home collection. Laboratory contamination and mix-up of samples in the laboratory are 

also a possibility, although laboratory quality control procedures and performance of test samples 

were designed to minimize this source of contamination.

A contaminated sample will probably show up as an obvious outlier in the dataset for a given worker. 

If the dataset shows an unusually high urinalysis result for a radionuclide other than tritium, and if 

follow-up samples were not consistent with the high result, the high result can be considered an 

outlier. However, if the result is not obviously an outlier, it is favorable to claimants to assume the 

result is real.

For in vivo measurements, contamination can occur as external to the body or, in the case of chest 

counting, as external to the lung. If a follow-up in vivo count obtained the same day or within a few 

days shows a dramatic decrease in activity or no detectable activity, external contamination can be 

assumed. Radon progeny and medical diagnostic or therapeutic procedures involving radionuclides 

can cause interference to in vivo measurements, especially for NaI detectors. However, unless the 

count was invalidated or noted as being influenced by such interference, use the results as recorded.

5.5.2 Uncertainty

Uncertainties for bioassay measurements were not stated in the records. For results near or at the 

reporting levels, the assumption provided in the Internal Dose Reconstruction Implementation 

Guideline (NIOSH 2002) should be used for results near or at the reporting levels when no uncertainty 

is reported in the DOE claim file; that is, the standard deviation is 0.3 times the MDA or reporting 

level. For results greater than 3 times the MDA or reporting level, the standard deviation can be 

assumed to be 0.1 times the result, based on quantification level [as cited in NIOSH (2002)]. The 

reported value should be used if actual standard deviations or other indications of error are reported 

with a bioassay measurement result.

5.5.3 Less-Than Values

At NTS, the term less-than refers to data that are reported as less than some reporting level (Allen et 

al. 1993). For example, a plutonium urine bioassay might be reported as <1 E-12 (less than 1 × 10-12) 
Document No. ORAUT-TKBS-0008-5 Revision No. 03 Effective Date: 08/17/2012 Page 47 of 116

μCi/mL or less than the LLD. Less-than data can be used as a constraint on the iteration; the 

predictions of a model should agree with the less-than data. For example, if a model predicts a urine 

concentration of 5 × 10-4 μCi/mL and the measured concentration is less than the LLD, the 

expectation, empirical, and results are in agreement as long as the LLD is higher than the expectation 

value. Less-than data should not be used for residual plots, test runs, or least squares fitting 

procedures.

5.6 DOSE RECONSTRUCTION RECOMMENDATIONS

5.6.1 Evaluation of Potential Internal Worker Exposure

NTS was unique in the DOE weapons complex in that no production processes occurred at this site. 

Chronic exposure would for most workers be based on ambient levels of radiation. Devices were 

assembled from materials produced elsewhere and field tested at NTS. The source term was specific 

to each testing category (atmospheric, tunnel, safety, etc.) and in some cases quantities of specific 

radionuclides and ratios remain classified. Residual soil concentrations at various NTS locations are 

available in the environmental section of this TBD (ORAUT 2010a).

As indicated in this TBD, during atmospheric testing the emphasis was on measurement of external 

radiation. Internal monitoring was based on the worker’s assigned responsibilities and/or the worker 

being identified as potentially contaminated. Therefore, only a worker who was suspected of 

contamination would have received the initial screening provided by nasal swab. Nasal swabs were 

taken on a regular basis, particularly for anyone involved in reentries, decontamination activities, or 

other activities requiring respiratory protection. If a nasal swab result indicated contamination, the 

individual received bioassay (usually urinalysis). For most cases, the DR will review the worker’s 

duties and job history and will base their assignment of internal dose on one or a combination of the 

following:

• The bioassay results in the worker’s DOE records (Attachment A),

• Environmental measurements (ORAUT 2009, 2010a), and/or

• Air sampling limits after 1992 (ORAUT 2005a).

Internal dose is assigned based on current revisions of Project guidance (NIOSH 2002; ORAUT 

2004a, 2005a, 2007).

5.6.2 Workers with Pre-1963 Employment

A class of employees from NTS has been added to the SEC for the period January 27, 1951 through 

December 31, 1962. For some claims it is required that a partial dose reconstruction be conducted. 

As a result of the SEC class for NTS, NIOSH has determined, and the Secretary of Health and 

Human Services has concurred, that in the absence of bioassay results for the worker, internal doses 

cannot be reconstructed between 1951 and 1962 inclusive for an EE. 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. Any bioassay results found in the 

DOE files for NTS workers before 1963 should be assumed to be valid, [HASL-300 (Harley 1976) 

procedures were used in the early bioassay program], therefore these results can be used to evaluate 

internal dose. If internal monitoring exists for an individual in the SEC timeframe, it should be 

assumed to be a nonroutine sample and associated with some discrete event.

Some of the standard techniques used for process facilities in the DOE complex cannot be employed 

at NTS because internal monitoring was limited and exposure was episodic. Based on the bioassay 

results for the worker, both acute and/or chronic intakes should be evaluated and the best fit assigned 

when a best estimate is required. NTS bioassay results would normally be identified with an acute 
Document No. ORAUT-TKBS-0008-5 Revision No. 03 Effective Date: 08/17/2012 Page 48 of 116

event (test or venting) associated with the worker’s employment period, duties, and or access records 

included in the DOE file. However, based on the records for certain workers indicating involvement in 

multiple tests, a chronic exposure may be appropriate.

Only the measured isotope can be used. For example, if there is a urine sample result for plutonium, 

the DR should not apply ratios for uranium and fission products to that intake. Only the plutonium 

should be estimated. If the bioassay report includes other analyses, such as gross gamma, a worstcase gamma emitter can be assumed.

Results from urine, chest counts, fecal, and WBCs found in the worker’s records should be evaluated 

for intakes. Other analyses such as nasal swabs are a qualitative measure used to evaluate the 

potential for exposure. For NTS, the dose assignments based on bioassay results are consistent with 

project guidance (NIOSH 2002; ORAUT 2007).

For miners and other who are identified as working underground (i.e., tritium results found in the case 

records), the results can be evaluated. Radon and/or thoron assignment in accordance with Section 

4.4.4 of the environmental chapter of this TBD (ORAUT 2010a) may also be considered depending on 

the cancer type (Allen 2007).

5.6.3 Workers with Employment From 1963 through 1992

NIOSH has determined, and the Secretary of Health and Human Services has concurred, that NIOSH 

lacks sufficient information that would allow it to adequately estimate internal exposures during the 

period 1963 through 1992. NIOSH believes that the cessation of all nuclear testing, coupled with the 

implementation of the 1993 NTS internal technical basis document that demonstrates NTS 

compliance with 10 CFR Part 835, supports NIOSH’s ability to bound internal dose for the evaluated 

class starting in 1993. Although HHS has added this class to the SEC, NIOSH intends to use any 

available internal and external data for the recommended period (that can be interpreted using 

existing NIOSH processes and/or procedures) for the purpose of partial dose reconstructions.

Some of the standard techniques used for process facilities in the DOE complex cannot be employed 

at NTS because internal monitoring was limited and exposure was episodic. Based on the bioassay 

results for the worker, both acute and/or chronic intakes should be evaluated and the best fit assigned 

when a best estimate is required. NTS bioassay results would normally be identified with an acute 

event (test or venting) associated with the worker’s employment period, duties, and or access records 

included in the DOE file. However, based on the records for certain workers indicating involvement in 

multiple tests, a chronic exposure may be appropriate.

Only the measured isotope can be used. For example, if there is a urine sample result for plutonium, 

the DR should not apply ratios for uranium and fission products to that intake. However, in the event 

that 239plutonium results are identified in the absence of results for 241americium and 238plutonium, the 

absent result shall be accounted for by assuming the ratios to 239plutonium provided in Section 5.2.3. 

If the bioassay report includes other analyses, such as gross gamma, a worst-case gamma emitter 

can be assumed.

Results from urine, chest counts, fecal, and WBCs found in the worker’s records should be evaluated 

for intakes. Other analyses such as nasal swabs are a qualitative measure used to evaluate the 

potential for exposure. For NTS, the dose assignments based on bioassay results are consistent with 

project guidance (NIOSH 2002; ORAUT 2007).

For miners and others who are identified as working underground (i.e., tritium results found in the 

case records), the results can be evaluated. Ambient environmental radon and/or thoron assignment 
Document No. ORAUT-TKBS-0008-5 Revision No. 03 Effective Date: 08/17/2012 Page 49 of 116

in accordance with Section 4.4.4 of the environmental chapter of this TBD (ORAUT 2010a) may also 

be considered depending on the cancer type (Allen 2007).

5.6.4 Approach to Internal Dose Reconstruction for NTS

Starting April 1, 1957, all persons entering NTS were required to wear a film badge with the correct 

monthly color coding (Smith 2003). Because all personnel were issued film badges, and later 

thermoluminescent dosimeters (TLDs), the presence of external monitoring results does not equate to 

the worker entering radiation exclusion (radex) areas (i.e., areas of elevated alpha surface 

contamination measured by portable alpha survey meter). This means any distinction in the internal 

dose assigned has to be made based on the available information about the job duties of the 

individual and/or positive external dose results in the records. For example, you can assign 

environmental dose to workers not directly involved in post-test activities (e.g., cement truck driver, 

cafeteria worker, housekeeper, etc). This can be done starting in 1963.

After 1992, for individuals with job classifications where environmental internal dose assignment is 

not appropriate, the efficiency approach based on air sampling limits may be used to overestimate

cases (ORAUT 2005a) with employment after 1992.

Note: For respiratory cancers (e.g., Lung, ET1, and ET2) when underground work has been 

identified, ambient environmental radon exposures should be applied (whether monitored or 

unmonitored) according to Section 4.4.4 of the environmental section of this TBD (ORAUT 2010a). 

Radon dose is assigned as working level month (WLM) in the radon section of the Interactive 

RadioEpidemiological Program (IREP) input sheet. Ensure that exposure to thoron (220Rn) is also 

addressed. Thoron levels are provided for two locations in ORAUT (2010a). If the individual worked 

in other locations, thoron levels can be assigned using the thoron:radon ratio in these tables (typically 

the thoron:radon ratio is 2:1, other values can be used as determined by the DR). NIOSH (2008) 

provides a method for determining the dose from thoron. This dose is assigned as alpha dose in the 

standard IREP dose section using a lognormal distribution with the same geometric standard 

deviation as radon (entered into the radon section of the IREP sheet). Radon and thoron exposures 

should be prorated based on the estimated time underground and employment periods. This should 

all be discussed in the report.

The DR must be diligent in investigating the type of work the individual performed, where the work 

was performed, and what nuclear testing was being performed while the individual was at the site. 

This information will be used to determine the most appropriate intake type (chronic vs. acute), and 

what radionuclides need to be evaluated (e.g., cesium, strontium, radon, iodine, etc.).

In general as an efficiency method, the assumption of a chronic exposure will often work (especially 

when calculating missed dose), because it will approximate a series of acute intakes (i.e., there were 

enough tests that some individuals could have had an almost constant potential for intakes during 

some years). DRs should review the compiled list of events that occurred during the testing program 

(DOE 2000) to aid in this determination. However, there is still room for judgment, particularly in the 

case of a single positive result, and where there is a higher likelihood of the result being associated 

with a particular test. This information can be compared to the employee’s period of employment to 

determine if the EE was potentially exposed during an event. During events the number of EEs 

associated with the test would have been extremely small.

5.6.4.1 Single Positive Result

Bioassay was not performed on a routine basis for most individuals, it was typically performed when 

an intake for the individual was suspected or the bioassay was specified as part of the Radiation 

Safety Plan written for a specific event. Therefore, it is likely that a positive sample result is 
Document No. ORAUT-TKBS-0008-5 Revision No. 03 Effective Date: 08/17/2012 Page 50 of 116

associated with a recent intake. For a single result that can be associated with a particular event, it is 

best modeled as an acute intake. For NTS, the DR may need to associate the intake with an event if 

there is an indication that the individual was involved in reentry after the test rather than at the time of 

the test itself. In this case, the DR might need to put the exposure date no more than 1 week before

the sample or select the date that best agrees with later negative data. There will often be no followup bioassay.

Note: “Bioassay” includes both in vivo and in vitro measurements. If only a baseline or a termination 

bioassay is included in the records, then the DR can consider the individual to be not monitored.

5.6.4.2 For Individuals with Urine Bioassay Results (Positive or <MDA)

Note: Bioassay results are very limited in the NTS records. Most will be from suspected intakes or a 

termination result. Be aware that often no follow-up bioassay was performed to help in the 

determination of assigning acute or chronic intake. Occupations that did receive routine bioassay by 

the 1960s are security guards, radiation monitors, and individuals included in immediate reentry after 

an event. The NTS Site Description includes a listing of job titles and duties that will help in the 

determination of chronic vs. acute exposures.

For each radionuclide (plutonium, uranium, strontium, tritium, etc.) but not gross gamma, gross alpha, 

or GFPs (gross beta) for which there are urine bioassay results:

• Review all site area(s) where the individual worked using his or her entire work history.

• Select the absorption type (F, M, S, or Super S) most favorable to the claimant for the 

radionuclide and organ of interest based on the work areas or job classification. If this cannot 

be determined, use the absorption type most favorable to the claimant listed for the entire site.

• For 238Pu and, 239Pu <MDA, determine the 239Pu intake based on half of the MDA [standard 

missed dose calculation per ORAUT (2007)] and add intakes of the associated radionuclides, 

238Pu and 241Am. These intakes are based on the NTS soil inventories of these radionuclides 

relative to 239Pu. The intake of 238Pu = 0.91 × 239Pu and 241Am = 0.59 × 239Pu (see Table 4-6 of

ORAUT 2010a). These ratios should also be used for positive 239Pu results when results for 

238Pu and 241Am are <MDA or not available. If lung counter results are provided in conjunction 

with the in vitro results, determine the projected chest count values and compare them to the 

measured results to assure that the calculated intakes are not in disagreement with the 

measured values.

• For uranium results <MDA, calculate missed dose based on MDA for total uranium and then, 

for purposes of dose calculation, assume the calculated intake (in units of activity) is 100% 

234U. Bioassay methods for uranium were developed at NTS, and uranium results may be 

found for some individuals. The type of uranium and the specific activity to apply will vary 

according to the duties of the individual and the timeframe when the individual was on site. 

The IMBA program includes defaults that can be used. For an underestimate, if lung counter 

results are provided in conjunction with the in vitro results, calculate the potential missed 

intake rates based on the lung counter MDAs to ensure that the intakes of 235U and 238U have 

not been overestimated using the urine data.

• Calculate 3

H doses per the technical information bulletin on tritium (ORAUT 2004b). Tritium is 

an issue for individuals in Area 12 B Tunnel (U12b). The presence of tritium in this area was 

identified in 1961 and individuals who worked there have been routinely monitored since then. 

The original source of the tritium was thought to be from 1958 test activities.
Document No. ORAUT-TKBS-0008-5 Revision No. 03 Effective Date: 08/17/2012 Page 51 of 116

For unspecified in vitro bioassay results (alpha, beta, gamma, and fission products):

• For gross alpha results (positive), assume analysis was for 239Pu if no other bioassay results 

specific to plutonium are available.

• For gross beta and GFP results, assume the analysis was for 90Sr unless there is evidence 

that other radionuclides were present. Use the GFP MDAs applied as 90Sr.

• For gross gamma results, assume analysis was for 137Cs unless there is evidence that other 

nuclides are present. Use the gamma MDAs assigned as cesium.

5.6.4.3 For Individuals with Whole-Body Count Results

For WBC results all <MDA or no remarks:

• Use the NTS Chooser tool/aid to determine the nuclide favorable to the claimant based on the 

length of employment and organ identified with the cancer type using half of the published 

MDA.

• Determine the organ dose associated with each organ.

• Use a triangular distribution for missed dose.

• For specified fission products with positive WBCs, use standard project methods (ORAUT 

2007).

5.6.4.4 For Individuals with Lung Count Results

Lung (or chest) counts were typically performed after a suspected intake of thorium, uranium, or a 

TRU element.

• If there are urine sample results for any of the above nuclides, use the lung counts in 

conjunction with the urine sample results to ensure that the two sets of measurements are not 

in disagreement.

• If there are no urine sample results and the results are all <MDA, assume the radionuclide of 

interest was 241Am, used as a tracer for plutonium intakes, unless there is indication of other 

radionuclides. Calculate missed dose using the MDA for 241Am and using an Associated 

Radionuclide of 241Pu in IMBA. This is necessary because 241Am is being produced by the 

decay of 241Pu in the plutonium mixture and this ingrowth must be taken into account. Add 

intakes of associated radionuclides (239Pu and 238Pu). Calculate missed dose.

5.6.5 Examples of the Assignment of Internal Dose for Selected Job Titles

Support workers not directly involved in testing (e.g., cafeteria worker, clerk, housekeeper, bus 

driver, warehouseman, expediter, etc.) and who were not assigned underground duties. The 

assignment of internal dose based on environmental values (ORAUT 2010a) should be an 

overestimate of exposure. Bioassay records above the MDAs in the DOE records should be 

evaluated and included to the internal dose estimate. Claims that include this environmental dose can 

be completed as over or best estimates. 

Workers directly involved with testing [e.g., drill crews or anyone with a job title indicating a fulltime underground work assignment (full shift and a 40-hour workweek)]. ORAUT (2008) contains job 
Document No. ORAUT-TKBS-0008-5 Revision No. 03 Effective Date: 08/17/2012 Page 52 of 116

titles and identifies the potential for tunnel work. For respiratory tract cancers, include radon (ORAUT 

2010a) and thoron (NIOSH 2008) for both over and best estimate cases. If the worker records include 

tritium monitoring, evaluate tritium as discussed in Section 5.1.3.2 and include it for all cancer types. 

Bioassay records found in the worker records should be evaluated and included in the internal dose 

estimate using a method appropriate for underestimate, best estimate or overestimate based on 

project guidance (ORAUT 2007).

Engineers, scientists, supervisors, waste management and safety personnel, and allied crafts. 

These workers could require a combination of environmental and air sampling. Some job categories 

might spend some time in the tunnels and some time in town or in an onsite office. For a favorable to 

claimant assignment of internal dose, DRs should assign these workers environmental dose. Worker 

bioassay records above the MDAs should be evaluated and included in the internal dose estimate. 

Workers who were involved in the assembling of devices and the analysis of debris from events will 

likely have positive external dosimetry results as well as bioassay results.

The records might indicate that workers were underground for several years and then moved to a job 

classification that put them above ground. For example, some RCTs and industrial hygiene personnel 

were assigned full-time underground for their entire term of service. Others were assigned full-time 

underground for limited periods (e.g., a quarter or a year). Still others might have gone underground 

only on special assignment (e.g., reentry support or special monitoring tasks) or to backfill for 

regularly assigned employees. It would be reasonable and favorable to claimants to apply 

environmental internal dose for all years when the length of time the worker actually worked in the 

tunnels cannot be determined.

A reasonable or minimizing approach would be to assign internal dose based on the bioassay results 

for the worker. Both acute and chronic intakes should be evaluated and the best fit assigned. 

Because exposure at NTS was episodic, the bioassay results would normally be identified with an 

acute event (test or venting) associated with the worker’s employment period, duties, and or access 

records included in the DOE file. However, based on the records for certain workers indicating 

involvement in multiple tests, a chronic exposure may be appropriate.

5.6.6 Unusual or Unique Working Environments

The following circumstances indicated in the DOE records will require special treatment or evaluation:

5.6.6.1 Potential for Iodine Exposure

For a worker with thyroid cancer whose records indicate that there was involvement in reentry after a 

test or in a venting or working drill-backs before 1964, DRs should evaluate iodines based on the 

information in Section 5.3.3. From the records for the claim, a reentry time after the event for the 

worker should be estimated and used in the determination of the relative abundances of the iodines. 

Baneberry, the most significant venting to take place at NTS, provides a bounding concentration that 

can be used to develop an iodine dose estimate when the worker received no internal monitoring. 

Based on the maximum reported on site concentration from the Baneberry venting, no significant 

exposure occurred for the unmonitored worker (see Section 5.3.3).

5.6.6.2 Cleanup after Safety Tests

If a DR suspects but cannot confirm intake of plutonium, americium, and thorium, a more sensitive 

urine result from a later period can be used to determine the worst-case intake at a time when 

analysis was not as sensitive. The DR should use MDAs from later urinalyses or, if many samples 

show no detection, one-half the MDA. Plutonium is a consideration for workers on Project 56 (1955 

and 1956) and/or Project 57 (1957), which were plutonium dispersal experiments. Reference to 
Document No. ORAUT-TKBS-0008-5 Revision No. 03 Effective Date: 08/17/2012 Page 53 of 116

“Plutonium Valley” or Area 11 or work at the Nellis Air Force Range during this period is an indication 

of involvement. Partial remediation of the Project 57 area was completed in 1981 and soils 

remediation is included in the ongoing environmental restoration program for both locations. 

Plutonium urine or fecal sampling in the record is a possibility for workers who participated in the 

original tests or cleanup activities up to the present time. Plutonium calculations should also include 

evaluation for Type Super S plutonium (ORAUT 2010c).

5.6.6.3 Nuclear Reactor Development Station

The dose estimation methods and models for nuclear reactor tests in the Naval Research Defense 

Laboratory (NRDL) report (NRDL 1968) have been evaluated for their usefulness and application to 

the NIOSH dose reconstruction process for personnel participating in reentry operations at NTS. The 

dose models in the NRDL report are applicable only to the relatively small population of NTS workers 

that were involved in reentry operation after a nuclear propulsion rocket test at the Nuclear Reactor 

Development Station (NRDS). The data are experiment specific and require knowledge of specific 

physical parameters.

The NRDL report provides methods of computing doses from both individual particles and infinite field 

contamination levels from the nuclear rocket tests. The model provides methods for calculating doses 

to the lungs, GI tract, skin, and gonads for particular exposure scenarios. If a worker who participated 

in NRDS nuclear rocket re-entries has a covered cancer in a location where a hot particle exposure 

has been documented, and the worker can be associated with a specific test, and the requisite 

physical parameters for the test are available from NTS records, DRs may consider using the models 

and methods in the NRDL report. If this information is not available for NRDS workers, internal 

exposures can be addressed through procedures developed for the project (NIOSH 2002) and the 

IMBA EXPERT codes when bioassay results are available in the dosimetry records. If the worker’s 

dosimetry records include dose reconstruction using the NRDL methods, the resulting doses should 

be included in the IREP evaluation.

5.6.6.4 Incidents

Workers directly involved in accidental releases will require internal evaluation if they were in the area 

within 10 days of the release. This information would be in the DOE records. Table A-26 lists 

releases with identified radionuclides. Relative abundances can be determined according to methods 

in ORAUT (2010a). These workers might have been monitored or not monitored for internal 

exposure.

For a worker with a metabolic cancer and bioassay results, DRs should review Attachment A to 

determine what other radionuclides would affect development of the cancer and perform the 

evaluation based on the radionuclides that provided 90% of the dose. Attachment A contains 

information on radionuclides present during test categories and historical MDAs. Based on theoretical 

areal concentrations associated with testing, intakes can be assigned and doses calculated.

5.7 ATTRIBUTIONS AND ANNOTATIONS

All information requiring identification was addressed via references integrated into the reference 

section of this document.
Document No. ORAUT-TKBS-0008-5 Revision No. 03 Effective Date: 08/17/2012 Page 54 of 116

REFERENCES

AEC (U.S. Atomic Energy Commission), 1956, Report of the Test Manager on “56 Project - NTS”, 

Albuquerque Operations Office, Albuquerque, New Mexico, June 12. [SRDB Ref ID: 2493]

AEC (U.S. Atomic Energy Commission), 1961, “Radiological Safety,” Chapter 0524, Standard 

Operating Procedure, Nevada Test Site Organization, NTSO-0524-01, Washington, D.C., 

June 26. [SRDB Ref ID: 12468]

AEC (U.S. Atomic Energy Commission), 1970, Test Manager’s Rad Safe Advisor - Status Report, 

Nevada Operations Office, Las Vegas, Nevada, December 24. [SRDB Ref ID: 9791]

Allen, D., 2007, “NTS Internal Dose,” e-mail to J. Griffin et al., National Institute for Occupational 

Safety and Health, March 20. [SRDB Ref ID: 34077]

Allen, T. E. Jr., E. Zervoudis, C. T. Bastian, and H. F. Kerschner, 1993, Technical Basis for Internal 

Dosimetry at the Nevada Test Site, DOE/NV/10630-64, Reynolds Electrical & Engineering 

Company, Las Vegas, Nevada, June 30. [SRDB Ref ID: 1948]

Arent, L. J., and C. Smith, 2004, “Nevada Test Site (NTS) Technical Basis Document (TBD), Chapter 

5, Occupational Internal Dosimetry - Record of Interview with Ms. Martha DeMarre, Bechtel 

Nevada, memorandum to File, Oak Ridge Associated Universities Team, Oak Ridge, 

Tennessee, August 18. [SRDB Ref ID: 14054]

Author unknown, ca. 1978, Radiological Data Reporting, Mercury, Nevada. [SRDB Ref ID: 39415]

Author unknown, ca. 2004, Radiological Considerations, Nuclear Weapons Testing Source 

Documents. [SRDB Ref ID: 39409]

Author unknown, undated a, NTS General Information 1951–1973. [SRDB Ref ID: 2614]

Author unknown, undated b, Internal Doses. [SRDB Ref ID: 14046]

Author unknown, undated c, General Radiation Protection Considerations, CIC 317286. [SRDB Ref 

ID: 13984]

Bicker, A. E., 1975, “Current Tunnel Survey, Radiological Conditions and Reentry Requirements,”

RRS-75-3, letter with attachments to R. Ray (U.S. Atomic Energy Commission, Nevada 

Operations Office), Reynolds Electrical & Engineering Company, Las Vegas, Nevada, January 

27. [SRDB Ref ID: 8550]

Black, S. C. and G. D. Potter, 1986, “Historical Perspectives on Selected Health and Safety Aspects 

of Nuclear Weapons Testing,” Health Physics, volume 51, number 1, pp. 17–33. [SRDB Ref 

ID: 37777]

Black, S. C., and Y. E. Townsend, editors, 1996, Nevada Test Site Annual Site Environmental Report

for Calendar Year - 1995, DOE/NV/11718-037, Bechtel Nevada, Las Vegas, Nevada, 

September. [SRDB Ref ID: 1791]

BN (Bechtel Nevada), 1998, Technical Basis for Internal Dosimetry at the Nevada Test Site, 

DOE/NV/10630-64, Rev. 0, Change Notice CN-DOE/NV-10630-64-01, Las Vegas, Nevada, 

August 27. [SRDB Ref ID: 39408]
Document No. ORAUT-TKBS-0008-5 Revision No. 03 Effective Date: 08/17/2012 Page 55 of 116

BN (Bechtel Nevada), 2000, Technical Basis for Internal Dosimetry at the Nevada Test Site,

DOE/NV/10630-64, Rev. 3, Las Vegas, Nevada, January 6. [SRDB Ref ID: 1961]

Brown, B. L., 1963, Reentry Problems Associated with Radiation from Underground Nuclear 

Detonations, RRS-63-10, Reynolds Electrical & Engineering Company, Las Vegas, Nevada. 

[SRDB Ref ID: 8054]

Bybee, R. S., 1957, “Entry of Saturn Tunnel, August 30, 1957,” memorandum to W. E. Nolan, 

University of California, Lawrence Radiation Laboratory, Livermore, California, September 5. 

[SRDB Ref ID: 7970]

Church, B. W., 1967, “Report on State of the Art of Whole-Body Counting,” memorandum to Manager, 

Radiation Services, Pan American Airways, National Reactor Development Station, Nevada, 

August 4. [SRDB Ref ID: 8055]

DeMarre, M., 2003, “RE: Information Request,” e-mail to L. J. Arent (Oak Ridge Associated 

Universities Team), U.S. Department of Energy, Nevada Operations Office, Las Vegas, 

Nevada, September 17. [SRDB Ref ID: 14053]

DeMarre, M., 2004a, “RE: NTS Code Questions,” e-mail to L. J. Arent (Oak Ridge Associated 

Universities Team), U.S. Department of Energy, Nevada Operations Office, Las Vegas, 

Nevada, August 9. [SRDB Ref ID: 13975]

DeMarre, M., 2004b, “RE: Bioassay Code Question - NIOSH Project,” e-mail to L. J. Arent (Oak 

Ridge Associated Universities Team), U.S. Department of Energy, Nevada Operations Office, 

Las Vegas, Nevada, June 11. [SRDB Ref ID: 14048]

DeMarre, M., 2005, “RE: GFP Bioassay Questions (Sr & Cs),” e-mail to L. J. Arent (Oak Ridge 

Associated Universities Team), U.S. Department of Energy, Nevada Operations Office, Las 

Vegas, Nevada, February 22. [SRDB Ref ID: 39410]

DOE (U.S. Department of Energy), 1988, Radiation Protection for Occupational Workers, DOE 

5480.11, Office of Environment, Safety and Health, Washington, D.C., December 21. [SRDB 

Ref ID: 8073]

DOE (U.S. Department of Energy), 2000, United States Nuclear Tests, July 1945 through September 

1992, DOE/NV--209-REV 15, Nevada Operations Office, Las Vegas, Nevada, December. 

[SRDB Ref ID: 2433]

DOE (U.S. Department of Energy), 2007, “Prime Contracts,” National Nuclear Security Administration, 

Nevada Site Office, Las Vegas, Nevada, June 20, 

http://www.nv.doe.gov/about/business/contracts.htm, accessed August 9, 2007. [SRDB Ref 

ID: 34075]

Dummer, J. E. Jr., 1958, Los Alamos Handbook of Radiation Monitoring, LA-1835 (3d ed.), University 

of California, Los Alamos Scientific Laboratory, Los Alamos, New Mexico, November. [SRDB 

Ref ID: 967]

Eubank, B. F., editor, 1965, Operation Whetstone Onsite Radiological Safety Report, July, 1964 

through June, 1965, NVO-410-24, Reynolds Electrical & Engineering Company, Las Vegas, 

Nevada, January. [SRDB Ref ID: 8053]
Document No. ORAUT-TKBS-0008-5 Revision No. 03 Effective Date: 08/17/2012 Page 56 of 116

French, C. S., and K. W. Skrable, 1995, Evaluation of Protection, Bioassay, and Dose Assessment 

Programs for Internal Radiation Exposures at Nevada Test Site Particularly as Related to 

Three Exposure Situations, September 30. [SRDB Ref ID: 14050]

Geiger, E. L., 1959, “Air Sampling for Radioactive Particulate Material, memorandum/Information 

Bulletin #12 to Rad-Safe Division Personnel, Reynolds Electrical & Engineering Company, Las 

Vegas, Nevada, August 6. [SRDB Ref ID: 7830]

Geiger, E. L., and E. L. Whittaker, 1961, Analytical Procedures of the Radiological Safety Laboratory, 

Nevada Test Site, Reynolds Electrical & Engineering Company, Las Vegas, Nevada, April. 

[SRDB Ref ID: 8023]

Glasstone, S., 1971, Public Safety & Underground Nuclear Detonations, TID-25708, U.S. Atomic 

Energy Commission, Washington, D.C., June. [SRDB Ref ID: 7929]

Harley, J. H., editor, 1976, HASL Procedures Manual, HASL-300, U.S. Energy Research and 

Development Administration, Health and Safety Laboratory, New York, New York, August. 

[SRDB Ref ID: 31894]

Harris, P. S., C. Lowery, A. Nelson, S. Obermiller, W. J. Ozeroff, and S. E. Weary, 1981, Plumbbob 

Series, 1957, DNA 6005F, JAYCOR, Alexandria, Virginia, September 15. [SRDB Ref ID: 

1591]

Hartzell, L. J., 1962, “Quantities of Samples Required for Routine Laboratory Analysis,” memorandum 

to All Supervisors and Monitors, Reynolds Electrical & Engineering Company, Las Vegas 

Nevada, August 28. [SRDB Ref ID: 8047]

Hawthorne, H. A., editor, 1979, Compilation of Local Fallout Data from Test Detonations 1945-1962 

Extracted from DASA 1251, Volume I-Continental U.S. Tests, DNA 1251-1 EX, General 

Electric Company-TEMPO, Santa Barbara, California, May 1. [SRDB Ref ID: 2578]

HNS (Helgeson Nuclear Services), 1967, Report of Whole Body Counting for Humboldt Bay Nuclear 

Power Plant of Pacific Gas & Electric Co., Pleasanton, California, January 20. [SRDB Ref ID: 

14051]

Holland, J. Z., 1964, “Physical Origin and Dispersion of Radioiodine – A Review,” Biology of 

Radioiodine, L. K. Bustad, editor, Pergamon Press, Oxford, England, pp. 15–23.

HPS (Health Physics Society), 1996, An American National Standard, Performance Criteria for 

Radiobioassay, ANSI N13.30-1996, American National Standards Institute, New York, New 

York. [SRDB Ref ID: 23441]

IAEA (International Atomic Energy Agency), 1970, Directory of Whole-Body Radioactivity Monitors, 

Vienna, Austria.

ICRP (International Commission on Radiation Protection), 1959, Report of Committee II on 

Permissible Dose for Internal Radiation, Publication 2, Pergamon Press, Oxford, England. 

[SRDB Ref ID: 11106]

ICRP (International Commission on Radiation Protection), 1979, Limits for Intakes of Radionuclides 

by Workers, Publication 30, Part 1, Pergamon Press, Oxford, England. [SRDB Ref ID: 29220]
Document No. ORAUT-TKBS-0008-5 Revision No. 03 Effective Date: 08/17/2012 Page 57 of 116

ICRP (International Commission on Radiation Protection), 1994, Human Respiratory Tract Model for 

Radiological Protection, Publication 66, Pergamon Press, Oxford, England. [SRDB Ref ID: 

22732]

Johnson, E. B., 1962, “Employees, NTS, Mercury, Nevada – Personnel Exposures,” memorandum to 

N. E. Woodruff, U.S. Atomic Energy Commission, Nevada Operations Office, Las Vegas, 

Nevada, January 5. [SRDB Ref ID: 8021]

Johnson, W. H., J. Cramer, B. Thompson, and W. Sitko, 1966, Postshot Drilling Handbook, UCID15034, University of California, Lawrence Radiation Laboratory, Livermore, California, 

October. [SRDB Ref ID: 39413]

Kathren, R. L., 1964, “Nucleonics Data Sheet, Activity and Thyroid Dose from Radioiodines,” 

Nucleonics, volume 22, number 11, p. 60. [SRDB Ref ID: 8020]

LASL (Los Alamos Scientific Laboratory), 1952, Operation Ranger, Nevada Test Site, Jan. - Feb. 

1951, Volume 4, Program Reports—Gross Weapons Measurements, WT-201, University of 

California, Los Alamos, New Mexico, June. [SRDB Ref ID: 8036]

Leavitt, M. O., 2006a, HHS Designation of Additional Members of the Special Exposure Cohort under 

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]

Leavitt, M. O., 2006b, HHS Designation of Additional Members of the Special Exposure Cohort under 

the Energy Employees Occupational Illness Compensation Program Act, Designating a Class 

of Employees from Pacific Proving Grounds, Enewetak Atoll, U.S. Department of Health and 

Human Services, Office of the Secretary, Washington, D.C., June 26. [SRDB Ref ID: 39412]

LRL (Lawrence Radiation Laboratory), 1961, General Re-Entry Procedure for Underground Nuclear 

Events, CN-294, University of California, Livermore, California, November 9. [SRDB Ref ID: 

8038]

LRL (Lawrence Radiation Laboratory), 1965, Palanquin Reentry and Recovery Safety Procedure, 

University of California, Livermore, California. [SRDB Ref ID: 8037]

Lyons, C. L., 1992, Operational Air Sampling Report, DOE/NV/10630-42, Reynolds Electrical & 

Engineering Company, Las Vegas, Nevada, September. [SRDB Ref ID: 8575]

McClelland, J., 1955, Analytical Procedures of the Industrial Hygiene Group, LA-1858, University of 

California, Los Alamos Scientific Laboratory, Los Alamos, New Mexico, August. [SRDB Ref 

ID: 13981

McClelland, J., 1958, Analytical Procedures of the Industrial Hygiene Group, LA-1858 (2nd edition), 

University of California, Los Alamos Scientific Laboratory, Los Alamos, New Mexico, 

September 1. [SRDB Ref ID: 12948]

McClendon, L., and B. F. Eubank, editors, 1964, Operation Storax On-Site Radiological Safety 

Report, July 1962 through June 1963, NVO-162-14, Reynolds Electrical & Engineering 

Company, Nevada Test Site, Mercury, Nevada, September. [SRDB Ref ID: 16657]
Document No. ORAUT-TKBS-0008-5 Revision No. 03 Effective Date: 08/17/2012 Page 58 of 116

McMahan, M., and A. G. Ogurek, 2003, Technical Basis for Internal Dosimetry at the Nevada Test 

Site, TBD-E211-002, Rev. 5, Bechtel National, Las Vegas, Nevada, April. [SRDB Ref ID: 

13987]

MJW (MJW Corporation), 2003, Uranium Background Levels in Non-Occupationally Exposed 

Populations Residing Near the Nevada Test Site, Mercury, Nevada, Williamsville, New York, 

June 2. [SRDB Ref ID: 13985]

NBS (National Bureau of Standards), 1953, Maximum Permissible Amounts of Radioisotopes in the 

Human Body and Maximum Permissible Concentrations in Air and Water,” Handbook 52, U.S. 

Department of Commerce, Washington, D.C., March 20. [SRDB Ref ID: 12332]

NBS (National Bureau of Standards), 1959, Maximum Permissible Body Burdens and Maximum 

Permissible Concentrations of Radionuclides in Air and Water for Occupational Exposure, 

Handbook 69, Washington, D.C., June 5. [SRDB Ref ID: 11110]

NCI (National Cancer Institute), 1997, Estimated Exposures and Thyroid Doses Received by the 

American People from Iodine-131 in Fallout Following Nevada Atmospheric Nuclear Bomb 

Tests, Publication No. 97-4264, U.S. Department of Health and Human Services, National 

Institutes of Health, Bethesda, Maryland.

NCRP (National Council on Radiation Protection and Measurements) 1988, Exposure of the 

Population in the United States and Canada from Natural Background Radiation, Report 94, 

Bethesda, Maryland. [SRDB Ref ID: 22736]

Nielsen, D. E., 1957, “Saturn Tunnel Re-Entry,” memorandum to G. W. Johnson, University of 

California, Lawrence Radiation Laboratory, Livermore, California, December 12. [SRDB Ref 

ID: 7968]

NIOSH (National Institute for Occupational Safety and Health), 2002, Internal Dose Reconstruction 

Implementation Guideline, OCAS-IG-002, Rev. 0, Office of Compensation Analysis and 

Support, Cincinnati, Ohio, August 1.

NIOSH (National Institute for Occupational Safety and Health), 2006, SEC Petition Evaluation Report, 

Petition SEC-00055, Rev. 0, Office of Compensation Analysis and Support, Cincinnati, Ohio, 

April 10.

NIOSH (National Institute for Occupational Safety and Health), 2007, SEC Petition Evaluation Report, 

Petition SEC-00084, Office of Compensation Analysis and Support, Cincinnati, Ohio, 

September 25.

NIOSH (National Institute for Occupational Safety and Health), 2008, Lung Dose Conversion Factor 

for Thoron WLM, OCAS-TIB-0011, Rev. 3, Office of Compensation Analysis and Support, 

Cincinnati, Ohio, January 28.

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

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, November.

NIOSH (National Institute for Occupational Safety and Health), 2010b, SEC Petition Evaluation 

Report, Petition SEC-00084, Rev. 1, Office of Compensation Analysis and Support, Cincinnati, 

Ohio, January 25.
Document No. ORAUT-TKBS-0008-5 Revision No. 03 Effective Date: 08/17/2012 Page 59 of 116

NRC (National Research Council), 2003, A Review of the Dose Reconstruction Program of the 

Defense Threat Reduction Agency, National Academies Press, Washington, D.C. [SRDB Ref 

ID: 17579]

NRDL (Naval Radiological Defense Laboratory), 1968, Hazards to Personnel Re-Entering the Nevada 

Test Site following Nuclear Reactor Tests, NRDL-TR-68-149, San Francisco, California, 

December 19. [SRDB Ref ID: 8033]

Ogurek, A., 2004, “Fw: NIOSH Dose Reconstruction Project,” e-mail to L. J. Arent (Oak Ridge 

Association Universities Team), July 11. [SRDB Ref ID: 39414]

ORAUT (Oak Ridge Associated Universities Team), 2004a, Technical Information Bulletin –

Assignment of Environmental Internal Doses for Employees Not Exposed to Airborne 

Radionuclides in the Workplace, ORAUT-OTIB-0014, Rev. 00, Oak Ridge, Tennessee, June 

22.

ORAUT (Oak Ridge Associated Universities Team), 2004b, Technical Information Bulletin: Tritium 

Calculated and Missed Dose Estimates, ORAUT-OTIB-0011, Rev. 00, Oak Ridge, Tennessee, 

June 29.

ORAUT (Oak Ridge Associated Universities Team), 2005a, Internal Dose Overestimates for Facilities 

with Air Sampling Programs, ORAUT-OTIB-0018, Rev. 01, Oak Ridge, Tennessee, August 9.

ORAUT (Oak Ridge Associated Universities Team), 2005b, Interpretation of Dosimetry Data for 

Assignment of Shallow Dose, ORAUT-OTIB-0017, Rev. 01, Oak Ridge, Tennessee, October 

11.

ORAUT (Oak Ridge Associated Universities Team), 2005c, Application of Internal Doses Based on 

Claimant-Favorable Assumptions for Processing as Best Estimates, ORAUT-OTIB-0033, Rev. 

00, Oak Ridge, Tennessee, April 20.

ORAUT (Oak Ridge Associated Universities Team), 2007, Internal Dose Reconstruction, ORAUTOTIB-0060, Rev. 00, Oak Ridge, Tennessee, February 6.

ORAUT (Oak Ridge Associated Universities Team), 2008, Nevada Test Site – Site Description, 

ORAUT-TKBS-0008-2, Rev. 01 PC-1, Oak Ridge, Tennessee, May 27.

ORAUT (Oak Ridge Associated Universities Team), 2009, Los Alamos National Laboratory –

Occupational Internal Dose, ORAUT-TKBS-0010-5, Rev. 01, Oak Ridge, Tennessee, October 

15.

ORAUT (Oak Ridge Associated Universities Team), 2010a, Nevada Test Site – Occupational 

Environmental Dose, ORAUT-TKBS-0008-4, Rev. 02, Oak Ridge, Tennessee, April 15.

ORAUT (Oak Ridge Associated Universities Team), 2010b, Nevada Test Site – Occupational External 

Dose, ORAUT-TKBS-0008-6, Rev. 02, Oak Ridge, Tennessee, April 15.

ORAUT (Oak Ridge Associated Universities Team), 2010c, Estimating Doses for Plutonium Strongly 

Retained in the Lung, ORAUT-OTIB-0049, Rev. 01 PC-1, Oak Ridge, Tennessee, November 

29.

ORDOSE (ORNL Center for Biokinetic and Dosimetric Research), 2003, Rad Toolbox v.1.0.0, Oak 

Ridge National Laboratory, Oak Ridge, Tennessee. [SRDB Ref ID: 26858]
Document No. ORAUT-TKBS-0008-5 Revision No. 03 Effective Date: 08/17/2012 Page 60 of 116

OTA (Office of Technology Assessment), 1989, The Containment of Underground Nuclear 

Explosions, OTA-ISC-414, U.S. Congress, Washington, D.C., October. [SRDB Ref ID: 8029]

Ponton, J., S. Rohrer, C. Maag, R. Shepanek, and I. Grivitis, 1982, Operation Hardtack II, 1958, DNA 

6026F, JRB Associates, McLean, Virginia, December 3. [SRDB Ref ID: 1634]

REECo (Reynolds Electrical & Engineering Company), 1957, Operation Plumbbob On-Site 

Radiological Safety Support Report, OTO-57-2, Nevada Test Site, Mercury, Nevada, July. 

[SRDB Ref ID: 8050]

REECo (Reynolds Electrical & Engineering Company), 1958, Operation Hardtack – Phase II, On-Site 

Radiological Safety Report, OTO 58-5, Nevada Test Site, Mercury, Nevada. [SRDB Ref ID: 

2503]

REECo (Reynolds Electrical & Engineering Company), 1961, Standard Operating Procedures for the 

Radiological Safety Division, Nevada Test Site, Mercury, Nevada, January. [SRDB Ref ID: 

8044]

REECo (Reynolds Electrical & Engineering Company), 1965, Report of Investigation of Thyroid 

Overexposure - U3ct, Type “B” Incident, Las Vegas, Nevada. [SRDB Ref ID: 8049, p. 6]

REECo (Reynolds Electrical & Engineering Company), 1968–1987a, collection of “Section 25 [and 

Section L1], Laboratory Instruments and Equipment,” REECo Standard Procedures, Las 

Vegas, Nevada. [SRDB Ref ID: 2115

REECo (Reynolds Electrical & Engineering Company), 1968–1987b, collection of “Section 28 [and 

Section L4], Instrumental Radiological Analysis,” REECO Standard Procedures, Las Vegas, 

Nevada. [SRDB Ref ID: 2105]

REECo (Reynolds Electrical & Engineering Company), 1969–1987, collection of “Section 27 [and 

Section L4], Chemical and Radiological Analysis,” REECo Standard Procedures, Las Vegas, 

Nevada. [SRDB Ref ID: 2116]

REECo (Reynolds Electrical & Engineering Company), 1970–1987, collection of “Section 26 [and 

Section L2], Sample Preparation,” REECo Standard Procedures, Las Vegas, Nevada. [SRDB 

Ref ID: 2118]

REECo (Reynolds Electrical & Engineering Company), 1973, Onsite Environmental Sciences 

Activities During the Baneberry Event, NVO-410-29, Las Vegas, Nevada, June. [SRDB Ref 

ID: 8045]

REECo (Reynolds Electrical & Engineering Company), 1976, Operation Latchkey OnSite Radiological 

Safety Report, July, 1966, through June, 1967, NVO-410-5, Las Vegas, Nevada. [SRDB Ref 

ID: 8826]

REECo (Reynolds Electrical & Engineering Company), 1978, “Section L5, Personnel Dosimetry and 

Records,” REECo Standard Procedures, Las Vegas, Nevada. [SRDB Ref ID: 2021]

REECo (Reynolds Electrical & Engineering Company), 1979, “Section L5, Personnel Dosimetry and 

Records,” REECo Standard Procedures, Las Vegas, Nevada. [SRDB Ref ID: 2022]
Document No. ORAUT-TKBS-0008-5 Revision No. 03 Effective Date: 08/17/2012 Page 61 of 116

REECo (Reynolds Electrical & Engineering Company), ca. 1980s, Urine Analysis for Radioactivity, 

Purpose and Alert Level, Las Vegas, Nevada (ca. 1980s per M. DeMarre). [SRDB Ref ID: 

39416]

REECo (Reynolds Electrical & Engineering Company), 1981, “Section L5, Personnel Dosimetry and 

Records,” REECo Standard Procedures, Las Vegas, Nevada. [SRDB Ref ID: 2028]

REECo (Reynolds Electrical & Engineering Company), 1983, “Section L5, Personnel Dosimetry and 

Records,” REECo Standard Procedures, Las Vegas, Nevada. [SRDB Ref ID: 2031]

REECo (Reynolds Electrical & Engineering Company), 1985, “Determination of Radiokrypton and 

Radioxenon in Air,” Section 7, Environmental Sciences Department Detailed Operating 

Procedures, Las Vegas, Nevada, September. [SRDB Ref ID: 13986]

REECo (Reynolds Electrical & Engineering Company), 1987, “Section L5, Personnel Dosimetry and 

Records,” REECo Standard Procedures, Las Vegas, Nevada. [SRDB Ref ID: 2045]

REECo (Reynolds Electrical & Engineering Company), undated, Iodine 1960–1963, Las Vegas, 

Nevada. [SRDB Ref ID: 1479]

Reeves, J. E., 1962, “Procedures to Limit Radiation Exposure of Miners at NTS,” memorandum to A. 

R. Luedecke, U.S. Atomic Energy Commission, Nevada Operations Office, Las Vegas, 

Nevada. [SRDB Ref ID: 8042]

Rich, B. L., S. L. Hinnefeld, C. R. Lagerquist, W. G. Mansfeld, L. H. Munson, E. R., Wagner, and E. J. 

Vallario, 1988, Health Physics Manual of Good Practices for Uranium Facilities, EGG-2530, 

EG&G Idaho, Idaho National Engineering Laboratory, Idaho Falls, Idaho, June. [SRDB Ref ID: 

1781]

Robison, W. L., and V. E. Noshkin, 1998, Radionuclide Characterization and Associated Dose from 

Long-Lived Radionuclides in Close-In Fallout Delivered to the Marine Environments at Bikini 

and Enewetak Atolls, UCRL-JC-130230, University of California, Lawrence Livermore National 

Laboratory, Livermore California, September. [SRDB Rev ID: 37747]

Robison, W. L., V. E. Noshkin, T. F. Hamilton, C. L. Conrado, and K. T. Bogen, 2001, An Assessment 

of the Current Day Impact of Various Materials Associated with the U.S. Nuclear Test Program 

in the Marshall Islands, UCRL-LR-143980, University of California, Lawrence Livermore 

National Laboratory, Livermore California, May 1. [SRDB Rev ID: 37748]

Rollins, E. M., 2004, “Record of Telephone Conversation with Ms. Martha DeMarre, Bechtel Nevada, 

Tritium Missed Dose Assignment for the Nevada Test Site,” memorandum to File, Oak Ridge 

Associated Universities Team, Cincinnati, Ohio, November 17. [SRDB Ref ID: 39418]

SAIC (Science Applications International Corporation), 1985, FIIDOS—A Computer Code for the 

Computation of Fallout Inhalation and Ingestion to Organs, Computer User’s Guide, DNA-TR84-375, McLean, Virginia, December 12. [SRDB Ref ID: 39419]

Schoengold, C. R., M. E. DeMarre, and E. M. Kirkwood, 1996, Radiological Effluents Released from 

U.S. Continental Tests 1961 through 1992, DOE/NV-317 (Rev. 1), U.S. Department of Energy, 

Nevada Operations Office, Las Vegas, Nevada, August. [SRDB Ref ID: 2435]
Document No. ORAUT-TKBS-0008-5 Revision No. 03 Effective Date: 08/17/2012 Page 62 of 116

Smith, C. Y., 2007, “Meeting in Las Vegas with Martha DeMarre and Tom Hayes,” documented 

communication on September 8-9, 2003 with M. DeMarre (National Securities Technologies), 

Oak Ridge Associated Universities Team, Oak Ridge, Tennessee, August 13. [SRDB Ref ID: 

34088]

Smith, F., D. W. Boddy, and M. Goldman, 1952, Operation JANGLE, Project 2.7, Biological Injury 

from Particle Inhalation, WT-396, Public Health Service, National Institutes of Health, 

Bethesda, Maryland, June18. [SRDB Ref ID: 7818]

Taplin, G. V., O. M. Meredith, Jr., and H. Kade, 1958, Operation TEAPOT, Nevada Test Site, 

February – May 1955, Project 37.3, Evaluation of the Acute Inhalation Hazard from 

Radioactive Fall-Out Materials by Analysis of Results from Field Operations and Controlled 

Inhalation Studies in the Laboratory, WT-1172, University of California at Los Angeles, Los 

Angeles, California, February 28. [SRDB Ref ID: 7823]

Teasdale, C., 1985, Whole Body Counting Facility, Nevada Test Site, Reynolds Electrical & 

Engineering Company, Las Vegas, Nevada, November. [SRDB Ref ID: 14045]

Wilcox, F. W., 1958, “Decontamination of Tunnel U-12f,” memorandum to W. S. Johnson, Reynolds 

Electrical & Engineering Company, Las Vegas, Nevada, September 27. [SRDB Ref ID: 8030]

Wilcox, F. W., 1959, “Operational Guides for Above-Ground Drill Sites into Ground Zero Areas,” 

memorandum/Information Bulletin #24 to Division Personnel, Reynolds Electrical & 

Engineering Company, Las Vegas, Nevada, November 13. [SRDB Ref ID: 7832]

Wilcox, F. W., 1962, “Operation Guides for Tunnel Areas,” memorandum/Information Bulletin #20 

(Revised) to Division Personnel, Reynolds Electrical & Engineering Company, Las Vegas, 

Nevada, July 10. [SRDB Ref ID: 8040]

Wilcox, F. W., and J. S. Coogan, 1959, Atomic Weapon Accidents – Alpha Emitter Dispersal and 

Decontamination, RSS-59-1, Reynolds Electrical & Engineering Company, Nevada Test Site, 

Mercury, Nevada. [SRDB Ref ID: 8026]

Wilcox, F. W., and J. S. Coogan, 1960, Radiological Safety for Underground Nuclear Explosions, 

RRS-60-3, Reynolds Electrical & Engineering Company, Las Vegas, Nevada, June 9. [SRDB 

Ref ID: 8043]
Document No. ORAUT-TKBS-0008-5 Revision No. 03 Effective Date: 08/17/2012 Page 63 of 116

GLOSSARY

alert level

A urinalysis result that calls the attention of the observer to the fact that further action related 

to the sample or the person from whom it came is required. The action might consist of a 

more informative level of analysis of the same sample, a request for additional samples, or a 

recommendation of in vivo counting of the person involved. The following are alert levels for 

specific analyses:

• gamma – The alert level for routine quarterly urine samples was “any detectable” for

91Y, 103Ru, 106Ru106Rh, 141Ce, 144Ce144Pr, and 154Eu. Alert levels for all other gamma 

emitters were activity concentrations indicative of body burdens that are 10% of the 

MPBBs for continuous exposure given in Health Physics, volume 3, “Report of ICRP 

Committee II on Permissible Dose for Internal Radiation (ICRP 1959),” June 1960.

• tritium – The alert level for a single tritium sample was 1.53 × 10-2 μCi/cm3

. The dose 

“to infinity” to the fat-free soft tissue of the whole body indicated by a single exposure 

resulting in this concentration was 50 mrem.

• GFP – The alert level was 10-7 μCi/cm3

.

• Plutonium (or other transuranic elements) – The alert level was any detectable level.

• Strontium – The alert level was any detectable level.

atmospheric

A test conducted above ground or above water (i.e., in the open air). The last atmospheric 

test at NTS was in 1962.

containment failure

Unintentional release of radioactive material to the atmosphere due to a failure of the 

containment system. Called venting if it is a prompt massive release or seep if it is a slow 

small release that occurs soon after the test.

controlled release

A planned, filtered release frequently performed to reduce airborne radiation levels in the 

tunnel working environment.

controlled tunnel purging

An intentional release to allow recovery of experimental data and equipment or reuse of part of 

the tunnel system.

crater

A nuclear device placed shallow enough underground to produce a throw-out of earth when 

exploded.

detonation

A single nuclear device explosion; one or more comprise a test.

drill-back

Directional drilling operation performed after a shaft test to sample fission product materials in 

the test cavity.
Document No. ORAUT-TKBS-0008-5 Revision No. 03 Effective Date: 08/17/2012 Page 64 of 116

gas sampling

Operation usually performed after test activities have ceased to determine levels of noble 

gases present.

gross beta

Measure of the total beta activity.

H

Hour Detonation time (zero hour), the time the device was detonated.

hole

Alphanumeric designation given to the nuclear weapon test/detonation conducted at NTS. 

The alpha designator refers to the type of location (i.e., S = surface, U = underground). The 

numeric designator refers to the geographical area of the site. The remaining letters refer to 

the sequential ordering of the test/detonation. For most atmospheric tests, only the NTS 

geographical area was given.

kiloton (kt)

The energy of a nuclear explosion that is equivalent to the explosive power of 1,000 tons of 

TNT.

late-time seeps

[1] Small releases that occur days or weeks after a test when gases diffuse through pore 

spaces of the overlying rock and are drawn to the surface by decreases in atmospheric 

pressure. [2] A slow release of gases from test detonation sites that could exist from a few 

hours to even weeks after all other operations in the area have ceased (Schoengold, DeMarre, 

and Kirkwood 1996).

lower limits of detection (LLD, circa 1980s)

[1] The smallest amount of a sample activity that will be reported positive with a specified 

degree of confidence (95%). Before January 1, 1978, LLD = 2 so. After January 1, 1978, 

LLD = 3.29 so. The term so is the estimated standard error for the net sample activity (before 

1993). [2] A value selected above the MDA to reduce the probability of reporting false positive 

results (Allen et al. 1993).

minimum detectable activity (or amount) (MDA)

An a priori value used to evaluate the laboratory’s ability to detect an analyte in a sample 

(McMahan and Ogurek 2003).

monitored worker

A worker who participated in the internal dosimetry program at NTS. All persons entering NTS 

were monitored externally via film badges or thermoluminescent dosimeters.

operational release

Small consequential releases that occur when core or gas samples are collected or when the 

drill-back hole is sealed.

Radex areas

Full Radex Area: radiological contamination is >100 mR/hr (gamma) measured 3 ft from the 

ground, or >104 cpm/55 cm2 (alpha surface contamination measured by portable alpha survey 

meter). No one was allowed into a Full Radex Area unless accompanied by a certified monitor 

who remained with or near them during the entire period.
Document No. ORAUT-TKBS-0008-5 Revision No. 03 Effective Date: 08/17/2012 Page 65 of 116

Limited Radex Area: radiological contamination is ≥ 10 mR/hr but <100 mR/hr (gamma) 

measured 3 ft from the ground, or ≥103 to ≤104 cpm/55 cm2 (alpha surface contamination 

measured by portable alpha survey meter). No one was allowed into a Limited Radex Area 

unless accompanied by a certified monitor who would initially survey the area and return 

periodically to check radiological conditions.

reentry

In the context of NTS operations, the first entry into an area or tunnel (as soon as safety 

limitations permit) immediately after a nuclear detonation. Reentry is under the control and 

responsibility of the Technical Director assigned by the Test Manager. Reentry does not cover 

subsequent cleanup operations or preparing for a future detonation.

reporting level (1993)

The minimum level of a bioassay measurement result, which requires the measurement lab to 

provide prompt notification to Dosimetry (Allen et al. 1993).

safety experiment

[1] Experiment designed to confirm a nuclear explosion will not occur in case of an accidental 

detonation of the explosive associated with the device. Some safety tests were set off at the 

bottom of unstemmed drilled holes, producing a "Roman candle" effect (a test conducted 

underground in an unstemmed hole to minimize, but not eliminate, the release of radioactivity 

to the atmosphere). Because incandescent gases were released, this was sometimes referred 

to as a "Roman candle" effect. [2] Experiment designed to confirm a nuclear explosion will not 

occur in case of an accidental detonation of the explosive associated with the device 

(Schoengold, DeMarre, and Kirkwood 1996).

seep

Uncontrolled slow release of radioactive material with little or no energy. Seeps are not visible 

and can be detected only by measuring for radiation.

shaft

A drilled or mined vertical hole in which a nuclear device is exploded. The hole depth and 

diameter were dictated by the projected yield and diagnostics. Typical hole depths ranged 

from 600 to 2200 feet, and typical hole diameters ranged from four to 10 feet.

stemming

Placement of materials such as grout, magnetite, bentonite, sand, and other materials to 

contain the nuclear explosion in a tunnel or shaft.

surface

A nuclear device placed on or close to the earth's surface.

test

Either a single underground nuclear explosion conducted at a test site, or two or more 

underground nuclear explosions conducted within an area delineated by a circle having a 

diameter of 2 km and conducted within a total period not to exceed 0.1 second (Threshold 

Test Ban Treaty).

tolerance level

The term for maximum permissible exposure.
Document No. ORAUT-TKBS-0008-5 Revision No. 03 Effective Date: 08/17/2012 Page 66 of 116

ton

A measure of the amount of energy released when a nuclear device explodes stated in terms 

of the quantity of TNT that would produce the same amount of explosive energy. A ton is 

equivalent to 1 ton of TNT.

tower

A nuclear device mounted at the top of a steel or wooden tower and exploded in the 

atmosphere.

tunnel

A nuclear device exploded at the end of a long horizontal drift mined into a mountain or mesa 

in such a way that the burst point is placed deep within the earth.

uncontrolled release

A spontaneous release occurring after a test but before post-test drilling operations 

commence. Schoengold, DeMarre, and Kirkwood (1996) uses this term when referring to 

tunnel tests.

underground

Underground nuclear test conducted in a tunnel or at the bottom of a drilled hole or shaft. 

Some underground nuclear tests were not designed to contain all radioactivity (e.g., cratering 

tests or safety experiments).

venting

Prompt, massive, 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.

weapons effects

A nuclear test to evaluate civil or military effects of a nuclear detonation on various targets, 

such as military hardware.

weapons related

A nuclear detonation conducted to test a nuclear device intended for a specific type of weapon

system.

yield

The total effective energy released in a nuclear explosion, usually expressed in terms of 

equivalent tonnage of TNT required to produce the same energy release in an explosion.
Document No. ORAUT-TKBS-0008-5 Revision No. 03 Effective Date: 08/17/2012 Page 67 of 116

TABLE OF CONTENTS

SECTION TITLE PAGE

A.1 Bioassay Codes and In Vitro Minimum Detectable Activities and Detection Levels ................ 68

A.1.1 Codes and Special Terminology Used in Bioassay Records....................................... 68

A.1.2 In Vitro Analyses for Individual Radionuclides ............................................................ 73

A.2 In Vivo MDAs and Reporting Practices................................................................................... 78

A.2.1 Whole-Body Counting................................................................................................. 78

A.2.2 Chest Counting........................................................................................................... 82

A.2.3 Thyroid Counts ........................................................................................................... 83

A.3 Area and Associated Radionuclides of Concern..................................................................... 84

A.3.1 Radionuclides of Concern and Specific Bioassay Programs for Facilities ................... 84

A.3.1.1 Drill-Backs (Yucca Flat and Pahute Mesa, or Areas 1-10 and 18-20) ......... 85

A.3.1.2 Reentry and Mine Back (Area-1 & Area-12)................................................ 87

A.3.1.3 Routine Tunnel Operations (Areas 1, 12, 15, and 16)................................. 89

A.3.1.4 Decontamination Facility (Area 6)............................................................... 89

A.3.1.5 Test Treatability Facility (Area 25) .............................................................. 89

A.3.1.6 Atmospheric Weapon Safety Tests (Area 3 and Area 11)........................... 90

A.3.1.7 Atmospheric Weapons Test Areas (all Areas except Area 22, Area 23, 

and Area 27 due to test location or airborne dispersion)............................. 90

A.3.1.8 Low-Level Waste Site (Area 3) ................................................................... 91

A.3.1.9 Low-Level Waste Site (Area 5) ................................................................... 91

A.3.1.10 Radiation Instrument Calibration Facilities (Area 6 and Area 23)................ 91

A.3.1.11 Radiography Operations (Area 23) ............................................................. 92

A.3.1.12 Well Logging Operations (Areas 1–10 and Areas 18–20) ........................... 92

A.3.1.13 Nuclear Explosive and Device Assembly Facilities (Area 6 and 

Area 27)...................................................................................................... 92

A.3.1.14 Nuclear Rocket Development Areas (Area 25 and Area 26) ....................... 92

A.3.1.15 Radioactive Source Storage Areas (Area 6 and Area 23)........................... 92

A.3.1.16 Radiochemistry and Counting Laboratories ................................................ 93

A.3.1.17 Other Facilities Identified After 1992 (the End of Testing)........................... 93

A.3.1.18 Summary Lists of Radionuclides of Concern............................................... 94

A.3.1.19 Atmospheric Testing Era ............................................................................ 95

A.3.2 Incidents..................................................................................................................... 96

A.3.3 Respiratory Protection Practices............................................................................... 110

A.3.4 Historical Practices and Contamination Levels ......................................................... 111

A.4 Reference Tables for Determining Internal Dose.................................................................. 113

The bioassay program monitored internal uptake of radionuclides using both in vitro analysis of urine 

and feces and in vivo monitoring using whole-body and chest counting. Internal monitoring at NTS 

was limited and was used only for specific job categories supporting events and after incidents.

The BN records archive has all available bioassay records for NTS contractors. The archive does not 

include records for national laboratory employees, which are held by the respective laboratories. 

ATTACHMENT A

OCCUPATIONAL INTERNAL DOSE FOR MONITORED WORKERS

Page 1 of 50

ATTACHMENT A

OCCUPATIONAL INTERNAL DOSE FOR MONITORED WORKERS

Page 1 of 50
Document No. ORAUT-TKBS-0008-5 Revision No. 03 Effective Date: 08/17/2012 Page 68 of 116

Records for REECo employees assigned to the Tonopah Test Range (TTR) under “Work for Others”

are held by Sandia National Laboratory.

Internal dosimetry records during the period from 1955 through 1959 include nasal swabs and 

plutonium bioassays performed by LASL. Beginning in 1958, NTS urine bioassays included tritium, 

gross beta, gamma scan, and plutonium. In 1990, fecal sampling was available. Other nuclear 

testing data are included in individual test radiological safety reports, final operation radiological safety 

reports, monitor or RCT logbooks, area access registers, air sampling forms, laboratory request forms 

and follow-up analyses, and meteorological data (DeMarre 2004a,b). The term used in earlier periods 

for RCTs was Radiation or Rad-Safe Monitors.

A.1 BIOASSAY CODES AND IN VITRO MINIMUM DETECTABLE ACTIVITIES AND 

DETECTION LEVELS

A.1.1 Codes and Special Terminology Used in Bioassay Records

Employer codes and job titles for some NTS contractors [REECo, EG&G, H&N, and RSN] are 

available to DRs in the SRDB maintained by the Division of Compensation Analysis and Support of 

the National Institute for Occupational Safety and Health. Other NTS contractors include BN; 

Lockheed Martin Nevada Technologies, Inc.; Johnson Controls Nevada, Inc.; and WSI. Former NTS 

contractors included H&N (1956–1990), FSN (1963–1990), EG&G (1951–1995), REECo (1953–

1995), and RSN (1990–1995). EG&G had offices and shops in Las Vegas, but some EG&G workers 

worked at NTS for extended periods. Some REECo workers were assigned to the TTR; Sandia 

National Laboratory is the custodian of TTR dosimetry records.

The computerized bioassay records contain the codes listed in Tables A-1 through A-5.

These codes were not identified with a particular period. The measurements of alpha, beta, gamma, 

and GFPs were just a "gross measurement" to determine if additional measurements were needed 

(DeMarre 2004a,b). No specific radionuclides were identified. This was a screening tool. For 

example, a proportional counter determined the counts of alpha and beta for nasal swipes. The 

proportional counter does not identify isotopes. Swipes taken for a worker on a safety test (post test), 

were used as an indicator for the presence of plutonium in the sample. For gamma, when a 

multichannel analyzer was used, specific radionuclides could be cited if they were detectable. For 

example ruthenium could occasionally be detected.

The main internal concerns were tritium, iodines, and plutonium (Arent and Smith 2004). If iodine was 

present, it was easily identified. In the case of the YUBA test (1963), even the G-M survey meter 

detected iodine present in workers leaving the controlled area. Because NTS did not have a WBC 

facility at the time, these individuals were sent to Donner Laboratory (California) for a full workup 

(whole body and urine). The iodine was easily detectable in urine and in the whole body.

Due to security concerns, the word "tritium" was not used. The terms “ACTIVITY or “ACT,” “MINT,” 

“EVERGREEN,” and “T” were code words for tritium. “PRODUCT” was a code word for 239Pu. 

Synonyms for uranium included “Oralloy” for enriched uranium, “tuballoy” for natural uranium, and “D38” for depleted uranium. “LT” or “-” in the bioassay records means less than the detection limit. The

ATTACHMENT A

OCCUPATIONAL INTERNAL DOSE FOR MONITORED WORKERS

Page 2 of 50
Document No. ORAUT-TKBS-0008-5 Revision No. 03 Effective Date: 08/17/2012 Page 69 of 116

Table A-1. Codes for analyte (DeMarre 2003).

For heading

“an_desc_co”

For heading

“analyze_de”

01 003-H (tritium)

02 ALPHA

03 BETA

04 GAMMA (gross)

05 GFP

06 239-PU

07 235-U

08 241-AM

09 238-PU

10 PGAMMA (gamma, specific isotopics)

11 239-PU

22 89-SR

23 90-SR

74 CO2 (carbon dioxide)

75 H2O (water)

code “-99” means not detected (Arent and Smith 2004). Table A-2 lists body part codes and 

abbreviations that are specific to different periods, as listed.

Table A-2. Codes for body parts (DeMarre 2003).

For 

heading(a)

“bdy_prt_cd”

For 

heading(b)

“bdy_prt_pr”

For heading

“bdy_prt_df” Period

01 WB Whole body 1945 to present

02 SK Skin (PL 98-542) 1945 to present

03 LH Left hand 1945 to present

04 RH Right hand 1945 to present

05 FA Forearms 1945 to present

06 EX Extremity 1945 to present

07 EY Eye 1945 to present

08 HD Head 1945 to present

09 GN Gonads 1945 to present

10 TH Thyroid (PL 98-542 & 100-321) 1945 to present

11 BO Bone (PL 98-542) 1945 to present

12 GI GI tract 1945 to present

13 MU Muscle 1945 to present

14 LU Lung (PL 98-542) 1945 to present

15 PA Pancreas (PL 98-542 & 100-321) 1945 to present

16 LI Liver (PL 98-542 & 100-321) 1945 to present

17 AG Adrenal gland 1945 to present

18 SP Spleen 1945 to present

19 KI Kidney (PL 98-542) 1945 to present

20 PR Prostate 1945 to present

21 LF Left foot 1945 to present

22 RF Right foot 1945 to present

23 OT Any tissues other than specific target organ 1945 to present

24 RW Right wrist 1945 to present

25 LW Left wrist 1945 to present

ATTACHMENT A

OCCUPATIONAL INTERNAL DOSE FOR MONITORED WORKERS

Page 3 of 50
Document No. ORAUT-TKBS-0008-5 Revision No. 03 Effective Date: 08/17/2012 Page 70 of 116

For 

heading(a)

“bdy_prt_cd”

For 

heading(b)

“bdy_prt_pr”

For heading

“bdy_prt_df” Period

26 WW Wrist 1945 to present

27 LL Lower large intestine 1945–1986

28 UI Upper large intestine 1945–1986

29 SI Small intestine (PL 100-321) 1945–1986

30 SW Stomach (PL 98-542 & 100-321) 1945–1986

31 BS Bone surfaces 1945 to present

32 RM Red marrow (PL 98-542 & 100-321) 1945–1986

33 BR Breast (PL 100-321) 1945–1986

34 BF Breast (female) (PL 98-542) 1945–1986

35 ES Esophagus (PL 98-542 & 100-321) 1945–1986

36 CO Colon (PL 98-542) 1945–1986

37 BL Bladder (PL 98-542) 1945–1986

38 SA Salivary gland (PL 98-542) 1945–1986

39 PH Pharynx (PL 100-321) 1945–1986

40 BI Bile duct (PL 100-321) 1945–1986

41 GA Gall bladder (PL 100-321) 1945–1986

42 LY Lymph gland(PL 100-321) 1945–1986

43 BN Brain 1945–1986

44 LX Larynx 1945–1986

45 MX Maxillary sinus 1945–1986

46 FR Frontal sinus 1945–1986

47 RU Rectum 1945–1986

48 AO All organs 1945–1986

49 HT Heart 1945–1986

50 UR Urethra 1945–1986

51 SC Spinal cord 1945–1986

52 FT Foot 1945–1986

53 AK Ankle 1945–1986

54 HP Hip 1945–1986

a. From 1987 to the present, this heading is “bp_code”.

b. From 1987 to the present, this heading is “bp_name”.

Table A-3 lists codes for internal radionuclides and specific operations that may be found in excreta 

records. The following codes were used as general information. The code 001 is used for nuclear 

weapons testing operations from 1945 to 1962 in the database. These are the one-record-per-page 

printouts. In general 001 is the main code used for 1945 to 1962. This information is added in 

updates for specific test operations where summary annual exposure data were replaced with 

individual dosimeter entries.

Radionuclides were not specifically for the bioassay program. They were generated for the main 

database, principally for 1945–1962 data, so one could distinguish between the reactor program and 

the weapons test program. The codes provided a potential flexibility for the database. The database 

design was developed in 1979 and augmented in 1983 to create a relational database. One of the 

potentials envisioned by designers was to support epidemiological studies with the data (DeMarre 

2004a,b).

ATTACHMENT A

OCCUPATIONAL INTERNAL DOSE FOR MONITORED WORKERS

Page 4 of 50
Document No. ORAUT-TKBS-0008-5 Revision No. 03 Effective Date: 08/17/2012 Page 71 of 116

Table A-3. Codes for radionuclides and sample types (DeMarre 2003).

int_nuc_cd int_nuc_df int_nuc_cd Int_nuc_df

No data available 022 Te-132

001 Detonation fission products 023 Te-134

002 Reactor fission products 024 I - 131

003 Criticality accident 025 I - 132

004 X-ray generator 026 I - 133

005 Accelerator 027 I - 134

006 Reactor neutrons 028 I - 135

007 Neutron generator 029 Xe-133

008 Cyclotron 030 Xe-135

009 Activation products 031 Cs-137

010 Photo neutrons 032 Ba-140

011 TRU elements 033 Th-232

012 H-3 034 U-234

013 Na-24 035 U-235

014 Fe-59 036 U-238

015 Sr-89 037 Pu-236

016 Sr-90 038 Pu-238

017 Sr-y-90 039 Pu-239

018 Nb-95 040 Pu-240

019 Ru-103 041 Am-241

020 Ru-Rh-106 042 Cf-2xx

021 Te-131 043 Pu-239, Am-241

In general, there was no consistent listing of the test name on the bioassay request form that went to 

the laboratory. Sometimes a location identifier was listed (e.g., U12b for B Tunnel). Until 1994, there 

were unannounced tests, the names of which did not appear on the bioassay request forms (DeMarre 

2004a,b).

Table A-4 lists codes for the type of sample to be analyzed, and Table A-5 lists the associated units.

Table A-4. Codes for sample types 

(DeMarre 2003).

sam_typ_co Sample_typ

06 Feces

09 Nasal smear

10 Urine

19 Wound swab

20 Tissue gross

21 Muscle

22 Blood

23 Skin

24 Liver

25 GI tract

26 Thyroid

27 Bone

28 Kidney

29 Milk

ATTACHMENT A

OCCUPATIONAL INTERNAL DOSE FOR MONITORED WORKERS

Page 5 of 50
Document No. ORAUT-TKBS-0008-5 Revision No. 03 Effective Date: 08/17/2012 Page 72 of 116

sam_typ_co Sample_typ

30 Lung

31 Whole body

60 Miscellaneous

Table A-5. Codes for units (DeMarre 2003).

Code Unit Description Code Unit Description

10 dpm disintegrations per minute 45 μCi/m3 microcuries per cubic meter

11 dpm/cm3 dpm per cubic cm 46 μCi/kg microcuries per kilogram

12 dpm/g dpm per gram 50 mCi microcuries

13 dpm/L dpm per liter 51 mCi/cm3 millicuries per cubic centimeter

15 dpm/m3 dpm per cubic meter 52 mCi/g millicuries per gram

16 dpm/kg dpm per kilogram 53 mCi/L millicuries per liter

20 Bq Becquerels 54 mCi/mg millicuries per milligram

21 Bq/cm3 Becquerels per cubic centimeter 55 mCi/m3 millicuries per cubic meter

22 Bq/g Becquerels per gram 56 mCi/kg millicuries per kilogram

23 Bq/L Becquerels per liter 60 cpm counts per minute (c/m)

25 Bq/m3 Becquerels per cubic meter 61 cpm/cm3 c/m per cubic centimeter

26 Bq/kg Becquerels per kilogram 62 cpm/g c/m per gram

30 pCi picocuries 63 cpm/L c/m per liter

31 pCi/cm3 picocuries per cubic centimeter 64 cpm/mg c/m per milligram

32 pCi/g picocuries per gram 65 cpm/m3 c/m per cubic meter

33 pCi/L picocuries per liter 66 cpm/kg c/m per kilogram

35 pCi/m3 picocuries per cubic meter 67 % percent

36 pCi/kg picocuries per kilogram 68 μg/g micrograms per gram

40 μCi microcuries 69 μg/L micrograms per liter

41 μCi/cm3 microcuries per cubic centimeter 70 μg micrograms U (Reg: kinetic 

phosphorescence analysis)

42 μCi/g microcuries per gram 71 μS/cm microsiemens per centimeter

43 μCi/L microcuries per liter 72 mS/cm millisiemens per centimeter

44 μCi/mg microcuries per milligram

NTS maintains a record set called the “dead bioassay database” (Arent and Smith 2004). There are 

no codes in this database that include information from 1955 to 1963. In the microfiche copy of the 

bioassay data, the years are listed as two digits (56, 57, etc.). The main purpose of the dead 

bioassay file/microfiche is to point to the raw data (note, reel, and frame citations). If an individual has 

a “deadbio” record, the microfiche page is followed by a copy of the original data forms for the data 

cited. DRs should review data from the original forms, not the microfiche “deadbio” index.

Table A-6 lists special location terminology found in NTS bioassay records.

Table A-6. Special terminology (ORAUT 2008, 2010b).

Term Explanation

A-23 Mercury, NV

A-24 Las Vegas

A-52 TTR

A-400 A-25 LANL term for NRDS area

A-401 A-25 LLNL term for Pluto/Ramjet area

A-6, CP-1 Test Control Room Building

A-6, CP-2 Radiological Safety (RadSafe) Building and dosimeter issue location

A-23, Building 1000 Badge office outside NTS main gate (Mercury)

ATTACHMENT A

OCCUPATIONAL INTERNAL DOSE FOR MONITORED WORKERS

Page 6 of 50
Document No. ORAUT-TKBS-0008-5 Revision No. 03 Effective Date: 08/17/2012 Page 73 of 116

Term Explanation

A-23, Buildings 111, 

155, and 650

RadSafe/dosimeter issue buildings

Alcove Side excavation in tunnel drift

BP Bypass (type of tunnel drift)

CP Control Point or A-6

Downhole shaft Vertical underground tunnel

Drift Horizontal tunnel

GZ Ground zero

Invert Part of bottom of tunnel made level by adding materials.

LOS Line of sight (drift or pipe)

Portal Tunnel entry point

Shaft designation Example “U20bb”

U = underground, 20 = Area 20, bb = hole number

Tunnel designation Example “U12b.10” 

U = underground, 12 = Area 12, b = B Tunnel, 10 = drift number 10

Tunnels are located in NTS areas (Wilcox and Coogan 1960; ORAUT 2008, 2010b):

• A-1 (U1a Complex, formerly known as the Low-Yield Nuclear Explosive Research Facility 

Complex)

• A-12 [B, C (“Saturn”), D, E, F, G, J, K, N, P, and T Tunnels]

• A-15 (15E Tunnel)

• A-16 (16A Tunnel)

• A-25 (X Tunnel, non-nuclear experiments only)

A.1.2 In Vitro Analyses for Individual Radionuclides

Table A-7 lists limits of detection for urine and fecal analyses. The MDA is an a priori value used to 

evaluate the laboratory’s ability to detect an analyte in a sample. LLD is defined in Allen et al. (1993) 

as a value selected above the MDA to reduce the probability of reporting false positive results. Limit 

of sensitivity is equivalent to the LLD. Detection limit is a general term related to the smallest amount 

of material detectable as a function of the measurement method and instrument background. All urine 

samples were collected over a 24-hour period.

Author unknown (ca. 1978) notes a change in the definition of LLD. The LLD was defined as the 

smallest amount of sample activity that was reported positive with a specified degree of confidence 

(i.e., 95% of the time). The LLD expression for data before January 1, 1978, was:

LLD = 2so (A-1)

where so = the estimated standard error for the net sample activity.

The LLD expression for data changed after January 1, 1978, to:

LLD = 3.29so (A-2)

ATTACHMENT A

OCCUPATIONAL INTERNAL DOSE FOR MONITORED WORKERS

Page 7 of 50
Document No. ORAUT-TKBS-0008-5 Revision No. 

0

3 Effective Date: 08/17/2012 Page 74 of 116

Table A-7. Limits of detection for urine and fecal analysis.

Radionuclide Period Urine MDA (pCi/L)a Fecal MDA (pCi/g)a Source

H-3 1958–1976 5 μCi/L - Arent and Smith 2004

H-3 1977–1987 1E-6 μCi/mL (detection limit)b - REECo 1968–1987a

H-3 1988c–1999d 300 - Allen et al. 1993

H-3 2000-2002 0.001 μCi/L - BN 2000

H-3 2003–present 0.005 μCi/L - McMahan and Ogurek 2003

Na-22 1970–presentd 200 - Allen et al. 1993

Na-24 1970–presentd 100e - Allen et al. 1993

Mn-54 1970–presentd 100 - Allen et al. 1993

Co-57 1970–presentd 100 - Allen et al. 1993

Fe-59 1970–presentd 600 - Allen et al. 1993

Co-60 1970–presentd 50 - Allen et al. 1993

Sr-85 1970–presentd 200 - Allen et al. 1993

Sr-90 1961–1969d 25 pCi/sample (limit of sensitivity)f - Author unknown undated a

Sr-90 1970–1999d 0.8g - Allen et al. 1993

Sr-90 2000–present 1 - BN 2000, McMahan and Ogurek 2003

Zr-95 1970–presentd 600 - Allen et al. 1993

Nb-95 1970–presentd 100 - Allen et al. 1993

Tc-99m 1970–presentd 100 - Allen et al. 1993

Rh-101 1970–presentd 100 - Allen et al. 1993

Rh-102 1970–presentd 100 - Allen et al. 1993

Rh-102m 1970–presentd 200 - Allen et al. 1993

Ru-103 1970–presentd 200 - Allen et al. 1993

Ru-106 1970–presentd 1000 - Allen et al. 1993

Sb-122 1970–presentd 100 - Allen et al. 1993

Sb-124 1970–presentd 100 - Allen et al. 1993

Sb-125 1970–presentd 300 - Allen et al. 1993

I-131 1961–1969d 10 pCi/sample (limit of sensitivity) - Author unknown undated a

I-131 1970–presentd 100 - Allen et al. 1993

I-132 1970–presentd 90 - Allen et al. 1993

Te-132 1970–presentd 100 - Allen et al. 1993

Ba-133 1970–presentd 200 - Allen et al. 1993

Ba-133m 1970–presentd 100 or 400h - Allen et al. 1993

I-133 1970–presentd 100 - Allen et al. 1993

Cs-134 1970–presentd 500 - Allen et al. 1993

ATTACHMENT A

OCCUPATIONAL INTERNAL DOSE FOR MONITORED WORKERS

Page 8 of 50
Document No. ORAUT-TKBS-0008-5 Revision No. 

0

3 Effective Date: 08/17/2012 Page 75 of 116

Radionuclide Period Urine MDA (pCi/L)a Fecal MDA (pCi/g)a Source

I-135 1970–presentd 200 - Allen et al. 1993

Cs-136 1970–presentd 100 - Allen et al. 1993

Cs-137 1961–1969d 10 pCi/sample (limit of sensitivity) - Geiger and Whittaker 1961, Author 

unknown undated a

Cs-137 1970–presentd 100 - Allen et al. 1993

Ce-139 1970–presentd 100 - Allen et al. 1993

Ba-140 1961–1969d 10 pCi/sample (limit of sensitivity)i - Author unknown undated a

Ba-140 1970–presentd 500 - Allen et al. 1993

La-140 1970–presentd 200 - Allen et al. 1993

Ce-141 1970–presentd 200 - Allen et al. 1993

Ce-143 1970–presentd 200 - Allen et al. 1993

Ce-144 1970–presentd 800 - Allen et al. 1993

Nd-147 1970–presentd 400 - Allen et al. 1993

Eu-152 1970–presentd 300 - Allen et al. 1993

Eu-154 1970–presentd 200 - Allen et al. 1993

Eu-155 1970–presentd 500 - Allen et al. 1993

Yb-169 1970–presentd 600 - Allen et al. 1993

Ta-182 1970–presentd 300 - Allen et al. 1993

W-187 1970–presentd 300 - Allen et al. 1993

Ir-192 1970–presentd 100 - Allen et al. 1993

Fission products- GFP 

(beta)

1977–1987 1E-10 μCi/mL (detection limit)b - REECo 1968–1987a

Fission products- GFP 1988c–presentd 3j - Allen et al. 1993

Gross alpha 1968–present - 1E-07 μCi/g (detection 

limit)k

Author unknown undated a

Gamma 1977–1987 5E-8 μCi/mL (detection limit)b - REECo 1968–1987a

Gamma (Cs-137) 1988c–present 100 - BN 2000, McMahan and Ogurek 2003

Ra-226 1958–1969d 0.88 dpm/sample (detection level)l - Dummer 1958

Ra-226 1970–presentd 300 0.4 or 0.04h Allen et al. 1993

Ra-226 2000–present 0.1 pCi/sample - ? BN 2000, McMahan and Ogurek 2003

Th-228 1970–presentd - 0.01 Allen et al. 1993

Th-230 1954–1992d 0.5 dpm/sample (detection level) m - McClelland 1955, Author unknown 

undated a

Th-230 1970–presentd - 0.01 Allen et al. 1993

Th-230 2003–present pCi/sample 0.04 pCi/sample McMahan and Ogurek 2003

ATTACHMENT A

OCCUPATIONAL INTERNAL DOSE FOR MONITORED WORKERS

Page 9 of 50
Document No. ORAUT-TKBS-0008-5 Revision No. 

0

3 Effective Date: 08/17/2012 Page 76 of 116

Radionuclide Period Urine MDA (pCi/L)a Fecal MDA (pCi/g)a Source

Th-232 1993–1999d - 0.01 Allen et al. 1993

Th-232 2000–present 0.02 pCi/sample 0.05 pCi/sample BN 2000, McMahan and Ogurek 2003

U-234 1970–presentd 0.02 - Allen et al. 1993

U-234 2000–present 0.04 pCi/sample 0.04 pCi/sample BN 2000, McMahan and Ogurek 2003

U-235n 1961–1969 0.03 dpm/sample (limit of 

sensitivity)

- REECo 1961, Author unknown undated 

a

U-235 1970–presentd 0.01 0.008 Allen et al. 1993

U-235 2000–present 0.04 pCi/sample 0.04 pCi/sample BN 2000, McMahan and Ogurek 2003

U-238 1970–1999d 0.02 0.008 Allen et al. 1993

U-238 2000–present 0.04 pCi/sample 0.04 pCi/sample BN 2000, McMahan and Ogurek 2003

Elemental U 1970–present 5 μg/Lo - Allen et al. 1993

Pu-238 1982–1987 2E-10 μCi/mL (detection limit)b - REECo 1968–1987a

Pu-238 1988c–1999d 0.01 - Allen et al. 1993

Pu-238 2000–present 0.006 pCi/sample 0.03 pCi/sample BN 2000, McMahan and Ogurek 2003

Np-239 1970–present 400 - Allen et al. 1993

Pu-239p 1954–1957d 2 dpm/24 hr (detection limit)q - McClelland 1955

Pu-239p 1958–1960d 0.05 dpm/sample (detection limit)r Dummer 1958, Author unknown 

undated a

Pu-239p 1961–1976d 0.005 dpm/sample (limit of 

sensitivity)

- REECo 1961, Author unknown undated 

a

Pu-239 1977–1987 5E-11 μCi/mL (detection limit)b - REECo 1968–1987a

Pu-239 1988c-2000d 0.01 (alpha spec)s; 50 (gamma 

spec)s

0.004 (alpha spec) Allen et al. 1993

Pu-239 2000–present 0.006 pCi/sample 0.03 pCi/sample BN 2000, 2003

Pu-240 1970–present 0.01 0.004 Allen et al. 1993

Pu-240 2000–present 0.006 - BN 2000, McMahan and Ogurek 2003

Am-241t 1954–1957d 2 dpm/sample (detection limit)u - McClelland 1955

Am-241t 1958–1981d 0.5 dpm-24-hr sample (LLD)v - Dummer 1958, REECo 1961, Author 

unknown undated a

Am-241 1982–1987 2E-11 μCi/mL (detection limit)b - REECo 1968–1987a

Am-241 1988c–1999d 0.03w 0.03 Allen et al. 1993

ATTACHMENT A

OCCUPATIONAL INTERNAL DOSE FOR MONITORED WORKERS

Page 10 of 50
Document No. ORAUT-TKBS-0008-5 Revision No. 

0

3 Effective Date: 08/17/2012 Page 77 of 116

Radionuclide Period Urine MDA (pCi/L)a Fecal MDA (pCi/g)a Source

Am-241 2000-2002 0.008 pCi/sample 0.03 pCi/sample BN 2000

Am-241 2003–present 0.006 pCi/sample 0.03 pCi/sample McMahan and Ogurek 2003

Cm-244 2000–present 0.008 pCi/sample - BN 2000, McMahan and Ogurek 2003

a. MDA units are pCi/L for urine and pCi/g for feces unless noted otherwise.

b. Reference states detection limit or “less-than” value.

c. The start date for this value was extended back from 1993 to 1988 to account for data gaps.

d. Only the period start date is confirmed (unless otherwise noted); the value is assumed to apply to the remainder of the period.

e. 1.4 × 10-7 μCi/mL for blood sodium radiobioassay.

f. With counter background of 17 cpm, a 60-minute counting period, 25 pCi of Sr-90/sample detected with an accuracy of +10% at 90% confidence level.

g. Method does not differentiate between Sr-89 and Sr-90.

h. Both values are listed in Allen et al. (1993).

i. REECo (1961) states the concentration of La-140 is estimated from the Ba-140 concentration and the history of the sample.

j. Gas flow proportional counting for GFP.

k. Assumed this was a fecal sample based on the units of μCi/g.

l. Reference states approximately 0.88 dpm of radium could be detected with the method if the sample was allowed to come to equilibrium.

m. McClelland (1955) states Th-230 could be detected down to 0.5 dpm.

n. References state uranium in urine. This was assumed to be U-235.

o. Measured by fluorometric analysis.

p. References state plutonium in urine. This was assumed to be Pu-239.

q. Reference states quantities of the order of 2 dpm of plutonium can be determined by the method.

r. Dummer (1958) states the detection limit at the 99% confidence level is approximately 0.05 dpm/sample.

s. Cannot differentiate Pu-239 from Pu-240 with alpha spectrometry; can differentiate at 50 pCi/L with gamma spectrometry; results reported as Pu-239.

t. Original references state americium in urine. This was assumed to be Am-241.

u. Reference states quantities of the order of 2 dpm of americium can be determined by the method. The 1958 revision lowers the value to 0.5 dpm.

v. Author unknown (undated a) states americium could be detected down to 0.5 dpm/sample. (dpm/sample = cpm/efficiency of counter; alpha counter at LASL were 

assumed to have an efficiency of 50%.) McClelland (1958) states for lack of better data, see plutonium (0.5 dpm/24-hr sample lower limit of detection). Alpha 

proportional counting does not separate thorium, plutonium, curium, actinium, and neptunium.

w. Cannot be chemically differentiated from californium, curium, or other isotopes of americium; results are reported as americium.

ATTACHMENT A

OCCUPATIONAL INTERNAL DOSE FOR MONITORED WORKERS

Page 11 of 50
Document No. ORAUT-TKBS-0008-5 Revision No. 03 Effective Date: 08/17/2012 Page 78 of 116

The change accounted for the possibility of a beta (β) error [i.e., to define the situation with 95% 

confidence that, on the average, activity is reported when it is present (1-β), as well as not reported 

when it is not present (1-α)].

All results for samples collected after January 1, 1978, appear in the REECo Sample Tracking Data

Management System as actual values [i.e., as they were calculated regardless of whether they were 

negative or positive (Author unknown ca. 1978)]. Off-station data were reported to two significant 

digits and only minimal detectable concentration (MDC) values were reported for less-than results. 

An MDC value is calculated for every sample and compared to the sample value. If the MDC value 

was greater than or equal to the sample value, an MDC was reported for off-station data. For onstation use, actual values with counting errors were reported to three significant figures on preliminary 

reports. In the data system, less-than values were stored as three significant digits in the counting 

error field. Table A-8 summarizes the method of data reporting.

Table A-8. Radiological data reporting.

Date Result Sample tracking data management system Off station

Pre 1978 <2 so <2 so

>2 so Result ± 2 so (2 significant figures)

Post 1978 <3.29 so Result ± MDC <MDC

>3.29 so Result ± 2 so (3 significant figures) Result ± 2 so (2 significant figures)

A.2 IN VIVO MDAs AND REPORTING PRACTICES

In vivo detection capabilities are discussed below. Error in estimated body content of radionuclides 

tends to increase as the estimate approaches the minimum detection limit of the in vivo measurement. 

The radionuclide content assessed from the results of a whole-body or lung count can be in error by 

as much as 100% or more from the true content of a low-energy, photon-emitting radionuclide. This is 

not significant in most cases because the MDA is a small fraction of the maximum permissible intake.

A.2.1 Whole-Body Counting

The MDA values for 1967 are based on those reported by HNS (1967) for a 480-second count. The 

values reported by HNS are at the 99% confidence level. To facilitate comparison with other MDA 

values reported at the 99% confidence level, the reported values have been multiplied by the ratio of 

kβ at the 95% confidence level (0.645) to the value of kβ at the 99% confidence level (2.326 – 0.71). 

HNS reported values for 60Co, 65Zn, 95Zr95Nb, and 137Cs; the values for the remaining radionuclides 

were derived from the HNS data. A nominal photon MDA of 2.1 nCi is the basis for the 1967 values in 

Table A-7 for photons in the range 100 to 2,000 keV, with correction for the yield of the primary 

measurement photon, Yγ – MDA = 2.1/Yγ nCi.

Specific efficiency data have not been found for the shadow shield counter with a 3-in. by 3-in. NaI 

crystal, so values for 1967 to 1977 using a 3-in. by 3-in. NaI crystal are based on values for a shadow 

shield counter of similar design (UK11.1) published by the IAEA (IAEA 1970). Since the decrease of 

detector efficiency with energy is essentially offset by the decrease in background with energy, a 

nominal MDA of 10 nCi is used for photon emitters in the range 0.1 to 2 MeV, with correction for the 

yield of the primary measurement photon, Yγ – MDA = 10/Yγ nCi.

The counter used during the period from 1977 to 1980 featured use of an 8-in. by 10-in. NaI detector. 

The efficiency reported for this detector is determined from E = 0.0089 × e-0.000175Eγ, where Eγ is in 

ATTACHMENT A

OCCUPATIONAL INTERNAL DOSE FOR MONITORED WORKERS

Page 12 of 50
Document No. ORAUT-TKBS-0008-5 Revision No. 03 Effective Date: 08/17/2012 Page 79 of 116

keV. Figure A-1 shows the efficiency curves reported for that detector, and the 11.5-in. detector 

installed in 1981. Figure A-2 shows the counting efficiency in terms of counts per minute per photon 

nCi. Background values for these detectors were not reported. Background values for 137Cs under 

similar conditions can be found in IAEA (1970). Six counting facilities reported values ranging from 60 

to 135 cpm. Using a value of 135 cpm, a counting time of 1,300 second (21.7 minutes), and an 

efficiency of 7.9 × 10-3 counts per photon, the 662-keV photon MDA is 0.93 nCi. (For 137Cs with a 

yield of 0.851 photon per disintegration, this becomes 1.1 nCi.) The values for 1978 to 1980 in Table 

A-9 are based on a nominal MDA value of 1 photon nCi, with correction for the yield of the primary 

measurement photon, Yγ – MDA = 1/Yγ nCi.

Figure A-1. NaI detector efficiency, counts per photon disintegration.

Figure A-2. NaI detector efficiency, cpm per photon nCi.

A new WBC facility began operation in 1981, with a shielded room, scanning bed, and 11.5-in. by 8-in. 

NaI detector (Teasdale 1985). The counting efficiency, E (counts per disintegration), can be 

determined from E = (1.3 × 10-2

) (e-0.0002778 Eγ), where Eγ is in keV. Little background information is 

available for such large crystals in IAEA (1970). However, it can be estimated from the value for the 

8-in. by 4-in. detector value by using the ratio of detector face areas:

0

0.002

0.004

0.006

0.008

0.01

0.012

0.014

0 500 1000 1500 2000 2500

Photon energy - keV

Eff. - c/dis.

8x4

11.5x4

0

5

10

15

20

25

30

0 500 1000 1500 2000 2500

Photon energy - keV

Eff. - cpm/nCi

8x4

11.5x4

ATTACHMENT A

OCCUPATIONAL INTERNAL DOSE FOR MONITORED WORKERS

Page 13 of 50
Document No. ORAUT-TKBS-0008-5 Revision No. 03 Effective Date: 08/17/2012 Page 80 of 116

Bkg11.5x4 = Bkg8x4 × 2.07 = 279 cpm (A-3)

The resulting estimate for the 662-keV MDA is 2,365 γ/min, or 1.06 nCi. In this case, the reported 

increase in detector efficiency does not fully compensate for the calculated increase in background. 

For the period from 1981 to 1993, the MDA values in Table A-9 are based on a nominal photon MDA 

of 1.1 nCi, with correction for the yield of the primary measurement photon, Yγ – MDA = 1.1/Yγ nCi.

Table A-9. Whole-body counting MDAs.

a

Radionuclide

Primary photon 

energiesb (keV)

Yγ

Photons per 

disintegration

Estimated MDAs (nCi)

1967c

1968–

1977d

1978–

1980e

1981–May 

1993f

June

1993g

Na-22 1,274 0.999 2.1 10 1.0 1.1 0.7

Na-24 1,369

2,754

1.00

0.999

2.1 10 1.0 1.1 1.1

Sc-46 889

1,121

1.00

1.00

2.1 10 1.0 1.1 1.1

Mn-54 835 1.00 2.1 10 1.0 1.1 1.2

Fe-59 1,099

1,292

0.565

0.432

3.8 18 1.8 2.0 2.3

Co-57 122 0.856 2.5 12 1.2 1.3 1.4

Co-60 1,173

1,333

1.00

1.00

2.0 10 1.0 1.1 0.9

Zn-65 1,116 0.506 4.0 20 2.0 2.2 -

Sr-85 514 0.96 2.2 10 1.0 1.1 1.8

Zr-95 724

757

0.442

0.540

3.8 19 1.9 2.1 1.9

Nb-95 766 1.00 2.1 10 1.0 1.1 1.1

Mo-99 (140) 0.89 2.4 11 1.1 1.2 10h

Tc-99m 7.1i

Rh-101 1.4i

Ru-103 497 0.909 2.3 11 1.1 1.2 1.2

Ru-106, Rh-106 (512)

(622)

0.20

0.0993

11 50 5.0 5.5 11

Rh-102 475

556

0.384

0.960

2.2 10 1.0 1.1 1.0

Rh-102m 475

631

0.56 2.3 18 1.0 1.1 7.9i

Sb-122 564 0.710 3.0 14 1.4 1.5 2.1

Sb-124 603

1,691

0.980

0.487

2.2 10 1.0 1.1 1.6

Sb-125 428

464

601

636

0.300

0.105

0.179

0.113

5.3 33 3.0 3.3 3.6

Te-132 228 0.88 2.4 11 1.1 1.2 1.5j

I-131 364 0.817 2.6 12 1.2 1.3 1.6j

I-132 667

773

0.99

0.756

2.1 10 1.0 1.1 1.4j

I-133 530 0.870 2.4 11 1.1 1.1 1.4j

ATTACHMENT A

OCCUPATIONAL INTERNAL DOSE FOR MONITORED WORKERS

Page 14 of 50
Document No. ORAUT-TKBS-0008-5 Revision No. 03 Effective Date: 08/17/2012 Page 81 of 116

Radionuclide

Primary photon 

energiesb (keV)

Yγ

Photons per 

disintegration

Estimated MDAs (nCi)

1967c

1968–

1977d

1978–

1980e

1981–May 

1993f

June

1993g

I-135, Xe-135m (526)

1,132

1,260

0.805

0.227

0.289

2.6 12 1.2 1.3 1.4j

Cs-134 605

796

0.976

0.855

2.2 10 1.0 1.1 1.5

Cs-136 341

818

0.422

1.00

2.1 10 1.0 1.1 1.4

Cs-137 662 0.851 2.1 12 1.2 1.3 1.7

Ba-133 356 0.620 3.6 16 1.6 1.8 1.8

Ba-133m 276 0.178 12. 56 5.6 6.2 1.1

Ba-140 4.8

La-140 (1,596) 0.954 2.2 10 1.0 1.1 0.4

Ce-139 166 0.80 2.7 12 1.2 1.3 1.4i

Ce-141 145 0.482 4.4 21 2.1 2.3 2.6i

Ce-143 293 0.428 5.0 23 2.3 2.5 1.4i

Ce-144 134 0.111 19. 90 9.0 9.9 12i

Nd-147 531 0.131 16. 76 7.6 8.4 5.5i

Eu-152 122

344

1,408

0.286

0.265

0.210

7.1 35 3.5 3.9 4.2i

Eu-154 123

723

1,274

0.408

0.202

0.352

5.3 25 2.5 2.8 3.2i

Eu-155 - - NA NA NA NA 4.1i

Yb-169 - - NA NA NA NA 0.2

Lu-174 - - (k) (k) (k) (k) 2.5i

Lu-174m - - (k) (k) (k) (k) 3.7i

Ta-182 1,121

1,189

1,221

1231

0.349

0.162

0.27

0.114

6.1 29 2.9 3.2 3.3i

W-181 - - (k) (k) (k) (k) 4.6i

W-187 480

686

0.218

0.273

7.8 37 3.7 4.1 4.6

Ir-192 316

468

0.828

0.478

2.6 12 1.2 1.3 0.9

Ra-226 & 

progeny

- - (k) (k) (k) (k) 504i,l

Ac-227, Th-227 (236)

(256)

0.123

0.070

17.0 81 8.1 8.9 28i.m

Th-228, Ra-224 (241) 0.041 51.8 244 24. 26. 33i.n

Pa-231 - - (k) (k) (k) (k) 6.7i

U-235 186 0.572 3.7 17 1.7 1.9 3.4i

Th-232, Ac-228 (911)

(965)

(969)

0.258

0.0499

0.158

8.5 39 3.9 4.3 5i.o

ATTACHMENT A

OCCUPATIONAL INTERNAL DOSE FOR MONITORED WORKERS

Page 15 of 50
Document No. ORAUT-TKBS-0008-5 Revision No. 03 Effective Date: 08/17/2012 Page 82 of 116

Radionuclide

Primary photon 

energiesb (keV)

Yγ

Photons per 

disintegration

Estimated MDAs (nCi)

1967c

1968–

1977d

1978–

1980e

1981–May 

1993f

June

1993g

U-233 - - (k) (k) (k) (k) 13i

U-237 208 0.212 9.9 47 4.7 5.2 4.8i

Np-237, Pa-233 (854)

(865)

0.34

0.157

6.4 29 2.9 3.2 2.8i,p

Np-239 106

228

278

0.272

0.108

0.144

7.8 37 4.7 5.2 5.6i

Am-241 - - (k) (k) (k) (k) 4.3i

Am-243 - - (k) (k) (k) (k) 2.8i

Cm-243 228

278

0.106

0.140

14.9 71 7.1 7.8 4.9i

a. Source: IAEA (1970).

b. Parenthesis denotes daughter photon energy used for detection.

c. HNS shadow shield, with 8-in. by 4-in. NaI detector, 8-minute count (HNS 1967).

d. Pan American or REECo shadow shield with 3-in. by 3-in. NaI detector, 20-minute count (Teasdale 1985).

e. REECo shadow shield with 8-in. by 4-in. NaI detector, 1,300-second count time.

f. REECo scanning bed with 11.5-in. by 4-in. NaI detector, 1,200-second count time (Teasdale 1985).

g. REECo germanium with chair geometry (Allen et al. 1993).

h. Only bioassay method identified for this radionuclide.

i. In conjunction with lung count.

j. In conjunction with thyroid count.

k. No gamma with energy greater than 100 keV with sufficient abundance on which to base the estimate.

l. 50 nCi for Ra-226; 4 nCi based on Pb-241 in equilibrium with Ra-226.

m. Based on Th-227 in equilibrium with Ac-227.

n. Based on Ra-224 in equilibrium with Th-228.

o. Based on Ac-228 in equilibrium with Th-232.

p. Pa-233 is used to measure Np-237 by whole-body count; the MDA for this procedure is 2.8 nCi of Pa-233 which, in 

equilibrium with Np-237, also represents 2.8 nCi of Np-237.

Consistent with industry-wide improvements in this area, a variety of hardware and software 

applications have been used for whole-body counting. In 1993, workers in the routine bioassay 

program received employment, annual, and termination WBCs. Workers who received routine 

bioassays included RCTs and field operation supervisors, chemists and laboratory technicians 

supporting the Analytical Services Department, industrial hygiene personnel trained as Radiation 

Workers, and workers in airborne or contamination areas or who shipped or disposed of radioactive 

material who were trained as Radiation Workers.

Whole-body counting was used to detect intake of most gamma-emitting fission and activation 

products. While an MDA was listed, it was not used for plutonium and americium bioassay due to 

their low-energy photon emissions. Some MDAs were reported on the forms for individual 

measurements; these should be used when they are available. The 137Cs WBC sensitivity for 1968 to 

1991 was 0.011 μCi for a 20-minute count and 0.010 μCi for a 40-minute count. The MPBB of 137Cs in 

standard man is 30 μCi.

A.2.2 Chest Counting

Based on Allen et al. (1993), lung counts occurred as soon as practicable after a suspected intake of 

thorium, uranium, or a TRU element. The first mention of lung counts in REECo procedures was in 

July 1983 (REECo 1983). There is no earlier mention of chest counting in REECo documentation; 

ATTACHMENT A

OCCUPATIONAL INTERNAL DOSE FOR MONITORED WORKERS

Page 16 of 50
Document No. ORAUT-TKBS-0008-5 Revision No. 03 Effective Date: 08/17/2012 Page 83 of 116

however, “lung counts” are noted in some safety reports without specific information provided. MDAs 

were generated individually for counts by processing software, with typical values listed in Table A-10.

Table A-10. 1993 MDAs for chest (lung) counting.

(a)

Radionuclide MDA (nCi) Radionuclide MDA (nCi)

Tc-99m 0.02 Pa-231 0.1

Rh-101 0.02 Th-232 (e) 0.2

Rh-102m 0.05 U-233 0.5

Ce-139 0.03 U-234 7

Ce-141 0.03 U-235 0.04

Ce-143 0.05 U-237 0.04

Ce-144 0.2 U-238(f) 9(0.4)

Nd-147 0.04 Np-237 0.05

Eu-152 0.03 Pu-238 2.8

Eu-154 0.04 Np-239 0.05

Eu-155 0.05 Pu-239 7.3

Lu-174m 0.05 Pu-240 2.9

Lu-174 0.04 Pu-241 (g) 200

W-181 0.05 Pu-242 40

Ta-182 0.04 Am-241 0.04

Ra-226 (b) 0.06 Cm-243 0.05

Ac-227 (c) 0.3 Cm-244 1.5

Th-228 (d) 0.7 Cf-252 4

Th-230 4 Am-243 0.01

Source: Allen et al. (1993).

a. Values are based on 1,000-second count.

b. Value based on Pb-214 in equilibrium with Ra-226 (MDA is approximately 1 nCi for 

Ra-226).

c. Based on Th-227 in equilibrium with Ac-227.

d. Based on Ra-224 in equilibrium with Th-228.

e. Based on Ac-228 in equilibrium with Th-232.

f. When U-238 is measured directly by lung count, the MDA is 9 nCi based on the detection of 

the 16-keV photon. When determined from Th-234, the MDA for U-238/Th-234 is 0.4 nCi 

based on the 63-keV photon from thorium, assuming equilibrium. [Note: Allen et al. (1993)

lists 0.4 nCi in the text (p. 49) and 0.04 nCi in the summary table (Table 4.1)].

g. Based on Am-241 measurement.

A.2.3 Thyroid Counts

Based on Allen et al. (1993), thyroid counts occurred as soon as practicable after a suspected 

radioiodine uptake. Processing software generated individual MDAs for counts, with typical values 

listed in Table A-11.

Table A-11. MDAs for thyroid counts.

Radionuclide Procedure MDA (nCi)

Te-132 0.004 prompt with special WBC 0.02

I-131 2.2E-4 (thyroid) prompt with special (0.006 μCi) WBC 0.03

I-132 0.15 μCi (thyroid) prompt with special (6.5 μCi) WBC 0.01

I-133 6.1E-4 prompt with special (0.01 μCi) WBC 0.04

I-135 0.07

Source: Allen et al. (1993).

ATTACHMENT A

OCCUPATIONAL INTERNAL DOSE FOR MONITORED WORKERS

Page 17 of 50
Document No. ORAUT-TKBS-0008-5 Revision No. 03 Effective Date: 08/17/2012 Page 84 of 116

Allen et al. (1993) states that radionuclides such as 132Te, 131I, 133I, and 135I are not reliably detected by 

routine whole-body counts because of their short physical half-lives. Special bioassay programs, 

such as air monitoring and breathing zone samples (based on work situation and exposure potential), 

are used to detect intakes of 132Te, followed by prompt special whole-body and thyroid counts for 

anyone suspected of being exposed to tellurium. Special whole-body and thyroid count bioassays are

promptly conducted for anyone suspected of being exposed to radioiodines. “Prompt” is interpreted to 

mean as soon as possible.

Table A-26 later in this attachment lists tests and incidents for which DRs might need to address 

iodine as an acute intake. Based on records provided by DOE, the internal dose can be calculated 

from bioassay results. If no results were provided, the DR should conclude that an acute intake 

evaluation is necessary based on information in the DOE access records indicating that the worker 

entered a controlled access area within 1 month of the test or was present during an incident or 

immediately after a test. From these records if available, the DR can determine the time and date of 

entry after the test, and compare this to the date that the test occurred. For a thyroid cancer, the 

relative amounts of various iodides over time are listed in Table 5-8.

Without specific access information for a worker, the DR should base internal exposure on 

environmental levels. Radionuclides listed in Table A-11 should be compared to the lists in Section 

A.4 that most closely characterize the type of test the worker was supporting, and the DR must decide 

which radionuclides are most significant to dose based on the cancer location. The list of 

radionuclides of concern varies according to the types of event and time interval before the worker 

was present. The DOE records might contain access records for test support (e.g., reentry, mineback, drill-back) including the times the worker entered and exited the controlled areas. If there are 

no specific access data, short-lived radionuclides will not be a concern. Exceptions to this will be 

workers who can be placed in the area where an uncontrolled venting occurred. Table A-26 lists 

releases from underground testing that DRs can review when determining if the worker was in the 

area during one of these releases.

In addition, DRs must consider the timeframe. DRs can determine when the worker was present from 

the “Other Monitoring” section in the DOE records. This section includes NTS Access Records for the 

worker’s reentry into an area after an event. If there are no “Other Monitoring” data, short-lived 

radionuclides are not a concern. Exceptions are workers who can be placed in the area where an 

uncontrolled venting occurred. Table A-23 in Section A.3.1.19 lists releases from underground testing 

that DRs can review when determining if the worker was in the affected area during one of these 

releases.

A.3 AREA AND ASSOCIATED RADIONUCLIDES OF CONCERN

A.3.1 Radionuclides of Concern and Specific Bioassay Programs for Facilities

The information in the following paragraphs is from the 1993 NTS Technical Basis (Allen et al. 1993), 

Chapter 7, Facility Descriptions and their Specific Routine Bioassay Programs. The NTS Technical 

Basis reflects radiological protection practices from about 1970 through the end of nuclear weapons 

testing in 1992, and is the best available source of internal dosimetry information for the nuclear 

weapons testing era (Arent and Smith 2004). The radionuclides of concern were determined in the 

following way:

ATTACHMENT A

OCCUPATIONAL INTERNAL DOSE FOR MONITORED WORKERS

Page 18 of 50
Document No. ORAUT-TKBS-0008-5 Revision No. 03 Effective Date: 08/17/2012 Page 85 of 116

• The radionuclides in a facility or area were identified from personal interviews, survey reports, 

radioactive material accountability reports, knowledge of past and present operations, and the 

open literature.

• The radionuclides in each area or facility, whose radiotoxicity and exposure potential could 

combine to deliver 90% of the maximum dose to an individual, were considered to be the 

radionuclides of concern.

• The radionuclides present in sufficient quantities, with gamma or beta emissions with 

detectable energies, were used to identify the possible presence of other radionuclides more 

difficult to detect. The indicator radionuclides were used in the Internal Dosimetry Program for 

screening routine bioassay samples. Further bioassay samples might have been collected if 

the screening isotopes were present.

The results of the urine samples might have indicated further bioassay measurements such as 

thyroid, lung, or whole-body counting, fecal sampling, or processing the urine samples for beta and 

alpha particle emitters. NOTE: Many radionuclides listed in the following paragraphs were not 

specifically measured by the bioassay program. Routine bioassay analysis was done for the following 

radionuclides:

• 241Am: Lung count

• Tritium: Urine, wound swabs

• Gamma emitters: Urine, WBC, wound swab

• Gross alpha: Wound swab

• GFP (gross beta): Urine, wound count/swab

• Iodines: Thyroid count

• 90Sr: Urine

• 238U: Lung count

• 238Pu: Urine, feces, lung count, wound count

• 239Pu: Urine, feces, lung count, nasal swabs, wound count

A.3.1.1 Drill-Backs (Yucca Flat and Pahute Mesa, or Areas 1-10 and 18-20)

Drill-back operations took place within days or weeks of an underground nuclear weapons test. Initial 

reentry surveys in the radiological exclusion area up to the trailer park usually began within 1 hour

after a test and the duration of the survey was typically about 1 hour depending on logistics (test 

location, weather, etc.) or whether there were identified problems. Initial reentry surveys were 

conducted by the REECo Rad-Safe Monitors. Typically, post-test drilling would begin within 1 to 4 

days after a test, but could begin more than a week after a test due to a variety of factors 

(containment success, initial survey results, test location, test type, weather, schedule, etc.). The 

duration of the post-test drilling operation was typically 2 to 6 days, but due to the factors mentioned 

above, plus concerns such as circulation problems, sampling issues, hole integrity, etc., a drill-back 

could last 2 to 3 weeks or more. Specific information about a particular drill-back might be available in 

the DOE records. The primary goal of the drill-back was to recover samples of the condensed fission 

and activation products remaining in the cavity created during the explosion. This was accomplished 

with a directional drilling technique to recover core samples from the cavity. Because of the short 

period after the test, it was possible to encounter high gas pressures that created a potential exposure 

to gaseous and particulate fission and activation products. Starting in 1964, engineering devices were 

ATTACHMENT A

OCCUPATIONAL INTERNAL DOSE FOR MONITORED WORKERS

Page 19 of 50
Document No. ORAUT-TKBS-0008-5 Revision No. 03 Effective Date: 08/17/2012 Page 86 of 116

used to prevent the escape of radioactive gases and particulates; however, there was a potential for 

release in some instances. The pathways for release and personnel exposure were (1) loss of 

containment during drilling or coring, and (2) resuspension of particulate fission or activation products 

during the coring operations.

Radionuclides for identification of a problem from containment loss and dose concern were 131I, 133I, 

and 137Cs. Table A-12 lists radionuclides for identification of a problem from resuspension and used 

to trigger searches in bioassays for the radionuclides of concern for resuspension. Table A-13 lists 

radionuclides of concern from drill-back resuspension.

Table A-12. Drill-back resuspension and mine back 

containment loss radionuclides for identification versus 

time after test. (Allen et al. 1993)

1 day 10 days 100 days 365 days 10,000 days

Na-24 Na-24a

Mn-54 Mn-54 Mn-54 Mn-54 Co-60

Fe-59 Fe-59 Fe-59 Co-57 Sr-90

Co-57 Co-57 Co-57 Co-60 Sb-125

Co-60 Co-60 Co-60 Sr-90 Ba-133

Y-91 Sr-89 Sr-89 Y-91 Cs-137

Zr-95 Sr-90 Sr-90 Zr-95

Zr-97 Y-91 Y-91 Nd-95

Mo-99 Zr-95 Zr-95 Ru-106

Tc-99m Nd-95 Nd-95 Sb-125

Ru-103 Mo-99 Ru-103 Ba-133

Ru-106 Ru-103 Ru-106 Cs-135

Sb-124 Ru-106 Sb-124 Ce-139

Sb-125 Sb-124 Sb-125 Ce-144

I-131 Sb-125 Ce-139 Pm-147

I-133 I-131 Ce-141 Ta-182

I-135 Te-132 Ce-143

Cs-137 Ba-140 Ce-144

Ba-140 La-140 Ta-182

La-141 Ce-141

Ce-143 Ce-144

Ce-144 Nd-147

Nd-147 Ta-182

Pm-149

Ta-182

a. Added in response to Sanford Cohen & Associates comment.

Table A-13. Drill-back resuspension, reentry/mine back 

resuspension, and decontamination facility, isotopes of 

concern for dose versus time after test (Allen et al. 1993). 

1 day 10 days 100 days 365 days 10,000 days

Zr-95 Sr-89 Sr-89 Sr-90 Sr-90

Zr-97 Y-91 Sr-90 Zr-95 Cs-137

Mo-99 Zr-95 Y-91 Ru-106

Ru-106 Ru-103 Zr-95 Ce-139

I-131 Ru-106 Ru-103 Ce-144

Te-132 I-131 Ru-106 Pm-147

ATTACHMENT A

OCCUPATIONAL INTERNAL DOSE FOR MONITORED WORKERS

Page 20 of 50
Document No. ORAUT-TKBS-0008-5 Revision No. 03 Effective Date: 08/17/2012 Page 87 of 116

1 day 10 days 100 days 365 days 10,000 days

I-133 Te-132 Ce-144

I-135 Ce-141

Ce-143 Ce-144

Ce-144

The routine bioassay for drill-backs was to collect quarterly urine samples after each drill-back 

operation. The urine samples were analyzed for gamma emitters by gamma spectroscopy for gross 

gamma-emitting radioactivity and for radionuclide identification. The primary isotopes used for 

identification of an intake were 131I, 133I, and 137Cs, and those listed in Table A-12. Annual WBCs were 

conducted on a routine basis for specific radiation workers, and on a special basis after any situation 

where intake was considered likely. Specific radiation workers include RCTs, radiological field 

operations supervisors, chemists and laboratory technicians, specific industrial hygiene personnel, 

and specific radiation worker-trained personnel who worked in airborne radioactive material areas or 

contamination areas, or who ship or dispose of radioactive material. Specific drilling job categories 

included:

• Driller Operator Supervisor (Oil Field or Core Drill Type)

• Driller Operator (Oil Field or Core Drill Type)

• Rotary Drill Operator/Rotary Drill Helper

• Driller Helper (Oil Field or Core Drill Type)

• Derrickman (Oil Field Type)

• Motorman (Oil Field Type)

• Fishing Tool Engineer (Oil Field Type)

• Drill Helper Trainee

Other historical job titles that might appear in the records include Rough Neck and Equipment Oiler. 

The use of engineering controls such as the blowout preventer was incorporated to contain 

radioactive gases in 1964. DRs will need to evaluate the potential for exposure based on review of 

the access records provided in the DOE claim file before 1964. ORAUT (2008) describes drill-back 

operations including potential exposures.

A.3.1.2 Reentry and Mine Back (Area-1 & Area-12)

Reentry and mine back operations were similar to those for drill-backs. The major differences were 1) 

the time after a test that the cavity was entered, 2) the horizontal method of drilling, and 3) the fact 

that operations took place in a confined underground environment. Additionally, in some cases, the 

cavity was opened and personnel entered.

Initial (surface) reentry surveys in the radiological exclusion area (from a preassigned location in the 

forward area (e.g., Gate 300) up to and including the tunnel portal yard and the Rainier Mesa Trailer 

Park) usually began within 1 to 2 hours of a test and the duration of the survey was typically about 

1 hour depending on the test location or if problems were identified. Initial surface reentry surveys 

were conducted by REECo Rad-Safe Monitors. Once the initial (surface) reentry survey was 

completed, remote gas sampling was conducted and the initial underground reentry began.

The initial underground reentry usually began 1 day after the test and the duration varied depending 

on various factors (tunnel condition, containment success, radiological readings, etc.). Multiple 

reentries and concurrent work activities were typically conducted until the Test Director declared the 

ATTACHMENT A

OCCUPATIONAL INTERNAL DOSE FOR MONITORED WORKERS

Page 21 of 50
Document No. ORAUT-TKBS-0008-5 Revision No. 03 Effective Date: 08/17/2012 Page 88 of 116

official end of the reentry phase. Teams with different missions, but with some overlapping duties, 

were involved in the underground reentry, including:

• Work Teams (miners and an RCT) assessed tunnel integrity and performed tunnel safety 

inspections, removed doors and panels, reinstalled train track, reestablished ventilation, 

conducted radiation surveys, and collected samples up to a preestablished location. This was 

the first team to enter the tunnel after the test.

• Initial Reentry Teams, Reentry Teams, Mine Rescue Teams (miners, Industrial Hygienists, 

RCTs, and other crafts as needed) advanced the reentry into the potentially contaminated 

area. These workers collected gas samples; performed radiation surveys; established utility 

services; opened crawl tube doors, panels, and valves; reestablished ventilation; installed 

security grates; and assessed tunnel integrity. These teams were typically the first to 

encounter radiological exposure and wore appropriate contamination control PPE (e.g., 

anticontamination clothing and self-contained breathing apparatus) as needed.

• Scientific Teams - Data Recovery Team, Special Assessment Team, Scientific Assessment 

Team (national laboratory personnel plus RCTs) collected experiments and data. Once the 

reentry teams surveyed and cleared the retrieval areas, these teams typically entered with 

minimal or no required contamination control PPE. Multiple entries were required to retrieve 

the required instrumentation.

Mining through barriers (e.g., gas seal plug, drift protection plug, overburden plug) could begin within 

2 or 3 days of a test, but could be delayed a month or more after the test depending on conditions. 

National laboratory and DNA personnel continued their photography and recovery efforts as the mineback progressed. Radiological survey support continued during each phase of the recovery period. 

The mission of the scientific teams was to enter the line-of-sight (LOS) pipe and collect experiments/ 

data from the test cells. The duration of the mine-back varied depending on test results and schedule.

Post-test drilling from the surface ground zero usually began 1 to 2 months after the tunnel test. 

Drilling was mentioned in the onsite radiological reports in the 1960s and early 1970s. It apparently 

did not occur after about 1974.

There were four pathways for exposure during reentry and mine backs. The first pathway is based on 

a loss of containment in the drilling or coring operations during routine drilling or by failure of 

containment equipment. The radionuclides of concern were gaseous fission and activation products 

or their particulate daughters. The second pathway was resuspension of particulate fission or 

activation products during the coring operations of the drill-back. The third exposure pathway was the 

possibility for gaseous fission or activation products to seep through fissures in the rock and reenter 

the working areas of the tunnel. The fourth intake possibility existed when the LOS pipe was opened 

to remove experiment equipment and samples. Underground operations and construction included 

drilling, mining and mucking, haulage and transportation, rock bolting, ground control and timbering, 

ventilation, grouting (applying cement as containment material), welding, electrical and utilities 

support, and barring down (inspection for loose ground). These operations used unique equipment 

including continuous Alpine mining machines, (load/haul/dump) mucking machines (small front-end 

loaders to remove mined material), dinky locomotives (train engines) and cars used to move 

personnel and materials (e.g., Moran cars for cement and grout, powder cars for explosives, high cars 

for elevated work), shotcrete machines [equipment to apply shotcrete or fibercrete, which is a cementATTACHMENT A

OCCUPATIONAL INTERNAL DOSE FOR MONITORED WORKERS

Page 22 of 50
Document No. ORAUT-TKBS-0008-5 Revision No. 03 Effective Date: 08/17/2012 Page 89 of 116

like material imbedded with metal wire and applied to the tunnel back and ribs (ceiling and walls) for 

ground control], jackleg and Jumbo drills, and spaders (similar to a jackhammer).

Specific tunnel job categories included:

• Miner

• Bull Gang (Underground Laborer)

• Mucker (Muck Machine Operator)

• Shifter

• Tunnel Walker

• Dinky Locomotive Operator

Radionuclides of dose concern from containment loss were 3

H, 131I, 133I, and 137Cs. Radionuclides for 

identification of a problem from containment loss and resuspension are listed in Table A-12. 

Radionuclides of dose concern from resuspension are listed in Table A-13. Radionuclides for 

identification of a problem from fissures and LOS pipe opening were 3

H, 7Be, 59Fe, 60Co, 124Sb, 131I, 

133I, 137Cs, and 182Ta. Isotopes of dose concern from fissures and LOS pipe openings were 3

H, 60Co, 

124Sb, 131I, 133I, and 137Cs.

The routine method for bioassay for reentry and mine backs was to collect urine samples based on 

the results of air samples. Tunnel air sampling began in 1957 in locations with the potential for 

airborne exposure (Arent and Smith 2004). Air samplers operated continuously. RCTs checked and 

exchanged the filters each shift. Bioassay was performed only if there was an indication of an effluent 

release (e.g., positive air sample) (Arent and Smith 2004). The urine samples were gamma counted 

and sampled for 3

H. Annual whole-body counts were conducted on a routine basis for specific 

radiation workers and on a special basis after any situation for which intake was considered likely.

A.3.1.3 Routine Tunnel Operations (Areas 1, 12, 15, and 16)

Tritium was the isotope of dose concern; routine bioassay was to collect quarterly urine samples 

which were processed for 3

H. (Allen et al. 1993 contains no bioassay information.) See Section 

5.2.2.4 for a discussion on assigning tritium dose.

A.3.1.4 Decontamination Facility (Area 6)

Table A-14 lists radionuclides used for identification of a problem at this facility. Table A-13 lists 

radionuclides of concern. In addition to the radionuclides listed in Table A-11, BN (2000) lists the 

following radionuclides of concern for this facility: 232Th, 234U, 235U, 238U, 239Pu, and 241Am. The routine 

bioassay method was to collect quarterly urine samples and conduct annual WBCs. Urine samples 

were gamma-counted and analyzed for GFP, 238Pu, 239Pu, 240Pu, and 241Am.

A.3.1.5 Test Treatability Facility (Area 25)

The Test Treatability Facility was a pilot project to bench-test technologies to be used for 

decontamination of soils containing TRU materials. The concentrations of radionuclides in the soils 

were not intended to exceed a few picocuries per gram of soil. Radionuclides for identification of a 

problem and dose concern were 238Pu, 239Pu, 240Pu, and 241Am. Routine bioassay was to collect 

quarterly urine samples and conduct annual whole-body counts. Urine samples were counted on a 

gamma analyzer for isotopic identification and analyzed for plutonium and americium.

ATTACHMENT A

OCCUPATIONAL INTERNAL DOSE FOR MONITORED WORKERS

Page 23 of 50
Document No. ORAUT-TKBS-0008-5 Revision No. 03 Effective Date: 08/17/2012 Page 90 of 116

Table A-14. Decontamination facility, isotopes for 

identification versus time after test.

1 day 10 days 100 days 365 days 10,000 days

Sr-89 Sr-89 Sr-89 Sr-90 Sr-90

Sr-90 Sr-90 Sr-90 Y-91 Ba-133

Zr-93 Y-91 Y-91 Zr-93 Cs-137

Zr-95 Zr-93 Zr-93 Zr-95 Pu-239

Zr-97 Zr-95 Zr-95 Ru-106 Am-241

Mo-99 Mo-99 Ru-103 Ba-133

Tc-99m Ru-103 Ru-106 Cs-135

Ru-103 Ru-106 Ba-133 Cs-137

Ru-106 I-131 Cs-135 Ce-139

I-131 Te-132 Cs-137 Ce-144

Te-132 I-133 Ce-139 Pm-147

I-133 Ba-133 Ba-140 Pu-239

Ba-133 I-135 Ce-141 Am-241

I-135 Cs-137 Ce-143

Cs-137 Ba-140 Ce-144

Ce-139 La-140 Pu-239

Ba-140 Ce-141 Am-241

La-140 Ce-144

Ce-141 Nd-147

Ce-143 Pu-239

Ce-144 Am-241

Nd-147

Pm-149

Pu-239

Am-241

A.3.1.6 Atmospheric Weapon Safety Tests (Area 3 and Area 11)

Isotopes for identification of a problem, sample screening, and of dose concern were 238Pu, 239Pu, 

240Pu, and 241Am. Routine bioassay was to collect quarterly urine samples and conduct annual wholebody counts. Urine samples were gamma-counted and analyzed for plutonium and americium.

A.3.1.7 Atmospheric Weapons Test Areas (all Areas except Area 22, Area 23, and Area 27 

due to test location or airborne dispersion)

Table A-15 lists radionuclides for identification of a problem and dose concern. Routine bioassay was 

to collect quarterly urine samples and conduct annual whole-body counts. Some area monitoring 

personnel also had periodic lung counts. Urine samples were analyzed for gamma-emitting 

radionuclides, GFP, plutonium, and americium.

Table A-15. Atmospheric weapons test areas, 

isotopes for identification and of concern for dose

(1993–present).

Co-60 Sb-125 Cs-137 Eu-155 Pu-239

Sr-90 Ba-133 Eu-152 Lu-174 Pu-240

Ru-101 Cs-134 Eu-154 Pu-238 Am-241

Ru-102m

ATTACHMENT A

OCCUPATIONAL INTERNAL DOSE FOR MONITORED WORKERS

Page 24 of 50
Document No. ORAUT-TKBS-0008-5 Revision No. 03 Effective Date: 08/17/2012 Page 91 of 116

A.3.1.8 Low-Level Waste Site (Area 3)

Table A-16 lists radionuclides for identification of a problem and dose concern. In addition to these 

radionuclides, BN (2000) lists 235U as a radionuclide of concern for this facility. Routine bioassay was 

to collect quarterly urine samples and conduct annual whole-body counts. Urine samples were 

gamma analyzed for GFP, plutonium, and americium.

Table A-16. Low-level waste site (Area 3), 

isotopes for identification and of concern for 

dose.

Mn-54 Ru-103 Ce-144 U-233 Pu-240

Co-60 Ru-106 Ac-227 U-234 Pu-241

Sr-85 Cs-134 Th-228 U-238 Pu-242

Sr-90 Cs-137 Th-230 Pu-238 Am-241

Zr-95 Ba-140 Th-232 Pu-239 Am-243

Nb-95 Ce-141

A.3.1.9 Low-Level Waste Site (Area 5)

Table A-17 lists radionuclides for identification of a problem and dose concern. Routine bioassay was 

to collect quarterly urine samples and conduct annual WBCs. Urine samples were analyzed for 

gamma emitters, sampled for 3

H, and analyzed for GFPs, plutonium, and americium.

Table A-17. Low-level waste site (Area-5), 

isotopes for identification and of concern for 

dose.

H-3 Mo-99 Ce-141 Ac-227 U-238

Na-22 Ru-103 Ce-144 Th-228 Pu-238

Mn-54 Ru-106 Eu-152 Th-230 Pu-239

Co-57 Sb-124 Eu-154 Th-232 Pu-240

Co-60 Sb-125 Eu-155 U-233 Pu-241

Sr-85 Ba-133 Yb-169 U-234 Am-241

Sr-90 Cs-134 Ta-182 U-235 Pu-242

Zr-95 Cs-137 Ir-192 Np-237 Am-243

Nb-95 Ba-140 Ra-226

A.3.1.10 Radiation Instrument Calibration Facilities (Area 6 and Area 23)

Table A-18 lists radionuclides for identification of a problem and dose concern. In addition to these 

radionuclides, BN (2000) lists 235U as a radionuclide of concern for these facilities. Routine bioassay 

for personnel who regularly worked with calibration sources was to collect quarterly urine samples and 

conduct annual whole-body counts. Urine samples were analyzed for 3

H and plutonium. A special 

bioassay whole-body count and/or urine sample collection was performed if the loss of calibration 

source containment was detected by a source leak test or other methods. The follow-up action and 

type of analyses conducted was determined by the particular source in question.

ATTACHMENT A

OCCUPATIONAL INTERNAL DOSE FOR MONITORED WORKERS

Page 25 of 50
Document No. ORAUT-TKBS-0008-5 Revision No. 03 Effective Date: 08/17/2012 Page 92 of 116

Table A-18. Radiation instrument calibration 

facilities, isotopes for identification and of 

concern for dose.

H-3 Sr-90 Ra-226 Th-230 Pu-239

Co-60 Cs-137 Th-228 Pu-238 Am-241

A.3.1.11 Radiography Operations (Area 23)

The isotope of concern was 192Ir. In addition to 192Ir, BN (2000) lists 60Co and 137Cs as radionuclides 

of concern for this operation. Bioassay (urine/WBC) was conducted if a loss of source containment 

was detected by swipes or other analysis. Urine samples were gamma-counted and follow-up actions 

taken accordingly.

A.3.1.12 Well Logging Operations (Areas 1–10 and Areas 18–20)

Isotopes for identification of a problem and dose concern were 60Co, 131I, 137Cs, 226Ra, 228Th, 238Pu, 

and 241Pu. Bioassay (urine/WBC) was conducted if there was a loss of source containment detected 

by swipes or other analysis. Urine samples were gamma-counted or processed for plutonium 

depending on the source type. Additional “Nuclides of Concern” listed in McMahan and Ogurek 

(2003) include the following well logging sources: 241Am7

Be, 238Pu7Be, and 226Ra7Be.

A.3.1.13 Nuclear Explosive and Device Assembly Facilities (Area 6 and Area 27)

Radionuclides for identification of a problem and dose concern were 3

H, 235U, 239Pu, and 240Pu. In 

addition to these radionuclides, BN (2000) lists 234U, 238Pu, 241Pu, and 241Am as radionuclides of 

concern for these facilities. No specific bioassay information was listed in Allen et al. (1993), BN 

(2000) or McMahan and Ogurek (2003) technical basis documents.

A.3.1.14 Nuclear Rocket Development Areas (Area 25 and Area 26)

Tests of nuclear reactors for use as propulsion units were conducted above ground into the 1960s. 

The fission and activation products from these tests were widely dispersed into the environment. 

Isotopes for identification of a problem and dose concern were 60Co, 90Sr, 137Cs, and 152Eu. In 

addition to these radionuclides, BN (2000) lists 234U, 235U, and 238U as radionuclides of concern for 

these areas. Routine bioassay was to collect quarterly urine samples and conduct annual WBCs. 

Urine samples were gamma-counted, sampled for 3

H, and analyzed for GFP and plutonium.

A.3.1.15 Radioactive Source Storage Areas (Area 6 and Area 23)

Table A-19 lists radionuclides for identification of a problem and dose concern. There was no routine 

bioassay program in support of normal source storage facility operations. If source leak test or other 

means indicated a stored source was leaking or had leaked radioactivity, special WBCs and/or urine 

sample collection and analysis were conducted. Specific bioassays and analysis depended on the 

source leaking.

ATTACHMENT A

OCCUPATIONAL INTERNAL DOSE FOR MONITORED WORKERS

Page 26 of 50
Document No. ORAUT-TKBS-0008-5 Revision No. 03 Effective Date: 08/17/2012 Page 93 of 116

Table A-19. Radiation instrument calibration 

facilities, isotopes for identification and of 

concern for dose.

Na-22 Ni-63 Cs-137 Ra-226 Pu-238

Mn-54 Sr-90 Eu-152 Th-228 Pu-239

Co-57 Cd-109 Eu-154 U-235 Am-241

Co-60 Ba-133 Ir-192

A.3.1.16 Radiochemistry and Counting Laboratories

Table A-20 lists radionuclides for identification of a problem and dose concern. In addition to these 

radionuclides, BN (2000) lists 226Ra, 232Th, 234U, and 238U as radionuclides of concern for these 

facilities. Routine bioassay was to collect quarterly urine samples and conduct annual WBCs only for 

those personnel routinely handling uncontained radioactive materials. Urine samples were gammacounted, sampled for 3

H, and analyzed for GFP and plutonium.

Table A-20. Radiochemistry and counting 

laboratories, isotopes for identification and of 

concern for dose.

H-3 Sr-85 Cd-109 Ce-144 Am-241

Na-24 Y-88 Sn-113 Eu-152 Pu-242

Co-57 Sr-90 Cs-137 Hg-20 Am-243

Co-60 Y-90 Ce-139 Pu-239 Cm-244

A.3.1.17 Other Facilities Identified After 1992 (the End of Testing)

McMahan and Ogurek (2003) included some facilities and radionuclides of concern that Allen et al. 

(1993) did not describe. The following paragraphs discuss these newer facilities and Site Monitoring 

Services (SMS)/Radioactive Material Control (RAMATROL):

• Big Explosive Experimental Facility (BEEF) (Area 4) – Most experiments at BEEF involve the 

use of high explosives. On occasion, depleted uranium or tritium could be used during an 

experiment. After detonation, particulates occur in the area. The nuclides of concern are 3

H, 

234U, and 238U.

• Waste Examination Facility (Area 5) – The WEF processes TRU waste drums in preparation 

for shipment to the Waste Isolation Pilot Plant. The drums are primarily from LLNL, but a 

small number are from LANL and Rocky Flats. The TRU waste is in 55-gal drums and 

processed in a glovebox. There is a storage area for the drums called the TRU Pad. The 

potential for internal exposure can come from a failure of glovebox integrity. Nuclides of 

concern at WEF are 233U, 238Pu, 239Pu, 240Pu, 241Am, 243Am, 244Cm, and 252Cf. Plutonium 

radionuclides and 241Am have been identified as the predominant nuclides of concern.

• SMS and RAMATROL – Personnel assigned to SMS and RAMATROL provide radiological 

survey services mainly on NTS but also at other sites as assigned. These services include 

contamination surveys of equipment and materials being removed from NTS, contamination 

surveys of equipment and buildings determined to be excess to government inventory, and 

surveys of radiological shipments entering and departing NTS. In addition, these personnel 

perform surveys of Radioactive Material Areas and of areas where handling of swipe samples 

ATTACHMENT A

OCCUPATIONAL INTERNAL DOSE FOR MONITORED WORKERS

Page 27 of 50
Document No. ORAUT-TKBS-0008-5 Revision No. 03 Effective Date: 08/17/2012 Page 94 of 116

and other potentially contaminated material occurs. The most common radionuclides these 

personnel encounter are aged fission products (e.g., 137Cs and 90Sr/90Y), 3

H, uranium, 239Pu, 

and 241Am. SMS and RAMATROL personnel could encounter these nuclides as removable or 

fixed contamination, typically with removable contamination less than 200 dpm/100 cm2 for 

beta emitters and 20 dpm/100 cm2 for alpha emitters.

A.3.1.18 Summary Lists of Radionuclides of Concern

Table A-21 summarizes radionuclides of dose concern for NTS. Table A-22 lists nuclides of concern 

at different NTS locations. Section 4.2.1.2 of this TBD includes tables of atmospheric radionuclide 

concentrations by year that can be used in conjunction with this table.

Table A-21. Radionuclides of concern for dose, 

summary list.

H-3 Mo-99 I-133 Lu-174 U-235

Na-22 Ru-101 Cs-134 Ta-182 U-238

Na-24 Ru-102m Cs-137 Ir-192 Np-237

Mn-54 Ru-103 Ba-140 Hg-203 Pu-238

Co-57 Ru-106 Ce-139 Ra-226 Pu-239

Co-60 Cd-109 Ce-141 Ac-227 Pu-240

Sr-85 Sn-113 Ce-144 Th-228 Pu-241

Y-88 Sb-124 Eu-152 Th-230 Pu-242

Sr-90 Sb-125 Eu-154 Th-232 Am241

Y-90 I-131 Eu-155 U-233 Am243

Zr-95 Ba-133 Yb-169 U-234 Cm244

Nb-95

Table A-22. Current nuclides of concern for NTS locations. 

NTS facility and area Radionuclides of concern

BEEFa U-234 U-238 H-3

Nuclear Explosive Assembly Facilities (DAF and 

Area 27)

H-3 U-234 U-235 Pu-238

Pu-239 Pu-240 Pu-241 Am-241

Routine tunnel operations (Areas 1 and 12) H-3

Decontamination Facility (Area 6) Sr-90 Cs-137 Th-232 U-234

U-235 U-238 Pu-239 Am-241

Legacy atmospheric weapons safety test areas Pu-238 Pu-239 Pu-240 Am-241

Legacy atmospheric weapons test areas; legacy 

weapons test waste trenches and support facilities

Co-60 Sr-90 Ru-101 Ru102m

Ba-133 Cs-137 Eu-152 Eu-154

Eu-155 Pu-238 Pu-239 Pu-240

Am-241

Tonopah Test Range and Area 13 safety test areas Pu-238 Pu-239 Pu-240 Am-241

Tonopah Test Range DU munitions tests U-234 U-238

Low-Level Waste Site, Area 3 Sr-90 Cs-137 Th-232 Pu-239

Am-241 U-234 U-235 U-238

Low-Level Waste Site, Area 5 H-3 Sr-90 Cs-137 Th-232

U-234 U-235 U-238 Pu-239

Am-241

ATTACHMENT A

OCCUPATIONAL INTERNAL DOSE FOR MONITORED WORKERS

Page 28 of 50
Document No. ORAUT-TKBS-0008-5 Revision No. 03 Effective Date: 08/17/2012 Page 95 of 116

NTS facility and area Radionuclides of concern

Ra-226 U-235 Pu-238 Pu-239

Am-241

Radiation instrument calibration facilities H-3 Co-60 Sr-90 Cs-137

Radiography operations Co-60 Cs-137 Ir-192

Well logging operations Co-60 Cs-137 Ra226/Be-7

Pu-238/Be-7 Am-241/Be-7

Nuclear Rocket Development Area (Area 25) Co-60 Sr-90 Cs-137 Nb-95b

U-234 U-235 U-238

Legacy biokinetic test areas (Test Cell A, Area 25) Am-241

Radioactive source storage areas Na-22 Mn-54 Co-57 Co-60

Ni-63 Sr-90 Cd-109 Ba-133

Cs-137 Eu-152 Eu-154 Ir-192

Ra-226 Th-228 U-235 Pu-238

Pu-239 Am-241

Radiochemistry and counting laboratories H-3 Na-24 Co-57 Co-60

Sr-85 Y-88 Y-90 Sr-90

Cd-109 Sn-113 Eu-152 Cs-137

Ce-139 Ce-144 Hg-203 Ra-226

Th-232 U-234 U-238 Pu-239

Pu-242 Am-241 Am-243 Cm-244

WEF U-233 Pu-238 Pu-239 Pu-240

Am-241 Am-243 Cm-244 Cf-252

SMS and RAMATROL H-3 Sr/Y-90 Cs-137 U-234 

U-235 U-238 Pu-239 Am-241

Source: McMahan and Ogurek (2003).

a. BN (2000) also lists H-3 as a radionuclide of concern.

b. Identified by Ogurek (2004) as a radionuclide of concern.

A.3.1.19 Atmospheric Testing Era

The following information is included for completeness. Based on the determination of Health and

Human Services, internal dose cannot be included for employment during the atmospheric testing era 

at either NTS or PPG (Leavitt 2006a,b).

Between June 1946 and October 1958, the United States used Enewetak and Bikini Atolls, part of the 

Marshall Islands approximately 2,500 nautical miles southwest of the Hawaiian Islands, as testing 

grounds for 67 nuclear devices. Some NTS workers might have been stationed in the Pacific for 

these tests. Materials associated with the nuclear tests conducted at or near Bikini and Enewetak 

Atolls include 210Po, 228Th, 230Th, 232Th, 233U, 238U, 241Am, 242Cm, and the fission product 137Cs 

(Robison et al. 2001). Radionuclides detected in the marine (lagoon) environment (plankton, fish, 

invertebrate, seawater, and sediments) from studies conducted from 1954 through the early 1970s 

included 3

H, 14C, 54Mn, 55Fe, 57Co, 58Co, 60Co, 63Ni, 65Zn, 90Sr, 95Zr, 99Tc, 102mRh, 106Ru, 110mAg, 144Ce, 

147Pm, 151Sm, 155Eu, 126Sb, 137Cs, 184W, 207Bi, 237Np, 237U, 239+240Pu, 241Pu, and 244Cm (Robison and 

Noshkin 1998). The limits found in NBS Handbooks 52 (NBS 1953) and 69 (NBS 1959) would have 

been in effect. SAIC (1985) lists the following radionuclides for Pacific testing: 148 fission products, 

17 activation products, and 18 actinide elements. The 148 Pacific testing fission products are as 

follows:

ATTACHMENT A

OCCUPATIONAL INTERNAL DOSE FOR MONITORED WORKERS

Page 29 of 50
Document No. ORAUT-TKBS-0008-5 Revision No. 03 Effective Date: 08/17/2012 Page 96 of 116

Table A-23. Fission products associated with above ground 

testing.

Se-79 Y-92 Mo-105 Sb-129 Cs-134m Pr-144

Br-82 Sr-93 Tc-105 Te-129 I-135 Pr-144m

Br-83 Y-93 Ru-105 Te-129m Xe-135* Ce-145

Kr-83m* Zr-93 Rh-105 I-129 Xe-135m* Pr-145

Br-84 Nb-93m Rh-105m Sn-130 Cs-135 Ce-146

Br-85 Y-94 Ru-106 Sb-130m Cs-136 Pr-146

Kr-85* Y-95 Rh-106 I-130 Xe-137* Ce-147

Kr-85m* Zr-95 Pd-107 Sb-131 Cs-137 Pr-147

Rb-86 Nb-95 Rh-109 Te-131 Ba-137m Nd-147

Kr-87* Y-96 Pd-109 Te-131m Xe-138* Pm-147

RB-87 Zr-97 Ag-109m I-131 Cs-138 Pr-148

Kr-88* Nb-97 Ag-110m Xe-131m* Cs-139 Pm-148

Rb-88 Nb-97m Ag-111 Sb-132 Ba-139 Pm-148m

Kr-89* Nb-98m Cd-113m Sb-132m Cs-140 Pm-149

Rb-89 Mo-99 Cd-115m Te-132 Ba-140 Pm-151

Sr-89 Tc-99 Sn-123 I-132 La-140 Sm-151

Rb-90 Tc-99m Sb-124 Sb-133 Ba-141 Eu-152

Rb-90m Mo-101 Sn-125 Te-133 La-141 Sm-153

Sr-90 Tc-101 Sb-125 Te-133m Ce-141 Eu-154

Y-90 Mo-102 Te-125m I-133 Ba-142 Eu-155

Rb-91 Tc-102 Sn-126 Xe-133* La-142 Eu-156

Sr-91 Tc-103 Sb-126 Xe-133m* La-143 Tb-160

Y-91 Ru-103 Sb-127 Te-134 Ce-143 Ho-166m

Y-91m Rh-103m Te-127 I-134 Pr-143

Sr-92 Tc-104 Te-127m Cs-134 Ce-144

* = gaseous fission product.

Table A-24. Activation products associated 

with above ground testing.

Be-7 Co-58 W-187

Na-24 Co-60 W-188

Mn-54 Cu-64 Au-198

Fe-55 Cu-67 Au-199

Fe-59 W-181 Pb-203

Co-57 W-185

Table A-25. Actinide elements associated 

with above ground testing.

Th-229 U-237 Pu-238

Th-230 U-238 Pu-239

Th-234 U-240 Pu-240

U-233 Np-237 Pu-241

U-234 Np-239 Am-241

U-235 Np-240 Cm-242

A.3.2 Incidents

Based on guidance from NTS, the DR should do the following when researching a claim:

1. Determine that the individual was working at NTS during a given operation.

ATTACHMENT A

OCCUPATIONAL INTERNAL DOSE FOR MONITORED WORKERS

Page 30 of 50
Document No. ORAUT-TKBS-0008-5 Revision No. 03 Effective Date: 08/17/2012 Page 97 of 116

2. Consult Compilation of Local Fallout Data from Test Detonations 1945–1962 (Hawthorne 

1979) in relation to atmospheric tests and Radiological Effluents Released from U.S. 

Continental Tests 1961–1992 (Schoengold, DeMarre, and Kirkwood 1996) in relation to 

underground tests to determine potential exposure to radioisotopes for a given test. United 

States Nuclear Tests, July 1945 through September 1992 (DOE 2000) provides a list of tests

by date or by test name. Other information contained in DOE (2000) includes operation name, 

test date and time, sponsor, location, hole number, latitude and longitude, surface elevation, 

test type, test purpose, and yield.

3. Establish if there is a link between the claimant and a release by timeframe, work location, and 

job title using ORAUT (2008) and Section A.4.1 of this document.

In addition, the DR might find useful the information in Table A-26 [from The Containment of 

Underground Nuclear Explosions (OTA 1989)] and Schoengold, DeMarre, and Kirkwood (1996) in 

identifying underground tests in which problems occurred.

The following paragraphs summarize other specific incidents that are important from an internal 

dosimetry standpoint at NTS (see Table A-26):

• AEC (1956), Part VI, Chapter 1, “Special Incidents,” describes an overexposure during a 

plutonium dispersion test. In January 1956, a work party was dispatched into the “hot” area to 

recover vital laboratory samples according to a plan that had been used successfully on earlier 

operations and for the first three tests of 56 Project – NTS. In performing this hazardous task, 

members of the party made three errors that resulted in four men receiving technical 

overexposures varying from 4.3 R to 28 R. An inquiry by the Test Director after the operation 

determined the causes of the overexposure. Routine Rad-Safe procedures in effect at NTS 

were amended to reduce the possibility of a reoccurrence of the errors committed.

• The Operation Plumbbob On-Site Radiological Report discusses four instances of possible 

internal exposure (REECo 1957):

– Escape of radioactive gas from the Tower 2-A cab contaminated the working area, 

resulting in 12 workers being exposed.

– One worker removed his respirator while working in an area highly contaminated with 

alpha-emitting material.

– Several workers were exposed while removing and cutting a cable highly contaminated 

with alpha-emitting material.

– Four workers without respirators entered a tunnel that was highly contaminated with alphaemitting materials.

• The Hardtack II report (Ponton et al. 1982) describes a tunnel incident in which an explosion 

occurred. Six REECo miners and two REECo radiation technicians were affected; no 

bioassay results were above permissible limits. The report also describes REECo tunnel 

decontamination methods, including air supplied respirators, the taking of nasal swabs, and 

full anticontamination clothing (clothing that might have consisted of coveralls, hoods, booties, 

and gloves).

ATTACHMENT A

OCCUPATIONAL INTERNAL DOSE FOR MONITORED WORKERS

Page 31 of 50
Document No. ORAUT-TKBS-0008-5 Revision No. 

0

3 Effective Date: 08/17/2012 Page 98 of 116

Table A-26. Releases from underground tests.

Test name Date Type

Purpose & 

yield Sponsor Location

Radionuclides 

identified In 

release

Release type & 

notes Release (Ci) Release summary

Platte 4/14/62 Tunnel Weaponsrelated

1.85 kt

LLNL U12k.01 K-40, Zr-95, 

Nb-95, Ru-103, 

Ru-105, I-131, 

I-133, I-135, 

Te-132, 

Ba-140/La-140, 

Ce-141, Ce-144

Test/prompt 

particle sampling 

release

Accidental 

release of 

radioactivity 

detected offsite

Test release at 

R+12 hours: 

1.9E+06

Venting occurred at tunnel 

portal, through fissures, and at 

sampling hole at H+1.5 

seconds. Fissures were created 

on side of hill, and radial cracks 

formed on top of hill. Persistent 

cloud was produced containing 

appreciable quantities of 

radioactivity associated with 

particles.

Eel 5/19/62 Shaft Weaponsrelated

4.5 kt

LLNL U9m Zr-95/Nb-95, 

Ru-103, Ru-106, 

Rh-105, I-131, 

I-133, I-135, 

Te-132,

Ba-140/La-140, 

Ce-141, Ce-144

Test/prompt 

particle sampling 

release

Accidental 

release of 

radioactivity 

detected offsite

Test release at 

R+12 hours: 

1.9E+06

Venting, in form of geyser, 

occurred at H+10 seconds from 

satellite hole U9m-2 and 

continued steadily until H+19 

minutes, 42 seconds. Similar 

venting occurred at H+15 

seconds from satellite hole 

U9m-3 and lasted until H+21 

minutes. Venting ceased with 

crater subsidence.

ATTACHMENT A

OCCUPATIONAL INTERNAL DOSE FOR MONITORED WORKERS

Page 32 of 50
Document No. ORAUT-TKBS-0008-5 Revision No. 

0

3 Effective Date: 08/17/2012 Page 99 of 116

Test name Date Type

Purpose & 

yield Sponsor Location

Radionuclides 

identified In 

release

Release type & 

notes Release (Ci) Release summary

Des Moines 6/13/62 Tunnel Weaponsrelated

2.9 kt

LLNL U12j.01 Ru-103, 

Ru-106/Rh-106, 

I-131, I-133, 

I-135, Te-132, 

Ba-140/La-140

Test/prompt 

particle sampling 

release

Accidental 

release of 

radioactivity 

detected offsite

Test release at 

R+12 hours: 

11E+06

Venting began at H+0.2 

seconds on top of hill at surface 

ground zero, then from sampling 

hole on face of hill, and finally 

through portal. Duration of 

release was about 5 minutes. 

Test vented from tunnel mouth 

with sufficient pressure and flow 

rate that radioactive debris was 

projected entirely across canyon 

and deposited on slope behind 

trailer shelter.

Baneberry 12/18/70 Shaft Weaponsrelated

10 kt

LLNL U8d GFPs

Isotopic 

analysise

: 

Mo-99, I-131, 

I-132, Te-132

Test release

Accidental 

release of 

radioactivity 

detected offsite

Test release at 

R+12 hours: 

6.7E+06

Venting occurred from fissure 

near surface ground zero at 

H+3.5 minutes.

Effluent venting rate steadily 

decreased with time, but visible 

vapor continued to emanate 

from fissure for 24 hours after 

detonation.

Camphor 6/29/71 Tunnel Weapons 

effects

<20 kt

LLNL

SNL

DOD

U12g.10 Test: Xe-133, 

Xe-135

Controlled: 

I-131, I-133, 

I-135

Test and 

controlled 

releases

Accidental 

release of 

radioactivity 

detected onsite 

only

Containment 

failurea

Test release at 

R+12 hours: 

360b

Test releases occurred from 

cable building (on mesa) at H+1 

hour, lasting for 30 minutes, and 

from portal at H+3.9 hours, 

lasting 4 days.

Controlled release through 

ventilation system of tunnel 

complex began at 1034 hours 

on July 27, 1971, and lasted 3 

days.

ATTACHMENT A

OCCUPATIONAL INTERNAL DOSE FOR MONITORED WORKERS

Page 33 of 50
Document No. ORAUT-TKBS-0008-5 Revision No. 

0

3 Effective Date: 08/17/2012 Page 100 of 116

Test name Date Type

Purpose & 

yield Sponsor Location

Radionuclides 

identified In 

release

Release type & 

notes Release (Ci) Release summary

Diagonal 

Line 

11/24/71 Shaft Weapons 

effects

<20 kt

LLNL

DOD

U11g Kr-85m, Kr-87, 

Kr-88, I-131, 

I-132, I-133, 

I-135, Xe-131m, 

Xe-133,

Xe-133m, 

Xe-135

Test and 

seepage

Accidental 

release of 

radioactivity 

detected offsite 

by aircraft only

Containment 

failurea

Test release 

and seepage at 

R+12 hours: 

6,800 

Test release (seepage) 

occurred from H+3.3 to H+20 

hours. Low-level seepage 

continued for about 3 days, but 

all significant activity had been 

released by H+20 hours. 

Effluent was primarily Xe-135

(80-85%), Kr-85m, Kr-87, Kr-88, 

Xe-131m, Xe-133, and 

Xe-133m, with trace quantities 

of I-131, I-132, I-133, and I-135

detected. Minor levels of 

radioactivity were detected 

offsite by aircraft only.

ATTACHMENT A

OCCUPATIONAL INTERNAL DOSE FOR MONITORED WORKERS

Page 34 of 50
Document No. ORAUT-TKBS-0008-5 Revision No. 

0

3 Effective Date: 08/17/2012 Page 101 of 116

Test name Date Type

Purpose & 

yield Sponsor Location

Radionuclides

identified in 

release

Release type & 

notes Release (Ci) Release summary

Hybla Fair 10/28/74 Tunnel Weapons 

effects

<20 kt

LLNL

DOD

U12n.09 Xe-133, 

Xe-133m

Controlled tunnel 

purge (release 

detected onsite 

only)

Controlled 

release activity 

at time of 

release: 500

Controlled 

release activity 

at R+12 hours: 

500

Release occurred from area 

between over burden plug 

(OBP) No. 1 and OBP No. 2 

from November 13 to November 

14, 1974. Stemming failed 

during test and noble gases 

seeped through or around OBP 

No. 2. All activity was 

successfully contained inside 

OBP No. 1. Effluent released 

during controlled ventilation of 

tunnel complex was activity 

contained between OBP No. 1 

and OBP No. 2 only. Activity 

was 99% Xe-133; remainder 

was Xe-133m. Second release 

occurred from U12n.09 drift 

complex from November 20, 

1974, to January 6, 1975. 

Activity released passed 

through HEPA and aerosol filter 

before being released through 

tunnel ventilation system. 

Effluent was 99% Xe-133 with 

some Xe-133m.

ATTACHMENT A

OCCUPATIONAL INTERNAL DOSE FOR MONITORED WORKERS

Page 35 of 50
Document No. ORAUT-TKBS-0008-5 Revision No. 

0

3 Effective Date: 08/17/2012 Page 102 of 116

Test name Date Type

Purpose & 

yield Sponsor Location

Radionuclides 

identified In 

release

Release type & 

notes Release (Ci) Release summary

Hybla Gold 11/1/77 Tunnel Weapons 

effects

<20 kt

LANL

DOD

U12e.20 Xe-133 Controlled tunnel 

purge (release 

detected onsite 

only)

Controlled 

release activity 

at time of 

release: 

5.0E-03

Controlled 

release activity 

at R+12 hours: 

5.0E-03

Activity was successfully 

contained inside drift protection 

plug until it had decayed to 

insignificant level. Controlled 

release occurred on November 

29, 1977.

Riola 9/25/80 Shaft Weaponsrelated

1.07 kt

LLNL U2eq Kr-85m, Kr-87, 

Kr-88, Xe-133, 

Xe-133, Xe-135, 

Xe-135m, 

tritium, and 

tritiated water

Test, seepage, 

and gas 

sampling

Release 

detected offsite 

(test only)

Containment 

failurea

Gas sampling 

release at time of 

release, in 

curies: 9.8

H-3 in curies: 9.8

Kr-85 in curies: 

1.5 E-04

Test release at 

R+12 hours: 

960 (mixed 

fission products)

Natural 

seepage at time 

of release: 2200 

(tritium and 

tritiated water)

Test release and seepage from 

surface ground zero area 

occurred at H+10 hours and 59 

minutes. Test release, 

consisting of xenons and 

kryptons, occurred through 

surface ground zero cracks and 

lasted until 1020 hours on 

September 26, 1980. Seepage 

continued until it was no longer 

positively quantified in March 

1981. Seepage rate varied 

throughout period as it was 

affected by atmospheric 

pressure changes.

Controlled gas sampling 

containment tank release 

occurred on December 6, 1982.

ATTACHMENT A

OCCUPATIONAL INTERNAL DOSE FOR MONITORED WORKERS

Page 36 of 50
Document No. ORAUT-TKBS-0008-5 Revision No. 

0

3 Effective Date: 08/17/2012 Page 103 of 116

Test name Date Type

Purpose & 

yield Sponsor Location

Radionuclides 

identified In 

release

Release type & 

notes Release (Ci) Release summary

Miners Iron 10/31/80 Tunnel Weapons 

effects

<20 kt

LANL

DOD

U12n.11 Xe-133, Xe-135 Controlled tunnel 

purge (release 

detected onsite 

only)

Controlled 

release activity 

at time of 

release: 

3.0E-01

Controlled 

release activity 

at R+12 hours: 

1.0E-01

Controlled release occurred 

from H+49.5 hours until H+67 

hours. Before that time, 

seepage from stemming area 

into open part of LOS pipe had 

occurred. Effluent was 87% 

Xe-135 and 13% Xe-133. 

Activity was contained in LOS 

pipe until controlled ventilation 

of pipe was established. 

Release point was N Tunnel 

mesa vent hole.

Huron 

Landing

(Simultaneo

us with 

Diamond 

Ace Test)

9/23/82 Tunnel Weapons 

effects

<20 kt

LLNL

DOD

U12n.15 Kr-85m, Kr-88, 

Xe-133, Xe-133, 

Xe-135

Controlled tunnel 

purge (release 

detected onsite 

only)

Controlled 

release activity 

at time of 

release: 280

Controlled 

release activity 

at R+12 hours: 

120

Controlled ventilation of tunnel 

occurred from H+27.8 hours 

until H+36 hours. Before that 

time, activity had been 

contained in OBP until 

ventilation to mesa could be 

established. Release point was 

N Tunnel mesa vent hole. 

Effluent was 86% Xe-135, 7% 

Kr-85m, 3% Xe-133, 3% 

Xe-133, and 1% Kr-88.

Mini Jade 5/26/83 Tunnel Weapons 

effects

<20 kt

LANL

DOD

U12n.12 Xe-133, Xe133mControlled tunnel 

purge (release 

detected onsite 

only)

Controlled 

release activity 

at time of 

release: 1.0

Controlled 

release activity 

at R+12 hours: 

1.0

Controlled ventilation occurred 

from H+5.2 days until H+6.2 

days. Before that, activity had 

been contained inside the drift 

protection plug (DPP) until 

ventilation to mesa had been 

established. Release point was 

N Tunnel mesa vent hole. 

Effluent was 89% Xe-133 and 

11% Xe-133.

ATTACHMENT A

OCCUPATIONAL INTERNAL DOSE FOR MONITORED WORKERS

Page 37 of 50
Document No. ORAUT-TKBS-0008-5 Revision No. 

0

3 Effective Date: 08/17/2012 Page 104 of 116

Test name Date Type

Purpose & 

yield Sponsor Location

Radionuclides

identified in 

release

Release type & 

notes Release (Ci) Release summary

Midas Myth 2/15/84 Tunne

l

Weapons 

effects

<20 kt

LANL

DOD

U12t.04 N/A Unexpected 

crater collapse

N/A Unexpected crater collapse 

occurred 3 hours after test 

above test tunnel causing 

injuries to personnel (1 fatality) 

and damaging equipment 

trailers. All radioactive material 

was contained in vessel with no 

release to atmosphere or tunnel.

Agrini 3/31/84 Shaft Weaponsrelated

<20 kt

LLNL U2ev Kr-85m, Kr-87, 

Kr-88, Xe-133, 

Xe-133, Xe-135, 

Xe-135m

Test, controlled, 

and drill-back

Accidental 

release of 

radioactivity 

detected onsite 

only

Containment 

failurea

Test release at 

R+12 hours: 

690

Controlled 

release activity 

at time of 

release: 

3.0E-02

H-3: 2.8E-02

Xe-133: 2.8E-04

Ar-37: 1.6E-03

Drill-back 

release activity 

at time of 

release: 

2.0E-03

Xenons: 

2.0E-03

Releases occurred as follows:

1. Seepage from crater from 

1530 hours on March 31, 

1984, to 1900 hours on April 

1, 1984;

2. Controlled, filtered release 

on June 13, 1984; and

3. Ventilation line release at 

0705 hours on April 5, 1984, 

during post-shot drilling 

operations.

ATTACHMENT A

OCCUPATIONAL INTERNAL DOSE FOR MONITORED WORKERS

Page 38 of 50
Document No. ORAUT-TKBS-0008-5 Revision No. 

0

3 Effective Date: 08/17/2012 Page 105 of 116

Test name Date Type

Purpose & 

yield Sponsor Location

Radionuclides 

identified In 

release

Release type & 

notes Release (Ci) Release summary

Kappeli 7/25/84 Shaft Weaponsrelated

20-150 kt

LLNL U20am Kr-85 Late-time seep 

(release detected 

onsite only)

Natural, latetime seepage at 

time of release: 

12

Seepage began months after 

test and continued as follows:

9/24/84 - 12/31/84: 0.5 Ci of Kr85

11/25/84 - 7/25/85: 3.6 Ci of Kr85

7/25/85 - 7/25/86: 5.0 Ci of Kr85

Tierra 12/15/84 Shaft Weaponsrelated

20-150 kt

LLNL U19ac Xe-133, 

Xe-131m, Kr-85, 

Ar-37 

Late-time seep 

(release detected 

onsite only)

Natural, latetime seepage at 

time of release: 

600

Xe-133: 

5.7E+02

Xe-131m: 4.0

Kr-85: 12

Ar-37: 9.0

Seepage occurred intermittently 

from December 26, 1984, to 

January 4, 1986.

Misty Rain 4/6/85 Tunnel Weapons

effects

<20 kt

LLNL

DOD

U12n.17 Xe-133, Xe-133, 

Xe-135

Controlled tunnel 

purge

Controlled 

release of 

radioactivity 

detected offsite

Controlled 

release activity 

at time of 

release: 63

Controlled 

release activity 

at R+12 hours: 

45

Controlled ventilation occurred 

from H+2.85 days until H+4 

days. Before that, activity had 

been contained inside gas seal 

plug until ventilation could be 

reestablished. Release points 

were N Tunnel portal and N 

Tunnel mesa ventilation lines. 

Effluent was 72% Xe-133, 22% 

Xe-135, and 6% Xe-133.

ATTACHMENT A

OCCUPATIONAL INTERNAL DOSE FOR MONITORED WORKERS

Page 39 of 50
Document No. ORAUT-TKBS-0008-5 Revision No. 

0

3 Effective Date: 08/17/2012 Page 106 of 116

Test name Date Type

Purpose & 

yield Sponsor Location

Radionuclides 

identified In 

release

Release type & 

notes Release (Ci) Release summary

Mill Yard 10/9/85 Tunnel Weapons

effects

<20 kt

LANL

DOD

U12n.20 Xe-133, Xe-135, 

Xe-135m

Controlled tunnel 

purge (release 

detected onsite 

only)

Controlled 

release activity 

at time of 

release: 5.9

Controlled 

release activity 

at R+12 hours: 

4.6

Controlled ventilations occurred 

as follows:

1. Controlled ventilation from 

working point side of 

U12n.20 drift was conducted 

from H+1.9 days until H+2.5 

days. Effluent was 80% 

Xe-135, 18% Xe-133, and 

2% Xe-133m.

2. Controlled release occurred 

during ventilation of MILL 

YARD cavity from H+16 

days until H+18 days. 

Effluent was 98% Xe-133 

and 2% Xe-133m.

Diamond 

Beech

10/9/85 Tunnel Weapons 

effects

<20 kt

LLNL

DOD

U12n.19 Xe-133, 

Xe-133m, 

Xe-135

Controlled tunnel 

purge (release 

detected onsite 

only)

Controlled 

release activity 

at time of 

release: 1.1

Controlled 

release activity 

at R+12 hours: 

1.0

Controlled ventilations occurred 

as follows:

1. Ventilation of tunnel to portal 

side of U12n.19 DPP 

occurred from H+1.8 days 

until H+2.5 days. Effluent 

was 80% Xe-135, 11% 

Xe-133m, and 9% Xe-133.

2. Ventilation of U12n.19 main 

drift occurred from H+8 days 

until H+9 days. Effluent was 

82% Xe-133 and 18% 

Xe-133m.

ATTACHMENT A

OCCUPATIONAL INTERNAL DOSE FOR MONITORED WORKERS

Page 41 of 50

ATTACHMENT A

OCCUPATIONAL INTERNAL DOSE FOR MONITORED WORKERS

Page 40 of 50
Document No. ORAUT-TKBS-0008-5 Revision No. 

0

3 Effective Date: 08/17/2012 Page 107 of 116

Test name Date Type

Purpose & 

yield Sponsor Location

Radionuclides

identified in 

release

Release type & 

notes Release (Ci) Release summary

Mighty Oak 4/10/86 Tunnel Weapons

effects

<20 kt

LLNL

DOD

U12t.08 Kr-85, I-131, 

Xe-133

Controlled tunnel 

purge (offsite)

Controlled 

release of 

radioactivity 

detected offsite

Controlled 

release activity 

at time of 

release: 

3.6E+04

Controlled 

release activity 

at H+12 hours: 

3.3E+04

Eight controlled ventilations 

occurred.f

Labquark 9/30/86 Shaft Weaponsrelated

20-150 kt

LLNL U19an Xe-133, Kr-85 Late-time 

seepage (onsite 

only)

Natural, latetime seepage 

activity at time 

of release: 16

Xe-133: 2.6

Kr-85: 13

Two late-time releases, due to 

seepage, occurred from 

October 25, 1986 to January 

13, 1987.

Bodie 12/31/86 Shaft Weaponsrelated

20-150 kt

LLNL U20ap Xe-133, 

Xe-133m, 

Xe-131m, Kr-85, 

Ar-37, Xe-135

Drill-back and 

late-time 

seepage (onsite 

only)c

Drill-back 

release activity 

at time of 

release: 1.2

Xe-133: 1.2

Xe-133m: 4.8 E02

Xe-135: 9.8 E04

Natural, latetime seepage 

activity at time 

of release: 50

Xe-133: 44

Xe-133m: 2.0

Xe-131m: 1.0

Kr-85: 2.0

Ar-37: 1.0

Five drill-back releases 

occurred from ventilation line 

from 2307 hours on December 

20, 1986, until 0215 hours on 

December 21, 1986, for total 

release time of 24.3 minutes. 

Seepage occurred continuously 

from December 15, 1986, to 

January 20, 1987, and 

sporadically, depending on 

atmospheric pressure, until 

December 16, 1987.

Mission 6/20/87 Tunnel Weapons LANL U12t.09 Kr-85 Controlled tunnel Controlled Activity was contained in cavity 

ATTACHMENT A

OCCUPATIONAL INTERNAL DOSE FOR MONITORED WORKERS

Page 41 of 50
Document No. ORAUT-TKBS-0008-5 Revision No. 

0

3 Effective Date: 08/17/2012 Page 108 of 116

Test name Date Type

Purpose & 

yield Sponsor Location

Radionuclides

identified in 

release

Release type & 

notes Release (Ci) Release summary

Ghost effects

<20 kt

DOD purge (onsite 

only)c

release at time 

of release: 3.0

Kr-85: 3.0

until ventilation was established 

on December 16, 1987. 

Release continued 

intermittently for about 3 weeks.

Releases from atmospheric tests 1951–1963 1.2 E+10

Other releases from 108 tests from 1970–1988d 5.5 E+03

Table A-26 contains information for NTS underground tests. A similar table is planned for NTS atmospheric tests [Reserved]. The “isotopes identified in the release” are 

from offsite monitoring (Schoengold, DeMarre, and Kirkwood 1996) and are not necessarily the isotopes of concern for workers.

a. Containment failures are normalized to 12 hour after the test.

b. The camphor failure includes Cs-140 from tunnel purging.

c. Bodie and Mission Ghost had drill-back releases.

d. Many operational releases are associated with tests that were not announced.

e. Isotopic analysis from the Baneberry Test Manager’s Rad Safe Advisor Status Report (AEC 1970).

f. MIGHTY OAK eight controlled ventilations occurred as follows:

1. Controlled ventilation from the gas seal plug (GSP) to the DPP was performed from 0950 hours on April 22 to 0611 hours on April 23, 1986. At the time of 

release, 340 Ci of activity were released (calculated to be 316 Curies at P+12).

2. Controlled ventilation of the tunnel complex, work point side of the DPP, was performed from 1040 hours to 1440 hours on April 25. At the time of release, 

3,400 Ci were released (calculated to be 3,200 Ci at P+12).

3. Controlled ventilation of the tunnel complex occurred from 1002 hours on April 28 to 0310 hours on April 29. At the time of release, 9,800 Ci were released 

(calculated to be 9,100 Ci at P+12).

4. Controlled ventilation of the tunnel complex occurred from 1034 hours to 1504 hours on April 29. At the time of release, 1,800 Ci were released (calculated to be 

1,700 Ci at P+12).

5. Controlled ventilation of the tunnel complex occurred from 1422 hours to 1805 hours on April 30. At the time of release, 1,200 Ci were released (calculated to be 

1,100 Ci at P+12).

6. Controlled ventilation of the tunnel complex occurred from 1011 hours to 1937 hours on May 1, 1986. At the time of release, 4,900 Ci were released (calculated 

to be 4,600 Ci at P+12).

7. Controlled ventilation of the tunnel complex occurred from 0946 hours on May 2 to 0450 hours on May 4. At the time of release, 9,000 Ci were released 

(calculated to be 8,400 Ci at P+12).

8. Controlled ventilation of the tunnel complex occurred from 1350 hours on May 5 to 1050 hours on May 19. At the time of release, 5,500 Ci were released 

(calculated to be 5,100 Ci at P+12).

Notes: The total release, at the time of release, was 36,000 Ci; at P+12, the total calculated activity was 33,000 Ci. The total release associated with MIGHTY OAK 

was assumed to be all Xe-133, but during the ventilation period, 2.4 Ci of I-131 and 4.3 Ci of Kr-85 were also released. All ventilations of the tunnel were 

accomplished with the approval of the Test Controller.

H = Time of detonation (NCI 1997). For example, “H + 12” means 12 hours after detonation.

R = Time of release (Schoengold, DeMarre, and Kirkwood 1996).

P = Time of controlled ventilation (“purge”) (Schoengold, DeMarre, and Kirkwood 1996).

ATTACHMENT A

OCCUPATIONAL INTERNAL DOSE FOR MONITORED WORKERS

Page 42 of 50
Document No. ORAUT-TKBS-0008-5 Revision No. 03 Effective Date: 08/17/2012 Page 109 of 116

• A memorandum for B Tunnel, U12b (Johnson 1962), describes the following tritium results for 

tests that occurred in November 1961: 42 workers were exposed above 3 rem (quarterly 

external limit) and 48 workers were exposed above 5 rem (yearly internal and external limit).

• In summary, 108 miners approached the quarterly exposure limit in 1961 due to tritium 

exposure in U12b. Twenty-seven of these employees were subjected to internal radiation 

greater than 3,000 mrem/yr. The original source of tritium was from 1958 activities (Arent and 

Smith 2004).

The Operation Storax On-Site Radiological Safety Report (McClendon and Eubank 1964) 

discusses internal radiation hazards for 1962 and 1963, which includes the YUBA test 

conducted at U12b.10 on June 5, 1963. Urinalysis results showed that 280 workers received 

50 mrem or more, and the maximum internal dosage was 2,435 mrem. Appendix D of this 

report, “Typical Detailed Safety Support Plan for Underground Nuclear Tests,” describes 

radiological support to post-test drilling. Air samples were collected on the drill platform (high 

and low volume) at breathing level. Nasal and urine samples were collected at the discretion 

of the shift supervisor. Appendix A of General Re-Entry Procedure for Underground Nuclear 

Events (LRL 1961) is the specific program protocol used for the YUBA test in 1963, during 

which an accidental release of radioactivity was detected on the site and an operational 

release of radioactivity was detected off the site from U12b (B Tunnel). In summary, nine 

REECo personnel received thyroid doses in excess of 30 rads during YUBA (Arent and Smith 

2004).

• Operation Whetstone, On-Site Radiological Report (Eubank 1965) for 1964 and 1965 

describes exposure to internal radiation up to NTSO-0524-01 limits (AEC 1961). Two posttest drilling personnel received internal doses to the thyroid above the quarterly limit of 10 rem. 

One of these exceeded the annual limit (30 rem) with a dose of 31 rem. In a collection of 

papers entitled REECo Historical Radiation Records, Operation Whetstone (July 1964 – June 

1965), #20 Merlin Event, U3CT a report entitled Report of Investigation of Thyroid 

overexposure - U3ct, Type “B” Incident (REECo 1965) also describes the iodine incident 

resulting in doses of 27 and 31 rem (131I and 133I, respectively) to the thyroid for two employees 

involved in post-test drilling. In all, six employees received measurable thyroid radiation 

exposures. A Type B Investigation was conducted (REECo 1965).

• Extensive records for the 1970 Baneberry test are available in the site research database. 

These include summary reports from the DOE Nevada Operations Office (DOE-NV), REECo 

radiological safety reports and procedures, and records of monitoring and decontamination 

efforts (REECo 1973). On December 18, 1970, the Baneberry underground nuclear test at 

NTS released radioactivity to the atmosphere. The release or venting resulted in a cloud of 

radioactive dust about 10,000 ft above the surface. Levels of radioactivity measured off NTS 

were below radiation guidelines. Approximately 86 employees were exposed to radiation from 

Baneberry, but none received exposure that exceeded the guideline for radiation workers. 

After Baneberry, new containment procedures were adopted to prevent a similar occurrence.

• Evaluation of Protection, Bioassay, and Dose Assessment Programs for Internal Radiation 

Exposures at NTS Particularly as Related to Three Exposure Situations (French and Skrable 

1995) describes internal uptakes at NTS E Tunnel (239Pu), Nellis Air Force Range Double 

Tracks (239Pu) at the Tonopah Test Range, and Building A-1 (Atlas Facility) in Las Vegas (3

H) 

during the mid–1990s. The three situations included:

ATTACHMENT A

OCCUPATIONAL INTERNAL DOSE FOR MONITORED WORKERS

Page 43 of 50
Document No. ORAUT-TKBS-0008-5 Revision No. 03 Effective Date: 08/17/2012 Page 110 of 116

• The exposure of 24 workers to 239Pu aerosols while working in the E Tunnel from June to 

August 1994, possibly from removal of timbers from the walls and roof of the tunnel, which 

might have had high specific activity. The activity of 239Pu in fecal samples of E Tunnel 

workers did not exceed 1 pCi and dose estimates ranged from 0 to 1,337 mrem based on the 

assumption of inhalation intakes of respirable aerosols. The bioassay data from E Tunnel 

workers were limited in the quantity and quality needed for making reasonable accurate 

estimates of their intakes and doses.

• The exposure of one worker in June 1995 to 239Pu aerosols, believed to have been generated 

during the handling and characterization of high-specific-activity fragments found during the 

Double Track Site soils characterization project. French and Skrable (1995) estimated the 

worker at the Double Track Site had an effective ingestion intake of 3,260 pCi (239Pu aerosol) 

from nonrespirable aerosols with a committed effective dose of 0.2 mrem, and an inhalation 

intake of 360 pCi of 1-μm activity median aerodynamic diameter class Y aerosols with a 

committed effective dose of 90 mrem.

• The exposure of workers in 1995 to tritium released from metal foils in Building A-1 at the Atlas 

Facility on Losee Road in North Las Vegas. French and Skrable (1995) concluded that 

considerable bioassay and air sampling data confirmed that the exposures were minimal with 

little dose consequence, and provided no further dose estimates. In 1995, EG&G discovered 

contamination in the same basement as a result of routine bioassay samples from workers. 

The source of contamination was later found to have resulted from a modification of a sealedtube neutron generator by an EG&G scientist. This modification included breaching the sealed 

tube that contained three tritide disks (130 Ci tritium total). According to the Annual Site 

Environmental Report for Calendar Year 1995 (Black and Townsend 1996), the maximum 

release into the building was 123 mCi.

A.3.3 Respiratory Protection Practices

The following paragraphs provide examples of NTS respiratory protection practices. Historical 

accounts of respirator usage are included to give the DR a chronological snapshot of the program that 

was in place during the early testing period. This is not a comprehensive review of respiratory 

protection practices at NTS.

The Operation Ranger Report (LASL 1952) stated that all persons entering a target area where 

background was higher than 3 mR/hr received respirators and that bulldozer operators received 

respirators with plastic hoods.

Project 57 was an atmospheric safety experiment that studied plutonium dispersal in NTS Area 11 

from 1955 to 1956. AEC (1956) stated that respirators or “full-face assault masks” were required for 

persons entering the CA. Nose swipes were required on leaving the CA; urinalysis was required for 

anyone who entered the CA at end of operations. AEC (1956) contains details on alpha air levels with 

workplace air sample results.

A memorandum for the Plumbbob Operation (Nielsen 1957) includes a detailed step-by-step narrative 

for a Saturn Tunnel (U12c) reentry conducted on December 12, 1957. The reentry party was fitted 

with “airpacks.” The memorandum lists measured contamination levels on PPE (up to 106 alpha cpm) 

and includes individual nasal swipes results (sanitized). A second memorandum (Bybee 1957) details 

the August 30, 1957, reentry into U12c during which team members wore respirators and MSA air 

ATTACHMENT A

OCCUPATIONAL INTERNAL DOSE FOR MONITORED WORKERS

Page 44 of 50
Document No. ORAUT-TKBS-0008-5 Revision No. 03 Effective Date: 08/17/2012 Page 111 of 116

packs. Specific results included individual nose swipe results (sanitized), filters from respirators 

measured at 2,000 to 3,000 cpm (after 3 days, the count was the same), and filter paper on the air 

sampler at the blast door that measured >100,000 cpm after a 3-day decay. Urine samples were 

reportedly submitted for plutonium analysis, but the memorandum included no results.

A 1958 REECo memorandum entitled “Decontamination of Tunnel U12f” (Wilcox 1958) discussed 

PPE worn by the reentry teams. The first team wore air-supplied masks and full anticontamination 

clothing. Alpha contamination was measured at >500 c/m/55cm2 in the main drift from Drift 01, and 2 

× 103 c/m/55cm2 in Drift 02 to the blast door. The second team wore full-face MSA masks with allservice canisters and anticontamination clothing. The third decontamination team wore Wilson fullface masks and anticontamination clothing; nasal swabs were collected. The memorandum stated 

“air sampling and nasal swabs indicated no significant air activity during latter stages of 

decontamination and afterwards.”

Atomic Weapon Accidents – Alpha Emitter Dispersal & Decontamination (Wilcox and Coogan 1959) 

described what to expect with an atomic weapon accident. Recommendations were based on NTS 

experience with large land areas contaminated with alpha-emitting material. The CA was defined as 

>2 × 103 dpm/55cm2

. Full-face masks equipped with dust, fume, and mist filters were worn, and nasal 

swabs were collected at the exit of the CA. If nasal swab results were >200 dpm (total both nostrils), 

workers contributed a urine sample. The report stated, “Airborne radioactive material was generally 

below permissible levels, but windy conditions occasionally caused concentration of airborne activity 

to 4 × 10-9 μc/mL. (Resuspension of radioactive particulate was reduced by keeping the area 

moistened with water.) Sample analysis indicated that of the more than 1,000 separate exposures to 

airborne alpha-emitters, no person received significant internal exposure.”

General Re-entry Procedure for Underground Nuclear Events (LRL 1961) described PPE worn during 

tunnel reentry. A McCaa 2-hour self-contained oxygen breathing apparatus and full Radex clothing 

were specified for the reentry party and rescue team. MSA all-service gas masks and full Radex 

clothing were designated for the surface radiological survey party (Rainier shaft). Appendix A of this 

document is the Re-entry Program for the YUBA test in 1963, in which an accidental release of 

radioactivity was detected on the site and an operational release of radioactivity was detected off the 

site from U12b (B Tunnel).

Palanquin Reentry and Recovery Safety Procedure (LRL 1965) stated that all reentry and recovery 

personnel would wear full protective clothing and respiratory protection (full-face mask with MSA 

Model “CMR” cartridge or equivalent). Workers were fit-tested with smoke tubes immediately before 

entry. Bioassays were to be performed if internal exposure to workers was suspected. “Spot” 

bioassay samples were to be taken from representative workers after reentry to verify the 

effectiveness of the control program. Table A-27 lists NTS respiratory protection levels for selected 

years (Author unknown undated a).

A.3.4 Historical Practices and Contamination Levels

The Plumbbob On-Site Radiological Safety Report (REECo 1957) is an example of the magnitude of 

early radiological support; it stated REECo collected the following samples: 4,230 nasal swabs, 109 

facial swipes, and 1,945 air samples. In addition, 73 urine kits and 10,755 respirators were issued.

ATTACHMENT A

OCCUPATIONAL INTERNAL DOSE FOR MONITORED WORKERS

Page 45 of 50
Document No. ORAUT-TKBS-0008-5 Revision No. 03 Effective Date: 08/17/2012 Page 112 of 116

Table A-27. Historical NTS respiratory protection action levels.

Year Respiratory protection action levels

1957 No respirator ≤9.6E+3 dpm/m3 hours (alpha) per quarter year

No respirator (short periods) ≤100 dpm/m3 (alpha)

Respirator required >100 dpm/m3 (alpha)

No respirator ≤2.2E+4 dpm/m3 (beta + gamma)

Respirator required >2.2E_4 dpm/m3 (beta + gamma)

Note: Respirators using filters have an efficiency up to 99.9% for particles as small as 0.3 μm in 

diameter.

1959 Ultra-filter respirator can be worn to levels <500 dpm/m3 (Pu-239 alpha)

Full-face mask with dust, fume, and mist canister required in levels from 500 to 10,000 dpm/m3

(Pu-239 alpha)

Air-supplied mask recommended when levels >10,000 dpm/m3 (Pu-239 alpha)

1968 0 to 100 dpm/m3 (alpha)—no respiratory protection for short-term exposure.

4E10-11 μCi/cm3 (beta, gamma)—respiratory protection required

100 to 100,000 dpm/m3 (alpha)—high-filtration, full-face respirator (99.9% effective)

Greater than the above levels—self-contained breathing apparatus

The Defense Nuclear Agency (DNA) report on the 1957 Plumbbob Operations (Harris et al. 1981) 

describes REECo activities, checkpoint during reentry, Radex area cards, and PPE in Radex areas 

(>100 mR/hr gamma). It also describes radiation safety support for the Project 57 internal alpha 

radiation hazard, and discusses REECo and Sandia surveys and PPE. Nose swipes and nasal 

swabs were analyzed by REECo. Urine samples were packaged and shipped off the site for analysis.

Radiological Safety for Underground Nuclear Explosions (Wilcox and Coogan 1960) describes 

general hazards, nasal swabs and urine samples, respirators, and air monitoring in tunnels. Workers 

received anticontamination clothing (coveralls, head covers, gloves, shoe covers) and respirators. 

Workers leaving the area at the end of the shift were monitored and decontaminated if necessary. Air 

samples were obtained and nasal swabs were taken to determine possible internal exposures. 

Individuals whose nasal swabs measured >3 × 103 dpm (beta, total both nostrils) or 2 × 102 dpm

(alpha, total both nostrils) were requested to submit urine samples. These samples were processed 

in the laboratory to determine body burdens of GFPs or alpha emitters.

“Procedures to Limit Radiological Exposures of Miners” (Reeves 1962) stated the following practices 

were to be implemented for underground operations:

• Anticontamination clothing and respiratory equipment when conditions indicate they are 

necessary; impermeable clothing when the situation warrants.

• Urine samples for specific activities collected on a routine basis from personnel working in 

areas where it is possible to receive internal exposures; results kept on cumulative records 

and included in the daily estimated exposure.

• Film, dosimeter, and internal exposure results added together to determine the daily estimated 

dose of each person; forwarded to Superintendent before shift (rotate men).

• Low-pressure weather – various radioactive gases seep into work areas from cracks. 

Evacuate area.

ATTACHMENT A

OCCUPATIONAL INTERNAL DOSE FOR MONITORED WORKERS

Page 46 of 50
Document No. ORAUT-TKBS-0008-5 Revision No. 03 Effective Date: 08/17/2012 Page 113 of 116

• Increased and expanded vent system to clear radioactive gases and reduce chance for 

internal exposure.

• One-side drift engineering of new tunnels.

• Table A-28 lists NTS contamination limits for selected years. Contamination limits were often 

used as indicators of when bioassay samples were to be taken.

Table A-28. NTS historical contamination limits.

Year Contamination limits

1957 Personnel: 1 mR/hr (gamma) and 100 cpm/55 cm2 (alpha)

Protective clothing: 7 mR/hr (beta + gamma) and 500 cpm/55 cm2 (alpha)

Respiratory devices: 1 mR/hr (beta + gamma) and 100 cpm/55 cm2 (alpha)

1958 Coveralls must be removed for eating if alpha levels on coveralls are >1E+3 dpm/55 cm2.

Decontaminate personnel:

• Outer clothes: >500 cpm/55 cm2 (alpha) or 7 mR/hr (beta + gamma)

• Shoes: >500 cpm/55 cm2 (alpha) or 7 mR/hr (gamma)

• Skin or underclothing: >100 cpm/55 cm2 (alpha) or 1 mR/hr (gamma)

• Equipment: <7 mR/hr (gamma) or 500 cpm/55 cm2 (fixed alpha)

Respiratory devices: <1 mR/hr (beta + gamma) or 100 cpm/55 cm2 (alpha fixed and removable)

Alpha contamination >5,000 cpm/55 cm2 requires field decontamination.

Access permits required to enter areas with beta-gamma levels >10 mR/hr or alpha levels 

>500 cpm/55 cm2.

1959 Nasal swab results >200 dpm (total for both nostrils) required urine samples. Outer pair of coveralls 

removed to eat if contaminated to levels in excess of 500 cpm/55 cm2. 

1963 Laundry items >1 × 104 dpm/55 cm2 (alpha) or 20 mrad/hr (beta-gamma) are disposed of or retained 

until radioactive decay reduces activity levels. Clothing and equipment are released for reissue after 

laundering if <7 mrad/hr (beta-gamma) or 1,000 dpm/55 cm2 (alpha) for coveralls and 1 mrad/hr (betagamma) or 200 dpm/55 cm2 (alpha) on respirators.

1970 Anticontamination coveralls must be free of loose contamination and below 100 cpm alpha before 

reuse. Beginning in 1959, studies were conducted to determine effects of environmental forces on 

shifting and resuspension of deposited alpha contamination and soil penetration since initial deposition 

or fixation (by windrowing). 

Source: Author unknown (undated a).

A.4 REFERENCE TABLES FOR DETERMINING INTERNAL DOSE

At NTS, which is a large outdoor testing facility, there was never a process that carried over from year 

to year; these differences have been identified as test categories in this document. DRs should 

assume that acute exposure occurred within one work shift, when single bioassay results are included 

in the DOE records for the worker. More than 1,000 different tests have occurred at NTS. Very few 

workers have been directly involved in these tests and they have been identified in the records 

provided by the DOE file for each claim. These are the workers who should be considered for 

potential iodine dose, tritium dose, or other unique internal dose consideration if their exposure can be 

linked to a specific release (see Table A-26).

Note: If the file provided by DOE does not include bioassay, DRs should base the internal dose 

assigned to the worker on the environmental ambient.

Tables A-29 through A-31 provide information about the characteristics of the source term at NTS. 

The DR needs to remember that this source term changes over the years based on the test category. 

ATTACHMENT A

OCCUPATIONAL INTERNAL DOSE FOR MONITORED WORKERS

Page 47 of 50
Document No. ORAUT-TKBS-0008-5 Revision No. 03 Effective Date: 08/17/2012 Page 114 of 116

Information has been provided about the radionuclides of concern as the primary focus; however 

other radionuclides are present depending on the timeframe and test category that the DR might have 

to address. The DR should use professional judgment when including or deleting specific 

radionuclides in the source term based on the test category with which the worker was associated and 

the cancer type.

Table A-29. Inhalation classes for radionuclides found at 

NTS.

Radionuclide(s) Allen et al. 1993 BN 1998 BN 2003

Ac-227 D - -

Am-241 W W W

Am-243 W - -

Ba-133 D - -

Ba-140 D - -

Ce-139 Y - -

Ce-141 Y - -

Ce-143 Y - -

Ce-144 Y - -

Cs-134 D - -

Cs-137 D D D

Co-57 Y - -

Co-60 Y W, Y

Cm-244 W - W

Eu-152 W - -

Eu-154 W - -

Eu-155 W - -

I-131 D D D

I-132 D - -

I-133 D - -

Ir-192 Y - -

Lu-174 W - -

Mn-54 W - -

Mo-99 Y - -

Na-22 D - -

Na-24 D - -

Nb-95 Y - -

Np-237 W - -

Pu-238 Wa W, Y W, Y

Pu-239 Wa W, Y W, Y

Pu-240 Wa - -

Pu-241 Wa - -

Pu-242 Wa - -

Ra-226 W W W

Rh-101 Y - -

Rh-102m Y - -

Ru-103 Y - -

Ru-106 Y - -

Sb-123 W - -

Sb-125 W - -

Sr-85 Y - -

Sr-90 (d) Y D, Y D, Y

ATTACHMENT A

OCCUPATIONAL INTERNAL DOSE FOR MONITORED WORKERS

Page 48 of 50
Document No. ORAUT-TKBS-0008-5 Revision No. 03 Effective Date: 08/17/2012 Page 115 of 116

Radionuclide(s) Allen et al. 1993 BN 1998 BN 2003

Ta-182 Y - -

Te-132 D - -

Th-228 W - -

Th-230 W - -

Th-232 W - W, Y

U-233 W - -

U-234b W, Yc D, W or Y D, W or Y

U-235b W, Y D, W or Y D, W or Y

U-238b W, Y D, W or Y D, W or Y

Zr-95 D - -

a. Class W was assigned for all plutonium radionuclides based on 

Pu-240, which has the lowest annual limit on intake (Allen et al. 

1993).

b. Class D was listed in the source documents and may be 

applicable to personnel working in laboratories with soluble 

forms of uranium.

c. Assumed based on information for U-235/238.

d. For X-10, guidance has been provided to the DRs indicating 

that Type Y (current Type S) should only be assigned when 

strontium titinate is identified, if this compound cannot be 

confirmed then Type D (current Type F) should be assumed. 

Table A-30. Fission products up to 1 year old as 

identified in 1954 documentation.

Radionuclide Half-life Beta energy (MeV)

Strontium-89 53 days 1.463

Strontium-91 9.7 hours 2.665 - 0.62

Strontium-90

Yttrium-90

19.9 years

61 hours

0.61

2.18

Yttrium-91 61 days 1.55

Yttrium-92 3.6 hours (?) 3.5

Yttrium-93 10 hours 3.1

Zirconium-95 65 days 0.84–0.371

Zirconium-97 17.0 hours 1.91

Niobium-95 35 days 0.16

Niobium-97 72.1 minutes 1.4

Molybdenum-99 67 hours 1.23–0.45

Ruthenium-103 39.8 days 0.22–0.7

Ruthenium-106 1.0 year 0.039

Rhodium-106 30 seconds 3.55–2.30

Rhodium-103 57 minutes ?

Rhodium-105 36.5 hours 0.570–0.25

Tellurium-132 77.7 hours 1.3–2.4

Iodine-131 8.14 days 0.815-0.250

Iodine-132 2.4 hours 2.2–0.9

Iodine-133 20.5 hours 1.3–0.4

Iodine-135 6.68 hours 1.4 0.5

Xenon-133 5.27 days 0.34

Xenon-135 9.13 hours 0.9

Cesium-137 33 years 0.51

Barium-137m 2.6 months -

Barium-140 12.8 days 1.02–0.47

ATTACHMENT A

OCCUPATIONAL INTERNAL DOSE FOR MONITORED WORKERS

Page 49 of 50
Document No. ORAUT-TKBS-0008-5 Revision No. 03 Effective Date: 08/17/2012 Page 116 of 116

Radionuclide Half-life Beta energy (MeV)

Lanthanum-140 40 hours 1.3–2.26

Lanthanum-141 3.7 hours 2.43–0.9

Cerium-141 33.1 days 0.58–0.442

Cerium-143 33 hour 1.39–0.71

Cerium-144 282 day 0.3–0.17

Praseodymium-144 17.5 minutes 2.97

Praseodymium-143 13.7 day 0.92

Neodymium-147 11.3 day 0.83–0.38

Promethium-147 2.6 year 0.22

Promethium-149 54 hour 1.05

Source: ORDOSE (2003).

Table A-31. Other common radionuclides not 

normally a part of fission products that might be 

present.

Radionuclide Half-life Beta energy (MeV)

Curium-242 162 day 6.1

Neptunium-239 2,433 day 0.715–0.33

Americium-243 10 year 5.27

Plutonium-241 14 year 0.02

Source: ORAUT (2010a).

NOTE: The International Commission on Radiation Protection (ICRP) Publication 30 (ICRP 1979) 

respiratory tract model describes the relative rate of clearance from the pulmonary region of the lungs. 

There are three defined inhalation classes for materials: D (clearance half-time less than 10 days), W 

(10 to 100 days), and Y (greater than 100 days). (Allen et al. 1993 defines D as days, W as weeks, 

and Y as years.)

The ICRP Publication 66 (1994) respiratory tract model describes the relative speed of dissolution and 

translocation to blood of material within the respiratory tract and is related to the physical chemistry of 

the material. There are three defined absorption types: F (fast), M (moderate), and S (slow). For the 

purpose of this project, the DR should assume that D is equivalent to F, M is equivalent to W, and S is 

equivalent to Y.

ATTACHMENT A

OCCUPATIONAL INTERNAL DOSE FOR MONITORED WORKERS

Page 50 of 50