ORAU TEAM

Dose Reconstruction

Project for NIOSH

Oak Ridge Associated Universities I Dade Moeller I MJW Technical Services

Page 1 of 113

DOE Review Release 04/21/2010

Document Title:

Nevada Test Site – Occupational Internal Dose

Document Number: ORAUT-TKBS-0008-5

Revision: 02

Effective Date: 04/15/2010

Type of Document: TBD

Supersedes: Revision 01

Subject Expert(s): Eugene M. Rollins

Site Expert(s): N/A

Approval: Signature on File Approval Date: 04/13/2010

Eugene M. Rollins, Document Owner

Concurrence: Signature on File Concurrence Date: 04/13/2010

John M. Byrne, Objective 1 Manager

Concurrence: Keith A. McCartney Signature on File for Concurrence Date: 04/14/2010

Edward F. Maher, Objective 3 Manager

Concurrence: Signature on File Concurrence Date: 04/14/2010

Kate Kimpan, Project Director

Approval: Signature on File Approval Date: 04/15/2010

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. 02 Effective Date: 04/15/2010 Page 2 of 113

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:

Eugene M. Rollins, TBD Team Leader

Signature on File 07/17/2006

John M. Byrne, Task 3 Manager

Signature on File 07/17/2006

Edward F. Maher, Task 5 Manager

Signature on File 07/18/2006

Kate Kimpan, Project Director

Signature on File 07/19/2006

James W. Neton, Associate Director for Science

Signature on File 07/20/2006

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. 02 Effective Date: 04/15/2010 Page 3 of 113

TABLE OF CONTENTS

SECTION TITLE PAGE

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

5.1 Introduction ............................................................................................................................10

5.1.1 Purpose......................................................................................................................11

5.1.2 Special Exposure Cohort Petition Information for the Nevada Test Site...................... 11

5.1.3 Scope.........................................................................................................................12

5.1.4 Historic Monitoring Perspective .................................................................................. 13

5.1.5 Radionuclides of Concern........................................................................................... 14

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

5.2 In Vitro Minimum Detectable Activities and Counting Methods...............................................17

5.2.1 In Vitro Urine and Fecal Analysis................................................................................ 18

5.2.2 In Vitro Methods for Individual Radionuclides ............................................................. 18

5.2.2.1 In Vitro Bioassay for Iodine........................................................................ 19

5.2.2.2 In Vitro Bioassay for Americium................................................................. 19

5.2.2.3 In Vitro Bioassay for Plutonium ................................................................. 20

5.2.2.4 In Vitro Bioassay for Tritium ...................................................................... 21

5.2.2.5 In Vitro Bioassay for Uranium.................................................................... 22

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

5.2.2.7 In Vitro Analysis for Thorium ..................................................................... 24

5.2.2.8 In Vitro Analysis for Radium ...................................................................... 24

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

5.2.2.10 In Vitro Analysis for Gamma Emitters........................................................ 26

5.2.3 Sample Collection and Correcting for Urinalysis Volume ............................................ 27

5.2.4 Fecal Sample Collection and Analysis ........................................................................ 29

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................................................ 34

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

5.3.3 Thyroid Counting ........................................................................................................ 39

5.3.3.1 Iodine ........................................................................................................ 39

5.3.4 Wound Counting.........................................................................................................40

5.4 Personal Air Sampling Data ...................................................................................................40

5.4.1 Drilling Operations...................................................................................................... 42

5.4.2 Tunnel Operations...................................................................................................... 44

5.4.3 Examples of Air Sampling Results.............................................................................. 45

5.5 Interferences and Uncertainty ................................................................................................48

5.5.1 Contamination of Samples.......................................................................................... 48

5.5.2 Uncertainty .................................................................................................................48

5.5.3 Less-Than Values....................................................................................................... 48

5.6 Dose Reconstruction Recommendations ...............................................................................49

5.6.1 Evaluation of Potential Internal Worker Exposure...................................................... 49

5.6.2 Workers with Pre-1963 Employment........................................................................... 49

5.6.3 Workers with Employment From 1963 through 1992 .................................................. 50
Document No. ORAUT-TKBS-0008-5 Revision No. 02 Effective Date: 04/15/2010 Page 4 of 113

5.6.4 Approach to Internal Dose Reconstruction for NTS .................................................... 50

5.6.4.1 Single Positive Result................................................................................ 51

5.6.4.2 For Individuals with Urine Bioassay Results (Positive or <MDA) ............... 52

5.6.4.3 For Individuals with Whole Body Count Results ........................................ 53

5.6.4.4 For Individuals with Lung Count Results.................................................... 53

5.6.5 Examples of the Assignment of Internal Dose for Selected Job Titles......................... 53

5.6.6 Unusual or Unique Working Environments ................................................................. 54

5.6.6.1 Potential for Iodine Exposure .................................................................... 54

5.6.6.2 Cleanup after Safety Tests........................................................................ 55

5.6.6.3 Nuclear Reactor Development Station (NRDS) ......................................... 55

5.6.6.4 Incidents.................................................................................................... 55

5.7 Attributions and Annotations ..................................................................................................55

References ........................................................................................................................................56

Glossary ............................................................................................................................................64

References ...................................................................................................................................... 110

ATTACHMENT A, OCCUPATIONAL INTERNAL DOSE FOR MONITORED WORKERS ................. 68

A.1 Bioassay Codes and In Vitro Minimum Detectable Activities and Detection Levels ................ 69

A.1.1 Codes and Special Terminology Used in Bioassay Records....................................... 69

A.1.2 In Vitro Analyses for Individual Radionuclides ............................................................ 74

A.2 In Vivo MDAS and Reporting Practices at NTS......................................................................74

A.2.1 Whole-Body Counting................................................................................................. 74

A.2.2 Chest Counting........................................................................................................... 81

A.2.3 Thyroid Counts ........................................................................................................... 82

A.3 Area and Associated Radionuclides of Concern.....................................................................83

A.3.1 Radionuclides of Concern and Specific Bioassay Programs for NTS Facilities ........... 83

A.3.1.1 Drill-backs (Yucca Flat and Pahute Mesa, or Areas 1-10 and 18-20) ........ 84

A.3.1.2 Reentry and Mine Back (Area-1 & Area-12) .............................................. 86

A.3.1.3 Routine Tunnel Operations (Areas 1, 12, 15, and 16)................................ 88

A.3.1.4 Decontamination Facility (Area 6).............................................................. 88

A.3.1.5 Test Treatability Facility (Area 25)............................................................. 88

A.3.1.6 Atmospheric Weapon Safety Tests (Area 3 and Area 11) ......................... 89

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)............................ 89

A.3.1.8 Low-Level Waste Site (Area 3).................................................................. 90

A.3.1.9 Low-Level Waste Site (Area 5).................................................................. 90

A.3.1.10 Radiation Instrument Calibration Facilities (Area 6 and Area 23)............... 90

A.3.1.11 Radiography Operations (Area 23)............................................................ 90

A.3.1.12 Well Logging Operations (Areas 1–10 and Areas 18–20).......................... 91

A.3.1.13 Nuclear Explosive and Device Assembly Facilities (Area 6 and Area 

27)............................................................................................................. 91

A.3.1.14 Nuclear Rocket Development Areas (Area 25 and Area 26)...................... 91

A.3.1.15 Radioactive Source Storage Areas (Area 6 and Area 23).......................... 91

A.3.1.16 Radiochemistry and Counting Laboratories............................................... 92

A.3.1.17 Other Facilities Identified after 1992 (the end of testing)............................ 92
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A.3.1.18 Summary lists of radionuclides of concern ................................................ 93

A.3.1.19 Atmospheric Testing Era ........................................................................... 94

A.3.2 Incidents.....................................................................................................................94

A.3.3 Respiratory Protection Practices at NTS................................................................... 103

A.3.4 Historical Practices and Contamination Levels ......................................................... 104

A.4 Reference Tables for Determining Internal Dose.................................................................. 106

References ...................................................................................................................................... 110
Document No. ORAUT-TKBS-0008-5 Revision No. 02 Effective Date: 04/15/2010 Page 6 of 113

LIST OF TABLES

TABLE TITLE

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

PAGE

15

5-2 2003 target minimum detectable activities for in vitro bioassay sample 

analysis..................................................................................................................................19

5-3 Minimum detectable activities for the Helgeson 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........................................................... 32

5-7 Whole-body count sensitivities ............................................................................................... 33

5-8 Examples of whole-body count ‘alert’ levels ........................................................................... 33

5-9 Mean body burdens of 137Cs from fallout in the United States ................................................ 35

5-10 Uranium L X-ray intensity for decay of 239Pu........................................................................... 37

5-11 Background and efficiency data for 239Pu detection using Phoswich 

detectors ................................................................................................................................37

5-12 Background and efficiency data for 241Am detection using Phoswich 

detectors ................................................................................................................................38

5-13 Relative abundance of radioiodides ....................................................................................... 40

5-14 Air sample results from 1957 surface plutonium dispersal test ............................................... 46

5-15 Additional air sampling data, Area 9 balloon test site ............................................................. 46

5-16 Historic air sampling limits...................................................................................................... 47

A-1 Codes for analyte...................................................................................................................70

A-2 Codes for body parts..............................................................................................................70

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

A-4 Codes for sample types ......................................................................................................... 72

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

A-6 Special terminology................................................................................................................73

A-7 Limits of detection for urine and fecal analysis ....................................................................... 75

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

A-9 Whole-body counting MDAs................................................................................................... 79

A-10 1993 MDAs for chest counting ............................................................................................... 82

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

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

identification versus time after test ......................................................................................... 85

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............................... 89

A-15 Atmospheric weapons test areas, isotopes for identification and of 

concern for dose ....................................................................................................................89

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

dose.......................................................................................................................................90

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

dose.......................................................................................................................................90

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

concern for dose ....................................................................................................................91

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

concern for dose ....................................................................................................................91
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A-20 Radiochemistry and counting laboratories, isotopes for identification and 

of concern for dose ................................................................................................................ 92

A-21 Radionuclides of concern for dose, summary list ................................................................... 93

A-22 Current nuclides of concern for NTS locations ....................................................................... 93

A-23 Fission products associated with above ground testing.......................................................... 94

A-24 Activation products associated with above ground testing...................................................... 95

A-25 Actinide elements associated with above ground testing........................................................ 95

A-26 Releases from underground tests .......................................................................................... 97

A-27 Historical NTS respiratory protection action levels................................................................ 104

A-28 NTS historical contamination limits....................................................................................... 105

A-29 Inhalation classes for radionuclides found at NTS................................................................ 107

A-30 Fission products up to 1 year old as identified in 1954 documentation ................................. 108

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

might be present .................................................................................................................. 109

LIST OF FIGURES

FIGURE TITLE

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

PAGE

78

A-2 NaI detector efficiency............................................................................................................ 79
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ACRONYMS AND ABBREVIATIONS

AMAD activity-median aerodynamic diameter

ANSI American National Standards Institute

BEEF Big Explosive Experimental Facility

BN Bechtel Nevada, Inc. 

CA contamination area

CAM continuous air monitor

cc cubic centimeter

cfm cubic feet per minute

CFR Code of Federal Regulations

Ci curie

cpm counts per minute

CWT chest wall thickness

DAC Derived Air Concentration

DL Decision Level

DOD U.S. Department of Defense

DOE U.S. Department of Energy

DPP drift protection plug

dpm disintegrations per minute

DR dose reconstructor

DU depleted uranium

EE Energy Employee

EG&G Edgerton, Germeshausen, and Greer Corporation 

FSN Fenix & Scisson of Nevada 

FWHM Full-Width-Half-Maximum

g gram

GeLi germanium-lithium detector

GFP gross fission product

H&N Holmes & Narver

HEPA high-energy particulate air (filter)

HEU highly enriched uranium

hr hour

ICRP International Commission on Radiation Protection

in. inch

IREP Interactive RadioEpidemiological Program

keV kilo electron volts

kt kiloton

L liter

LANL Los Alamos National Laboratory

LASL Los Alamos Scientific Laboratory

LLD Lower Limit of Detection

LLNL Lawrence Livermore National Laboratory
Document No. ORAUT-TKBS-0008-5 Revision No. 02 Effective Date: 04/15/2010 Page 9 of 113

LRL Lawrence Radiation Laboratory

m meter

MDA Minimum Detectable Activity; Minimum Detectable Amount

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

mrem millirem

MSA Mine Safety Appliances Company

NaI sodium iodide

NBS National Bureau of Standards

nCi nanocurie

NRDS Nuclear Reactor Development Station

NTS Nevada Test Site

OBP overburden plug

OCAS Office of Compensation Analysis and Support

pCi picocurie 

POC probability of causation

PPE personnel protection equipment

RAMATROL Radioactive Material Control

RAS retrospective air sampler

RCT Radiological Control Technician (also called Radiation Safety Monitor)

REECo Reynolds Electrical & Engineering Company, Inc.

RSN Raytheon Services Nevada

SEC Special Exposure Cohort

SMS Site Monitoring Services

SNL Sandia National Laboratories

SRDB Site Research Data Base

TBD technical basis document 

TRU transuranic

TTR Tonopah Test Range

U.S.C. United States Code

WBC Whole-Body Counter

WEF Waste Examination Facility 

WSI Wackenhut Services, Inc.

μCi microcurie
Document No. ORAUT-TKBS-0008-5 Revision No. 02 Effective Date: 04/15/2010 Page 10 of 113

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 facility” as defined in the Energy Employees 

Occupational Illness Compensation Program Act of 2000 [(42 U.S.C. Sections 7384l (5) and (12)]). 

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

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

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

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

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

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

facility encompassed by EEOICPA.

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

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

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

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

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

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

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

determined in accordance with the POC [probability of causation1

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

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

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

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

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

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

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

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

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

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

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

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

occupationally derived (NIOSH 2007a):

] guidelines established under 

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

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

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

• 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. 02 Effective Date: 04/15/2010 Page 11 of 113

5.1.1

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.

Purpose

5.1.2

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

Special Exposure Cohort Petition Information for the Nevada Test Site

Classes Added to the SEC

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

worked at the Nevada Test Site from January 27, 1951 through December 31, 1962 for a 

number of work days aggregating at least 250 work days, either solely under this employment 

or in combination with work days within the parameters (excluding aggregate work day 

requirements) established for other classes of employees included in the SEC, and who were 

monitored or should have been monitored (NIOSH 2006). 

Classes Recommended by NIOSH for Addition to the SEC

• 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 (NIOSH 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 [HASL 1972] 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.

NIOSH believes that there is insufficient information to adequately support bounding internal dose 

(reconstructing internal dose with sufficient accuracy) for the portion of the SEC-00084 NTS worker 

class who worked during the period of testing from 1963 through 1992 (NIOSH 2010). NIOSH also 

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. Based on this 

information, NIOSH has identified the need to modify its position originally stated in the SEC-00084

NTS Evaluation Report (NIOSH 2007b). NIOSH recommends including all employees of DOE, its 

predecessor agencies, and its contractors and subcontractors who worked at NTS from January 1, 

1963, through December 31, 1992, for a number of workdays aggregating at least 250 workdays, 

occurring either solely under this employment or in combination with workdays within the parameters 

established for one or more other classes of employees in the SEC. Although NIOSH recommends 

adding this portion of the class to the SEC, it intends to use available internal and external data for the 

recommended period (that can be interpreted using its existing processes or procedures) for partial 

dose reconstructions.
Document No. ORAUT-TKBS-0008-5 Revision No. 02 Effective Date: 04/15/2010 Page 12 of 113

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

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

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

dose for the proposed class, 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.

5.1.3

The Nevada Test Site (NTS) is unique in the U.S. Department of Energy (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.

Scope

Reynolds Electrical & Engineering Company, Inc. (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 prior to 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 (REECo 1993b). Bechtel Nevada (BN), was the site 

contractor from 1995 to 2006, and as such had responsibility for the bioassay program (BN 1998, 

2000, 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 regarding authorized external limits for certain job titles would be found in the Radiation 

Safety Plan. The 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 (DeMarre 2004a). 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 (REECo 1993a). Xenon, an external dose concern that is discussed in ORAUT (2007a), could 

have been present in significant quantities several days after a test.
Document No. ORAUT-TKBS-0008-5 Revision No. 02 Effective Date: 04/15/2010 Page 13 of 113

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, Inc.; Johnson Controls Nevada, 

Inc.; and Wackenhut Services, Inc. (WSI). Former NTS contractors included Holmes & Narver (H&N), 

1956–1990; Fenix & Scisson of Nevada (FSN), Inc., 1963–1990; Edgerton, Germeshausen, and 

Greer Corporation (EG&G), 1951–1995; REECo, 1953–1995; and Raytheon Services Nevada (RSN), 

1990–1995. 

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 1963 to the present.

5.1.4

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 (REECo unknown a): 

Historic Monitoring Perspective

“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,” [see 

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. 
Document No. ORAUT-TKBS-0008-5 Revision No. 02 Effective Date: 04/15/2010 Page 14 of 113

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 

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 above-ground testing ended (NAS 

2003). ORAUT (2008a) 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 (2008a) characterize the 

radiological exposure for the majority of the NTS work force.

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

sufficient personnel monitoring, air monitoring, or source term data to adequately reconstruct the 

internal exposures at NTS during the January 27, 1951 through December 31, 1962 time period. 

Consequently, NIOSH finds that it is not feasible to estimate with sufficient accuracy the radiation 

doses resulting from internal exposures during this period (HHS 2006a). 

5.1.5

Technical Basis for Internal Dosimetry at NTS (REECo 1993a) 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 Bechtel Nevada, Inc. (BN 2000, 2003a) cover the needs of 

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

Radionuclides of Concern

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) (REECo 1993a). 

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 

(HHS 2006b and HHS 2006a).
Document No. ORAUT-TKBS-0008-5 Revision No. 02 Effective Date: 04/15/2010 Page 15 of 113

5.1.6

As noted in the Historical Perspective (REECo date unknown a), 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 whole-body 

counts (REECo 1993a). 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. 

Bioassay Program Description

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 in 1967–1973, and by REECo at NTS in 1973. Prior to 1967, several 

analyses were done by Helgeson Nuclear Services, Inc.

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 1980–1989). 

By 1993, the common in vitro radiobioassay conducted at NTS included the following (REECo 1993a):

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

counting (70 min).

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

determined by beta counting (100 min).

• Gross fission products (GFPs) in urine – a 250-mL sample was determined by gross beta counting 

the oxide from a basic oxalate (100 min).

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

fluorometrically.
Document No. ORAUT-TKBS-0008-5 Revision No. 02 Effective Date: 04/15/2010 Page 16 of 113

• 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 min).

• 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.

REECo (1993a) 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 gross 

fission products, Pu, and tritium. Annual WBC.

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

(RCTs)/Rad-Safe Monitors] participated in the routine bioassay program. ORAUT (2007b) contains 

an expanded list of job titles with the potential for radiological exposure based on work assignment. 

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 

• Gross fission product (beta) 

• Specific radionuclides

Whole-body counts, lung counts, thyroid counts, and biological sampling were performed as soon as 

practicable after a suspected intake (REECo 1993a). Specific examples include:
Document No. ORAUT-TKBS-0008-5 Revision No. 02 Effective Date: 04/15/2010 Page 17 of 113

• Lung counts following a suspected intake of thorium, uranium, or a transuranic

• Whole-body counts 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

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

programs (REECo 1993a). Bioassay samples might have been taken to document that the worker 

was not internally contaminated as well as to confirm a suspected intake (REECo date unknown b).

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 (BN 

2003a). 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 BN (2003a), the DL is calculated as:

( )

KTs

3 Tb 1 65 Rb Ts 1 Ts

DL

× × + +

=

.

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 BN (2003a), 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. BN (2003a) 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 
Document No. ORAUT-TKBS-0008-5 Revision No. 02 Effective Date: 04/15/2010 Page 18 of 113

computed using the same value of alpha as used for the DL [Decision Level]. 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:

( )

KTs

Tb 3 3 29 Rb Ts 1 Ts

MDA





 



 + × × +

=

.

In REECo (1993a), the MDA equation was reported as:

MDA = [3 + (4.65(Cb)

1/2)] / ETVR

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 REECo (1993a) 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

MDA Values for Urinalysis and Fecal Analysis

In Vitro Urine and Fecal Analysis

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

2003a). 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

REECo (1993a) states that periodic urine samples were collected and analyzed for 238Pu, 239Pu, 

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

H, and gross fission products (GFPs). In vitro

radiobioassays included:

In Vitro Methods for Individual Radionuclides

• 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

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 
Document No. ORAUT-TKBS-0008-5 Revision No. 02 Effective Date: 04/15/2010 Page 19 of 113

Table 5-2. 2003 target minimum detectable activities for in vitro bioassay sample analysis.

Parameter and 

analysis method Sample typeb and container

Minimum detectable 

amounta 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

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: BN (2003a)

a. 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.

b. Sample collection was 24 hr.

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-hr void sample, obtaining reasonable recoveries, 

extending count times, and reasonable background levels.

d. Used in calibration activities 

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 (REECo 1993b). In 

1993, the 131I LLD was listed as 100 pCi/L urine (REECo 1993a). 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 LASL (1954) 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

gram 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 (LASL 1954).

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 (LASL 1958a, b). The LASL MDA value for 

americium was 9 × 10-1 pCi/24-hr sample (ORAUT 2004a). 
Document No. ORAUT-TKBS-0008-5 Revision No. 02 Effective Date: 04/15/2010 Page 20 of 113

REECo documentation (REECo 1969–1991) 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

μCi/mL for 241Am in urine (REECo 1977–1987). 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. 

REECo (1993a) 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 (REECo 1993a). The 

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

urine sample and a lung count. REECo (1993a) 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 gram 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 (LASL 1954). 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 thenoyltri 

fluoroacetone. 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 were 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 grams of plutonium). Samples showing 0.2 dpm/sample or more (1.5 × 

10-12 grams of plutonium) were rescheduled (LASL 1958a, b). The LASL MDA values for plutonium 
Document No. ORAUT-TKBS-0008-5 Revision No. 02 Effective Date: 04/15/2010 Page 21 of 113

(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 2004a).

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 (REECo 1993b). 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 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-hr 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 1968–1991b) and the 

permissible bone burden for 239Pu was listed as 0.04 μCi (REECo 1993b). From 1982 to 1987, the 

239Pu MDA was 5 × 10-11 μCi/mL and the 238Pu MDA was 2 × 10-10 μCi/mL (REECo 1977–1987).

REECo (1993a) 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 238Pu 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 prior to 1963. ORAUT (2008a) 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 0.17 and the ratio of 238Pu to 239Pu is 0.18. 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. Estimates of 241Pu are not necessary due to the

assumption that the intake estimates for 241Am account for any organ dose contribution from 241Pu. 

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 
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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 (LASL 1954). The LASL MDA value for tritium was 1 μCi/L (ORAUT 

2004a). In 1958, the procedure was the same except the tolerance for tritium in urine used was 85 

μCi/L (LASL 1958a). 

LASL (1958b) stated the following regarding tritium urinalysis: An acute dose resulting in 124 μCi/L in 

the body fluids (including urine) will expose the person receiving it to 0.6 rem in the first 2 weeks 

following the exposure and, until completely eliminated (assuming a 12-day elimination half-life), 

expose the person to 1.08 rem to the whole body. The 1958 tolerance for tritium in urine was listed as 

85 μCi/L. In 1959, the Maximum Permissible Concentration (MPC) for tritium was listed as 1.2 mCi in 

the body, which yielded 0.1 rem per week. Assuming 43.4 L of water in the body, 1-mCi body burden 

of tritium would result in a urine assay of 23 μCi/L. 

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 (REECo 1993b). 

In 1971, the urine sample tritium “alert level” was listed as 1 × 10-3 μCi/cc (REECo 1993b). The 

detection limit for 3

H in urine was listed as 1 × 10-6 μCi/mL from 1982 to 1987 (REECo 1977–1987).

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

2004). REECo (1993a) 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 2008b), 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 (BN 2003b). 

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 
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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 Waste Examination 

Facility (WEF). Uranium at the WEF is likely to be the 233U radionuclide, which can be present in 

some of the transuranic 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 the laser 

isotopic separations (AVLIS) method (Rich et. al. 1988, as cited in BN 2003a). 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 gram per 

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

LASL was 100 μg/L (LASL 1954). In 1958, the same method was sensitive to concentrations of 

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

1958a).

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 (LASL 1958a).

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 

(LASL 1958a).

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).
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A method for total uranium by fluorometry and alpha spectrometry and for 235U (anion exchange, 

purification, electrodeposition, alpha spectrometry) was listed in REECo documents published from 

1981 to 1983 (REECo 1968–1991a).

As described in REECo (1993a), 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-hr 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. 

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) (REECo 1993b). 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. 

REECo (1993a) 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 LASL (1954) 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 stainlesssteel 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 d/m of 230Th could be determined with this method. In 

1993, the 230Th fecal MDA was 0.01 pCi/g (REECo 1993a). 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 
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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 x 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 (LASL 1958a). REECO (1993a) 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 

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

distributed in the whole body.

The LASL (1958a) method for gross beta activity in urine was as follows. The nuclides of 90Sr90Y, 

140Ba-La, 144Ce-Pr, 89Sr, and gross fission products 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. LASL (1958a) states that if the 

gross beta (90Sr90Y, 140Ba-La) result is higher than 200 dpm/L, exposure should be suspected and 

investigated. The LASL MDA value for fission products was 50-100 dpm/L (ORAUT 2004a).

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 
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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 1970–1991). 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 1977–1987). 

REECo (1993a) 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.

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 

Assessment (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 log books, 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 whole-body count during any year and had a detectable fission product 

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

whole-body count. Whole-body counting began at offsite locations in 1964 (Pan Am 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 1977–1987).

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 
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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 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 

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 1968–1991).

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

such as whole-body (or specific organ) counting were preferred (REECo 1980–1989). 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 (REECo 1993a). 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 (REECo 1993a). 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

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 

Sample Collection and Correcting for Urinalysis Volume
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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.

REECo (1962a) 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-hr sample of at least 250 mL was required. To collect the 24-hr sample, a kit of three 

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

REECo (1970–1990) discussed how urine samples were received and processed. Specific 

instructions and minimum samples sizes were included for the following radionuclides (REECo 1969–

1991, 1993a; BN 1998).

• GFP: 250 mL (1969–1993)

• Plutonium-239 (urine) 

• 1 L (February 1969–June 1985) 

• 900 mL (July 1985–July 1987) 

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

• 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 REECo (1993a), 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-hr excretion. If a 24-hr 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-hr collection. In such a 

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

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

proper 12-hr 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.
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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).

REECo (1993a) 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 hr after intake (i.e., the “instantaneous” urinary 

excretion function was used) (REECo 1993a).

5.2.4

NTS performed fecal sample analyses as a special bioassay, usually if there were other indications 

that internal contamination had occurred. REECo (1993a) 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-hr 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. 

Fecal Sample Collection and Analysis

REECo (1993a) 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 following 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 following an incident (REECo 1993a). 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 hr after the incident (REECo 1993a). 
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5.3 IN VIVO MDAS, COUNTING METHODS, AND REPORTING PRACTICES

The 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 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 prior to 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 American Airways at the Nuclear Rocket 

Development Station as part of the Nuclear Space Propulsion Program (Teasdale 1985) for 

performing WBCs. This facility 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.3.1 and Table A9 for whole-body counting MDAs. 

On the basis of this information, it is assumed that chest counting at the new facility used Phoswich 

detectors. The first reference to chest counting was found in the 1983 REECo Standard Procedures, 

Chemical and Radiological Analysis (REECo 1969–1991). In vivo count measurements over 

minimum reporting limits are included in the annual report and were microfilmed for the historical files. 

Individual hard copies of whole-body counts were included in the files provided to DOE. 

As identified in REECo (1993a), the background subtraction algorithm used for these counts is 

designed to be invariant with respect to counting rates and the slope of the continuum. Identified 

photon peaks are corrected for natural background photon sources based on 24-hour environmental 

background counts. The background correction is not the traditional channel-by-channel subtraction, 

but rather the subtraction of normalized 30-minute background radioactivity (nanocurie amounts) from 

the radioactivity (nanocurie amounts) in the individual’s count.

The MDA is presently calculated as a function of the gross spectral counts and the full-width-halfmaximum (FWHM) energy and is consistent with American National Standards Institute (ANSI) 

standard N13.30 (ANSI 1996). The MDA (assuming a 5% probability of either a type 1 or type 2 error) 

is defined as:
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( )

T E Y k

C C Y

MDA

1

0.5

i

C W

1 2 C W

× × ×



  + × Σ

=

+

−

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

5.3.1

Helgeson Nuclear Services performed the first whole body counting at NTS. It used a shadow shield 

that was transported in a semitrailer (Helgeson 1967). The shadow shield employed an 8” x 4” 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). Helgeson estimated the sensitivity at two confidence levels: 50% 

and 99%. Table 5-3 summarizes these activities.

Whole-Body Counting

Table 5-3. Minimum detectable activities for the Helgeson 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

757

0.81

1.00

0.7 2.1

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

1,332

1.00

1.00

0.4 1.3

K-40* 71 1,461 0.107 7.1

(8.3 gK)

21.4

(24.9 gK)

Source: Helgeson (1967)

* 1 gK = 0.000118 g K-40 = 8.58138E-10 Ci = 0.858138 nCi/gK

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 ≥ 100 

keV, 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 whole-body counts performed prior to 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- × 3-in. 

NaI(Tl) crystal on the first visit and an 8- × 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.
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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: Pan Am (1967)

In January 1967, Pan American acquired and installed the shadow shield WBC. The detection 

system consisted of a 3-in. × 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 (Pan Am 1967). 

Table 5-5 lists results of the initial whole-body counts 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: Pan Am (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.

REECo standard operating procedures (REECo 1968–1991a) for chemical and radiological analysis 

describe the preliminary calibration of whole-body counts. The phantom consisted of polyethylene 

blocks stacked in a 3 × 5 × 5 arrangement with the 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 1969–1991). 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
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Radionuclide

Energy 

(keV)

Branching ratio 

gamma/disintegration

Y-88 898 0.91

Y-88 1,836 1.00

Source: REECo (1968–1991a)

The efficiencies for counting gamma rays from this phantom are described by the equation:

Efficiency(counts/disintegration) = 0.0089e -0.000175 * E

where E is energy in keV (REECo 1969–1991).

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 counts/sec.

Table 5-7. Whole-body count 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 (1968–1991a).

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 grams potassium, standard man equivalent, were reviewed by a senior health physicist to 

determine the need for further action.

Starting in 1978, REECo Radiological Safety Division procedures described whole-body counting 

(REECo 1969–1991). Documentation indicates that from 1978 through 1980, whole-body counting 

used a 20.3- by 10.1-cm-thick NaI crystal housed in a shallow shield. (The detection energy range 

was 100 to 2,000 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 (see Table 5-8), a

recount was performed after street clothes were exchanged for a paper suit. In addition, the 

background of the bed was rechecked. 

Table 5-8. Examples of whole-body count ‘alert’ levels 

(REECo 1969–1991).

Time (s)

Weight (lb)

155 175 195 215 235

Permissible counts

1,340 55,610 56,400 57,200 57,985 58,770

1,370 56,855 57,663 58,481 59,283 60,086

1,400 58,100 58,925 59,761 50,581 61,401

1,440 59,760 60,610 61,469 62,312 63,156

By 1981, the use of a paper suit became standard 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. 
Document No. ORAUT-TKBS-0008-5 Revision No. 02 Effective Date: 04/15/2010 Page 34 of 113

A 29.2-cm diameter by 10.1-cm thick NaI(Tl) detector was connected to a Canberra Series 30 

multichannel analyzer. Radon daughter products in the air were reduced by a high-energy particulate 

air (HEPA) filtration system. 

By 1983, two types of detection equipment were identified: (1) 29.2 cm by 10 cm thick NaI(Tl) crystal, 

and (2) Phoswich counting systems 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 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-ayear 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 containing 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.

The system used in 1993 was calibrated for whole-body counting and used for whole-body, thyroid,

and wound counting. 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 yield an arc with the detector of 50 cm, providing a full view of the 

body trunk in relation to the detector. Whole-body counts were performed (Teasdale 1985):

• 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 whole-body count and bioassay 

program

• For terminating employees

• For employees who had a suspected intake of radioactive material, particularly gamma-emitting 

fission and activation products.

The routine time for the whole-body count was 20 minutes. 

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. NCRP (1987b) provides mean body burdens of 137Cs for the United 

States for the years most likely to produce interference with occupational WBC results. Table 5-9 lists 
Document No. ORAUT-TKBS-0008-5 Revision No. 02 Effective Date: 04/15/2010 Page 35 of 113

those burdens. 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 in Table 5-9. 

Table 5-9. Mean body burdens of 137Cs from fallout in the 

United States.

Year

NCRP determined 

body burden (nCi)

a Year

Calculated body 

burden (nCi)b

1953 0.27 1978 1.07

1954 1.1 1979 1.05

1955 2.2 1980 1.03

1956 4.3 1981 1.00

1957 5.1 1982 0.98

1958 6.5 1983 0.96

1959 8.1 1984 0.94

1960 6.8 1985 0.91

1961 4.6 1986 0.89

1962 6.0 1987 0.87

1963 11 1988 0.85

1964 19 1989 0.83

1965 16 1990 0.81

1966 9.7 1991 0.80

1967 5.6 1992 0.78

1968 3.5 1993 0.76

1969 2.7 1994 0.74

1970 2.7 1995 0.73

1971 2.7 1996 0.71

1972 2.7 1997 0.69

1973 2.7 1998 0.67

1974 1.6 1999 0.66

1975 1.1 2000 0.65

1976 1.6 2001 0.63

1977 1.1 2002 0.62

2003 0.60

2004 0.59

a. From NCRP (1987b). NCRP only published Cs-137 data through 

1977. Extrapolated the date range from 1977 to present.

b. After 1978, the values have been calculated using the value for 

1977 as the initial activity and applying the activity equation A = 

A0e-(0.693/T1/2)t. Comparing any cesium bioassay results with these 

values is favorable to claimants because it does not take into 

account the Chinese atmospheric test or the accident at Chernobyl 

(i.e., any exposure for these events could be in the results for a 

worker, but will not be attributed to these events; it will be attributed 

to occupational exposure if the concentration is over the value 

found in Table 5-8 for the year the whole-body count was taken).

5.3.2

It has not been determined if lung counting was a viable bioassay method prior to construction of the 

WBC facility from pre-World War II steel. However, safety reports from the atmospheric testing era 

mention chest and lung counting.

Chest Counting

The first reference to lung or chest count was found in a REECo procedure dated July 1983 (REECo 

1977–1987). This was performed with two 5 in. diameter Phoswich detectors, which are used 
Document No. ORAUT-TKBS-0008-5 Revision No. 02 Effective Date: 04/15/2010 Page 36 of 113

primarily to detect low-energy X-ray and photon emissions from “heavy” elements such as 239Pu, 

241Am, etc. The total estimated sensitive area of the detectors was 500 cm2

.

The phoswich detectors were designed to detect low-energy photons (less than 100 keV) from 

radionuclides that had no other more readily detectable photon emission. They were designed to 

suppress detector background at the very low energies such as the uranium L X-rays from decay of 

the various plutonium radionuclides, particularly 238/239Pu. In addition, they have been used for 

detection of 241Am through the 59.5-keV gamma. Phoswich detectors can be used for other 

radionuclides, but their application is limited by poor resolution that makes identification difficult. This

is particularly true when compared with solid-state detectors such as GeLi.

One complicating factor in the measurement of low-energy photon emissions from the lung is the 

absorption of photons in the tissue overlying the lung – adipose (fat), muscle, cartilage, and bone. 

The thickness of these tissues and, as a consequence, the attenuation can vary significantly from one 

individual to the next. This is particularly serious for detection of the 17-keV plutonium X-rays. At this 

energy, 6 mm of muscle can attenuate half of the transmitted X-rays. In recent years, sophisticated 

ultrasound measurement techniques have been applied for accurate determination of the effective 

chest wall thickness (CWT). In the early days of lung counting, height/weight relationships had been 

used to estimate the CWT, but these were approximations, and could easily have led to errors of a 

factor of 2 or more. 

The method used at NTS in the early 1990s was based on published biometric relations for CWT 

measurement relations to weight and height. The chosen relation was (REECo 1993a):

CWT = 1.973 (W/H) – 2.0038

where:

CWT = chest wall thickness (cm)

W = weight (lb)

H = height (in.)

The situation is made more difficult by the significance in attenuation properties of the three primary 

tissues of concern: adipose, muscle, and bone. Although bone is obviously denser than adipose or 

muscle, the attenuation differences between those tissues can be significant. It is, therefore, not 

sufficient for DRs to determine the CWT; they must also estimate the relative fraction of each tissue. 

Another option is to use the WBC estimate; depending on body weight (120- to 185-lb range), the 

count rate could change by a factor of 1.57 (Pan Am 1967). 

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 or 100% muscle, 50% muscle – 50% adipose, and 13% 

muscle – 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.

In lieu of specific information on the efficiency and background characteristics of the phoswich system 

used at NTS, the International Calibration of Detectors Systems for the Measurement of Low Energy 

Photon Emitters In Vivo (IAEA 1991) contains such performance information for systems in use at 

other facilities around the world. This program focused on the use of phoswich detectors and the 

Livermore phantom. The IAEA study primarily addressed counting system detection capabilities for 

239Pu and 241Am. 
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In this intercalibration, 239Pu detection was based on uranium L X-rays, with a total intensity, Ix, of 

0.0457 X-ray per disintegration; Table 5-10 lists the contribution of the U L X-rays from 239Pu. The 

results of the intercalibration measurements can be used to estimate the expected range of MDA 

values. Table 5-11 lists background and efficiency data for measurements with 239Pu-loaded lungs. 

These are the results of one participating facility with higher-than-average backgrounds and lowerthan-average reported detection efficiencies. However, these values reflect those that a routine 

measurement program could have encountered.

Table 5-10. Uranium L X-ray intensity 

for decay of 239Pu.

X-ray energy–keV

X-ray intensity–

X-rays per 

disintegration

11.620 0.00106

13.442 0.00175

13.618 0.0157

15.400 0.00042

15.727 0.000296

16.410 0.0041

16.577 0.00046

17.068 0.00087

17.222 0.0157

17.454 0.00042

20.169 0.0038

20.487 0.00019

20.715 0.00018

20.844 0.0008

Total 0.045746

Source: LBNL (2004).

Table 5-11. Background and efficiency data for 239Pu detection using Phoswich detectors (12-25 keV) 

CWT–

cm

Normalized efficiency

counts/cm2/106 photons

Efficiencya

counts/photon

Efficiencyb

cpm/nCi

Estimated MDA–nCi

(40 minute count)

Torso 1.51 4.07 0.00204 0.207 21

13% muscle–

87% adipose

2.12 2.73 0.00136 0.138 32

2.79 1.68 0.00084 0.0852 51

3.34 1.24 0.00062 0.0629 69

4.03 0.84 0.00042 0.0426 102

50% adipose–

50% muscle

2.15 2.41 0.00120 0.1212 36

2.80 1.57 0.000785 0.0796 55

3.24 1.03 0.000515 0.0522 83

3.99 0.68 0.00034 0.0345 126

100% muscle 2.11 2.02 0.00101 0.102 42

2.75 1.18 0.00059 0.0599 73

3.32 0.79 0.000395 0.0401 109

3.96 0.45 0.000225 0.0228 191

Background 0.070 cpm/ cm2

Source: IAEA 1991 Working Material Document

a. For 500 cm2 detector

b. L-xray intensity = 0.0457 L-xrays/disintegration

Table 5-12 lists background and efficiency data for 241Am counting. The MDA values are markedly 

lower than those for 239Pu counting because:
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Table 5-12. Background and efficiency data for 241Am detection using phoswich detectors (40-80 

keV) 

CWT–

cm

Normalized efficiency

counts/cm2/106 photons

Efficiencya

counts/photon

Efficiencyb

cpm/nCi

Estimated MDA–nCi

(40 minute count)

Torso 1.51 158 0.079 63.0 0.11

13% muscle –

87% adipose

2.12 146 0.073 58.2 0.11

2.79 136 0.068 54.2 0.12

3.34 127 0.0635 50.6 0.13

4.03 116 0.058 46.2 0.14

50% adipose –

50% muscle

2.15 142 0.0710 56.6 0.12

2.80 131 0.0655 52.2 0.13

3.24 123 0.0615 49.0 0.14

3.99 112 0.056 44.6 0.15

100% muscle 2.11 144 0.072 57.4 0.12

2.75 132 0.066 52.6 0.13

3.32 123 0.0615 49.0 0.14

3.96 110 0.055 43.8 0.15

Background 0.163 cpm/ cm2

Source: IAEA 1991 Working Material Document

a. For 500 cm2 detector

b. 59.5 keV gamma intensity = 0.359

• The 59.5-keV gamma rays from 241Am are 7.8 times more abundant than the 239Pu uranium L Xrays.

• The 241Am gamma rays are far more penetrating than the L X-rays, so a much larger fraction of 

the photons emitted in the lung actually penetrate the overlying thoracic tissue to be detected.

REECo (1993a) identified lung counts with nonroutine bioassay for the suspected intake of thorium 

from the use of special nuclear materials, uranium, or any of the transuranics, especially californium 

and curium. Lung counts for low-energy photon emitters were performed with four Canberra Low 

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.

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 (BN 

2003a).
Document No. ORAUT-TKBS-0008-5 Revision No. 02 Effective Date: 04/15/2010 Page 39 of 113

5.3.3

During weapons testing at NTS, exposure to radioiodines was a major concern for the first 100 hr 

after detonation. Thyroid counts were a concern for workers involved in drill-backs or reentry/minebacks in this 100-hr timeframe. DRs can determine potential exposure within 100 hr 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 above-ground test or 

individuals on the ground following the fallout pattern. 

Thyroid Counting

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.”

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 1968–1991b). 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 (REECo 1993a). The following paragraphs contain information on 

the ratios of the iodines in the period immediately following 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 date unknown c), 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.

Following 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 5-13, 

derived from data cited in Kathren (1964), provides the relative activity normalized to that of 131I, at 

times ranging from 1 hr 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.

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 hr 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
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Table 5-13. Relative abundance of radioiodides.

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 -- -- -- --

radioactive decay of this short-lived species (T1/2 = 53 min) becomes predominant. The peak activity 

from this nuclide is 26 MCi per kt of explosive yield, compared to 0.11 MCi from 131I, which occurs 

about 5 hr 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 hr 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 hr after detonation, thereafter 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 

1.85E-12 μCi/cc was measured at Camp 12 on December 24, 1970, resulting in a calculated value of 

3.1E-12 μCi/cc 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 7.44E-6μCi is reasonable 

and results in a dose of less than 1 mrem to the thyroid.

5.3.4

Wound counting was performed (REECo 1993a) 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 wholebody 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. 

Wound Counting

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. BN (2003a) 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, 1958a); these reports 
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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), post-test drilling. In the 

early 1980s, brief activity and operations logs were included.

• 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 posttest drilling period.)

• A list of the radiation and toxic gas detection instruments in use.

REECo (1993a) 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 1989). 

The principal workplace air monitoring device used at NTS was the retrospective air sampler (RAS) 

(REECo 1993b). 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 (REECo 1993a). 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 μCi/mL. 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 μCi/mL 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 (REECo 1993a). 
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5.4.1

A 1959 memorandum titled “Operational Guides for Above-Ground Drill Sites into Ground Zero Areas” 

(REECo 1959a) stated that if surveys indicated significant beta-gamma concentrations, a 1-hr Staplex 

(8-in. × 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 (REECo 1959b):

Drilling Operations

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 

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 
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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 

(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: 90

Sr, 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 

(2007b) 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 (REEco 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 low-volume 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 
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measured 300 mR/hr or, in any case, at the end of 8 hr. In addition, a high-volume air sampler 

(similarly equipped and positioned) was used on each drill platform for 5 min 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 hr 

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-min 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 

μCi/m3 at mid-sampling time for each nuclide detected (other than radioxenons).

5.4.2

In 1961, REECo (1961) published specific air sampling guidelines for B and E Tunnel operations as 

follows:

Tunnel Operations

B Tunnel E 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 located near the 

work areas was checked for operation each shift and the 

filter was exchanged each Friday.

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 (radonthoron 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-hr 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-min air samples were collected rather than 5-min 

samples. In addition, after the 1-hr alpha check, the samples were transmitted to the Area 12 Change 

House counting facility (REECo 1962b).

Characterization of the tunnel environments occurred on a routine basis. Reports titled Current 

Tunnel Survey, Radiological Conditions and Reentry Requirements (REECo 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 prior to 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 

REECo (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.
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Fission product beta/gamma emitters are the radionuclides of greatest interest for NTS operations 

(REECo 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 

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 air flow calibrated liquid petroleum gas meters determined total air volume 

samples. Air sample data were reported in μCi/mL. 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 hr to attain 

an 80% saturation level. These data were also reported in μCi/mL (REECo 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

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.

Examples of Air Sampling Results

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 
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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 5-14 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.

Table 5-14. 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 5-15 lists additional air sampling data.

Table 5-15. 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 1958a) 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 (REECo 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.
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REECo (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 material and that

accompanying high levels of radon-thoron daughter products made measurements difficult.

According to REECo (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 5-16 lists examples of historical limits for air samples for years from 1950 to 1987 (REECo 

1985, date unknown b; 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 5-16. Historic air sampling limits (REECo 1993b) 

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 E12 μCi/mL (alpha)

When radionuclides were known, MPC was 3 

times above values based on exposures of 40-hr 

week

1958 Air tolerance for uranium was 50 μgm/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 & 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 < 1 × 10-8 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/cc for radon. Thoron levels 

were negligible.

1965 MPC for Pu-239 (in air) was 2 E-12 μCi/cc for 40-hr week
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Year Limit Note

1966–

1967

Alert level in air:

1 E-14 μCi/cc (alpha) 

1 E-11 μCi/cc (beta)

Alert levels based on MPC of unknown 

radionuclides in air over 168 hr were compared 

with sample activity results. 

1968 Kr-85 could be measured down to 2 E-07 μCi/cc

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

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. 

Contamination of Samples

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

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. 

Uncertainty

5.5.3

At NTS, the term less-than refers to data that are reported as less than some reporting level (REECo

1993a). For example, a plutonium urine bioassay might be reported as <1 E-12 (less than 1 × 10-12) 

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 

Less-Than Values
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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

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 2008a). 

Evaluation of Potential Internal Worker Exposure

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 2008a and ORAUT 2004b), and/or

• Air sampling limits (ORAUT 2005b).

Internal dose is assigned based on current revisions to project guidance [(NIOSH 2002), (ORAUT 

2004b), (ORAUT 2005b), (ORAUT 2005c), (ORAUT 2008b), (ORAUT 2007a) and (ORAUT 2007b)].

5.6.2

A class of employees from NTS has been added to the SEC. 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 1951-1962 inclusive 

for an EE. Based on the SEC petition evaluation, internal dose is not to be reconstructed for work 

prior to 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 prior to 1963 should be 

assumed to be valid, [HASL-300 (HASL 1972) 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. 

Workers with Pre-1963 Employment

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 
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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 worst 

case gamma emitter can be assumed.

Results from urine, chest counts, fecal, and whole body counts 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 and ORAUT 2007a). 

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 may also be considered 

depending on the cancer type (Allen 2007).

5.6.3

NIOSH believes that there is insufficient information to adequately support bounding internal dose 

(reconstructing internal dose with sufficient accuracy) for the portion of the SEC00084-NTS worker 

class who worked during the period of testing from 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 C.F.R. 835, supports NIOSH’s ability to 

bound internal dose for the evaluated class starting in 1993. Based on this information, NIOSH has 

identified the need to modify its position originally stated in the SEC00084-NTS Evaluation Report. 

NIOSH recommends including: 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 workdays aggregating at least 250 workdays, occurring 

either solely under this employment or in combination with workdays within the parameters 

established for one or more other classes of employees in the SEC. Although NIOSH recommends 

adding this portion of the 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.

Workers with Employment From 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 C.F.R. 835, 

supports NIOSH’s ability to bound internal dose for the evaluated class starting in 1993.

5.6.4

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 TLDs, 

the presence of external monitoring results does not equate to the worker entering 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 for jobs classified as 

having “little or no potential for internal” according to project guidance (ORAUT 2004b). 

Approach to Internal Dose Reconstruction for NTS 
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Environmental internal dose may be applicable as occupational internal dose for individuals with other 

job classifications which have very little chance of internal exposure (ORAUT 2004b) and no positive 

external dose. For individuals with job classifications where environmental internal dose assignment 

is not appropriate, the efficiency approach based on air sampling limits may be used for 

noncompensable cases (ORAUT 2005b) and is the preferred method for assigning internal dose for 

this site. The efficiency method described in ORAUT (2007e) can also be used for noncompensable 

cases consistent with the restrictions provided in the limitations of that bulletin.

Note: For respiratory cancers (e.g., Lung, ET1, and ET2) when underground work has been 

identified, radon exposures should be applied (whether monitored or unmonitored) according to the 

environmental section of this TBD (ORAUT 2008a). 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 (Rn-220) is also addressed. Thoron levels are provided for two 

locations in ORAUT (2008a). If the individual worked in other locations, thoron levels may be 

assigned using the thoron:radon ratio found in these tables (typically the thoron:radon ratio is 2:1, 

other values may be used as determined by the DR). OCAS (2008a) 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 

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 may need to put the exposure date no more than one week 

previous to the sample or select the date that best agrees with later negative data. There will often be 

no follow-up 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. 
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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 (Pu, U, Sr, tritium, etc.) but NOT gross gamma, gross alpha, or gross fission 

products (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, 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 2007a) 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 = 0.18 x 239Pu and 241Am = 0.17 x 239Pu (ORAUT 2008a). 

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 the duties of the individual and the timeframe when the individual was on site. The 

IMBA program includes defaults that can be used. For compensable cases, 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 tritium OTIB (ORAUT 2004c). 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.

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 gross fission products (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 Cs.
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5.6.4.3 For Individuals with Whole Body Count Results

For whole body count 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 whole body counts, use standard project methods 

(ORAUT 2007a).

5.6.4.4 For Individuals with Lung Count Results

Lung (or chest) counts were typically performed following a suspected intake of thorium, uranium, or a 

transuranic. 

• 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

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 2008a) 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 compensable or noncompensable. If the claim is completed as compensable, using 

environmental internals, only complete years of employment should be applied.

Examples of the Assignment of Internal Dose for Selected Job Titles

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-hr work week)]. DRs should base the 

assignment of internal dose to these workers on air sampling limits for noncompensable claims 

consistent with project guidance in ORAUT-OTIB-0018 (ORAUT2005b) and ORAUT-OTIB-0033 

(2005c). ORAUT (2007b) contains job titles and identifies the potential for tunnel work. For 

respiratory tract cancers, include radon (ORAUT 2008a) and thoron (OCAS 2006) for both 

compensable and noncompensable 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 compensable, best estimate or noncompensable based on project guidance (ORAUT 

2007b).
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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 an environmental dose plus 

some percentage of dose based on air sampling limits. Air sampling limits (ORAUT 2005b and 

2005c) can only be assigned for noncompensable cases. 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 and add internal dose based on air sampling limits for the 

years the worker was involved in underground work. For a favorable to claimant efficient evaluation of 

a noncompensable claim, air sampling limits should be used for years when the worker has positive 

recorded external dose and no bioassay results. 

An overestimate using the efficiency methods would be assigning environmental internal dose for 

support workers and air sampling internal dose (hypothetical internal dose) to a radiation worker. An 

overestimate could also be assigned based on project guidance (ORAUT 2007b). 

A more reasonable estimate would be prorating the hypothetical internal dose to reflect assumed 

duties of the worker and length of employment in a given category. For a worker with bioassay data, 

the results can be evaluated for dose and missed dose using IMBA and based on project guidance 

(ORAUT 2007b).

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

The following circumstances indicated in the DOE records will require special treatment or evaluation:

Unusual or Unique Working Environments

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 prior to 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).
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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 

“Plutonium Valley” or Area 11 or work at the Nellis Air Force Range (NAFR) 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 2008b).

5.6.6.3 Nuclear Reactor Development Station (NRDS)

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 re-entry 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 re-entry operation following a nuclear propulsion rocket test at the 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. 

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 2008a. 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.
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REFERENCES

AEC (U.S. Atomic Energy Commission), 1956, Report of the Test Manager on 56 Project - NTS, 

Albuquerque, New Mexico. [SRDB Ref ID: 2493]

AEC (U.S. Atomic Energy Commission), 1961, Standard Operating Procedure, NTS Organization, 

Radiological Safety, Chapter 0524, NTSO-0524-01, Washington, D.C. [SRDB Ref ID: 12468]

AEC (U.S. Atomic Energy Commission), 1962, “Employees, NTS, Mercury, Nevada – Personnel 

Exposures,” memorandum, Mercury, Nevada. [SRDB Ref ID: 8021]

Allen, David, 2007, “NTS Internal Dose,” e-mail to J. Griffin, March 20, Oak Ridge Associated 

Universities, Oak Ridge, Tennessee. [SRDB Ref ID: 34077]

ANSI (American National Standards Institute), 1996, Performance Criteria for Bioassay, ANSI N13.30, 

New York, New York.

Arent, L., and C. Smith, 2004, Personal Communications with Martha DeMarre 2003-2004, U.S. 

Department of Energy, Nevada Operations Office, Memorandum to File, NIOSH Dose 

Reconstruction Team, Cincinnati, Ohio. [SRDB Ref ID: 14054]

Black, S. C. and G. D. Potter, 1986, “Historical Perspectives on Selected Health and Safety Aspects 

of Nuclear Weapons Testing,” Health Physics, Vol. 51, pp. 17-33. [SRDB Ref ID: 37777]

Black, S. C., and Y. E. Townsend, 1996, Nevada Test Site Annual Site Environmental Report for 

Calendar Year 1995, DOE/NV/11718-037, U.S. Department of Energy, Nevada Operations 

Office, Las Vegas, Nevada. [SRDB Ref ID: 1791]

BN (Bechtel National, Inc.), 1998, Technical Basis for Internal Dosimetry at the Nevada Test Site,

DOE/NV/10630-64, Change Notice CN-DOE/NB-10630-64-01, Mercury, Nevada. [SRDB Ref 

ID: 39408]

BN (Bechtel National, Inc.), 2000, Technical Basis for Internal Dosimetry at the Nevada Test Site, DEAC08-96NV11718, Mercury, Nevada. [SRDB Ref ID: 1961]

BN (Bechtel National, Inc.), 2003a, Technical Basis for Internal Dosimetry at the Nevada Test Site,

TBD-E211-002, Mercury, Nevada. [SRDB Ref ID: 13987]

BN (Bechtel National, Inc.), 2003b, Uranium Background Levels in Non-occupationally Exposed 

Populations Residing Near the Nevada Test Site, Mercury, Nevada, prepared by MJW 

Corporation, Inc. [SRDB Ref ID: 13985] 

Brown, B., 1963, Re-entry Problems Associated with Radiation from Underground Nuclear 

Detonations, Reynolds Electrical & Engineering Company, Inc., RRS-63-10, Mercury, Nevada. 

[SRDB Ref ID: 8054]

DeMarre, M., 2003, “Information Request” (September 17), e-mail to L. Arent (ORAU Dose 

Reconstruction Team), U.S. Department of Energy, Nevada Operations Office, Las Vegas, 

Nevada. [SRDB Ref ID: 14053]

DeMarre, M., 2004a, Radiological Considerations, Presentation Slides, Bechtel Nevada Corporation, 

Las Vegas, Nevada. [SRDB Ref ID: 39409] 
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DeMarre, M., 2004b, “Measurements of Alpha, Beta, Gamma, and Gross Fission Products at NTS” 

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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 3, “Report of ICRP Committee 

II on Permissible Dose for Internal Radiation (1959),” June 1960.

• tritium – The alert level for a single tritium sample was 1.53 × 10 –2 μCi/cc. 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/cc.

• 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.
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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 detectability (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 (REECo 1993a).

minimum detectable activity (or amount) (MDA)

An a priori value used to evaluate the laboratory’s ability to detect an analyte in a sample (BN 

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. 
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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 following 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 (REECo 1993a). 

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 sec (Threshold Test 

Ban Treaty).

tolerance level

The term for maximum permissible exposure.
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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.
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TABLE OF CONTENTS

SECTION TITLE

A.1 Bioassay Codes and In Vitro Minimum Detectable Activities and Detection Levels ................

PAGE

69

A.1.1 Codes and Special Terminology Used in Bioassay Records....................................... 69

A.1.2 In Vitro Analyses for Individual Radionuclides ............................................................ 74

A.2 In Vivo MDAS and Reporting Practices at NTS......................................................................74

A.2.1 Whole-Body Counting................................................................................................. 74

A.2.2 Chest Counting........................................................................................................... 81

A.2.3 Thyroid Counts ........................................................................................................... 82

A.3 Area and Associated Radionuclides of Concern.....................................................................83

A.3.1 Radionuclides of Concern and Specific Bioassay Programs for NTS Facilities ........... 83

A.3.1.1 Drill-backs (Yucca Flat and Pahute Mesa, or Areas 1-10 and 18-20) ........ 84

A.3.1.2 Reentry and Mine Back (Area-1 & Area-12) .............................................. 86

A.3.1.3 Routine Tunnel Operations (Areas 1, 12, 15, and 16)................................ 88

A.3.1.4 Decontamination Facility (Area 6).............................................................. 88

A.3.1.5 Test Treatability Facility (Area 25)............................................................. 88

A.3.1.6 Atmospheric Weapon Safety Tests (Area 3 and Area 11) ......................... 89

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)............................ 89

A.3.1.8 Low-Level Waste Site (Area 3).................................................................. 89

A.3.1.9 Low-Level Waste Site (Area 5).................................................................. 90

A.3.1.10 Radiation Instrument Calibration Facilities (Area 6 and Area 23)............... 90

A.3.1.11 Radiography Operations (Area 23)............................................................ 90

A.3.1.12 Well Logging Operations (Areas 1–10 and Areas 18–20).......................... 91

A.3.1.13 Nuclear Explosive and Device Assembly Facilities (Area 6 and Area 

27)............................................................................................................. 91

A.3.1.14 Nuclear Rocket Development Areas (Area 25 and Area 26)...................... 91

A.3.1.15 Radioactive Source Storage Areas (Area 6 and Area 23).......................... 91

A.3.1.16 Radiochemistry and Counting Laboratories............................................... 92

A.3.1.17 Other Facilities Identified after 1992 (the end of testing)............................ 92

A.3.1.18 Summary lists of radionuclides of concern ................................................ 93

A.3.1.19 Atmospheric Testing Era ........................................................................... 94

A.3.2 Incidents.....................................................................................................................94

A.3.3 Respiratory Protection Practices at NTS................................................................... 103

A.3.4 Historical Practices and Contamination Levels ......................................................... 104

A.4 Reference Tables for Determining Internal Dose.................................................................. 106

References ...................................................................................................................................... 110

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.

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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. 

Records for REECo employees assigned to the 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). 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

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 Office of Compensation Analysis and Support 

(OCAS) 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.

Codes and Special Terminology Used in Bioassay 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 2004). 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), they 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, “turballoy” for natural uranium, and “D38” for depleted uranium. “LT” or “-” in the bioassay records means less than the detection limit. The 

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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 (gross fission products)

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

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

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For heading(a)

“bdy_prt_cd”

For heading(b)

“bdy_prt_pr”

For heading

“bdy_prt_df” Period

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. 

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 Transuranics 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

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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 

2004b). 

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 

2004b).

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

30 Lung

31 Whole body

60 Miscellaneous

Table A-5. Codes for units (DeMarre 2003).

Code Unit Description

10 dpm disintegrations per minute

11 dpm/cc d/m per cubic cm

12 dpm/g d/m per gram

13 dpm/L d/m per liter

15 dpm/m3 d/m per cubic meter

16 dpm/kg d/m per kilogram

20 Bq Becquerels 

21 Bq/cc Becquerels per cubic centimeter

22 Bq/g Becquerels per gram

23 Bq/L Becquerels per liter

25 Bq/m3 Becquerels per cubic meter

26 Bq/kg Becquerels per kilogram

30 pCi picocuries

31 pCi/cc picocuries per cubic centimeter

32 pCi/g picocuries per gram

33 pCi/L picocuries per liter

35 pCi/m3 picocuries per cubic meter

Code Unit Description

36 pCi/kg picocuries per kilogram

40 μCi microcuries

41 μCi/cc microcuries per cubic centimeter

42 μCi/g microcuries per gram

43 μCi/L microcuries per liter

44 μCi/mg microcuries per milligram

45 μCi/m3 microcuries per cubic meter

46 μCi/kg microcuries per kilogram

50 mCi microcuries

51 mCi/cc millicuries per cubic centimeter

52 mCi/g millicuries per gram

53 mCi/L millicuries per liter

54 mCi/mg millicuries per milligram

55 mCi/m3 millicuries per cubic meter

56 mCi/kg millicuries per kilogram

60 cpm counts per minute (c/m)

61 cpm/cc c/m per cubic centimeter

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Code Unit Description

62 cpm/g c/m per gram

63 cpm/L c/m per liter

64 cpm/mg c/m per milligram

65 cpm/m3 c/m per cubic meter

66 cpm/kg c/m per kilogram

67 % percent

Code Unit Description

68 μg/g micrograms per gram

69 μg/L micrograms per liter

70 μg micrograms U (Reg: KPA analysis)

71 μS/cm microsiemens per centimeter

72 mS/cm millisiemens per centimeter

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 2008b). 

Term Explanation

Area 23 (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

CP Control Point or A-6

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)

A-23, Building 111 Original RadSafe Building; this moved to Building 650 in A-23

Alcove Side excavation in drift

BP Bypass (type of tunnel drift)

CP Control Point or A-6

Drift Near-horizontal excavated passageway through earth; 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

Term Explanation

“U12b.10” Tunnel designation: U = underground

12 = Area 12 

b = B Tunnel

.10 = drift number

“U20bb” Downhole (i.e., vertical shaft) or crater designation: U = underground

20 = Area 20 

bb = hole or crater ID

Tunnels are located in NTS areas (REECo 1960 and ORAUT 2008b): 

• 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)

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A.1.2

Table A-7 lists limits of detection for urine and fecal analyses. 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. Lower Limit of Detection (LLD) is defined in REECo (1993a) 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-hr period. 

In Vitro Analyses for Individual Radionuclides

REECo (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 = 2 so

where so = the estimated standard error for the net sample activity.

The LLD expression for data changed after January 1, 1978, to:

LLD = 3.29 so

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 (REECo 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 on-station 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.

A.2 IN VIVO MDAS AND REPORTING PRACTICES AT NTS

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

The MDA values for 1967 are based on those reported by Helgeson (1967) for a 480-sec count. The 

values reported by Helgeson 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 

Whole-Body Counting

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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 1977–1987

H-3 1988c–1999d 300 - REECo 1993a

H-3 2000-2002 0.001 μCi/L - BN 2000

H-3 2003–present 0.005 μCi/L - BN 2003

Na-22 1970–presentd 200 - REECo 1993a

Na-24 1970–presentd 100e - REECo 1993a

Mn-54 1970–presentd 100 - REECo 1993a

Co-57 1970–presentd 100 - REECo 1993a

Fe-59 1970–presentd 600 - REECo 1993a

Co-60 1970–presentd 50 - REECo 1993a

Sr-85 1970–presentd 200 - REECo 1993a

Sr-90 1961–1969d 25 pCi/sample (limit of sensitivity)f - REECo 1993b

Sr-90 1970–1999d 0.8g - REECo 1993a

Sr-90 2000–present 1 - BN 2000, 2003

Zr-95 1970–presentd 600 - REECo 1993a

Nb-95 1970–presentd 100 - REECo 1993a

Tc-99m 1970–presentd 100 - REECo 1993a

Rh-101 1970–presentd 100 - REECo 1993a

Rh-102 1970–presentd 100 - REECo 1993a

Rh-102m 1970–presentd 200 - REECo 1993a

Ru-103 1970–presentd 200 - REECo 1993a

Ru-106 1970–presentd 1000 - REECo 1993a

Sb-122 1970–presentd 100 - REECo 1993a

Sb-124 1970–presentd 100 - REECo 1993a

Sb-125 1970–presentd 300 - REECo 1993a

I-131 1961–1969d 10 pCi/sample (limit of sensitivity) - REECo 1993b

I-131 1970–presentd 100 - REECo 1993a

I-132 1970–presentd 90 - REECo 1993a

Te-132 1970–presentd 100 - REECo 1993a

Ba-133 1970–presentd 200 - REECo 1993a

Ba-133m 1970–presentd 100 or 400h - REECo 1993a

I-133 1970–presentd 100 - REECo 1993a

Cs-134 1970–presentd 500 - REECo 1993a

I-135 1970–presentd 200 - REECo 1993a

Cs-136 1970–presentd 100 - REECo 1993a

Cs-137 1961–1969d 10 pCi/sample (limit of sensitivity) - Geiger and Whittaler 1961, REECo 1993b

Cs-137 1970–presentd 100 - REECo 1993a

Ce-139 1970–presentd 100 - REECo 1993a

Ba-140 1961–1969d 10 pCi/sample (limit of sensitivity)i - REECo 1993b

Ba-140 1970–presentd 500 - REECo 1993a

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Radionuclide Period Urine MDA (pCi/L)a Fecal MDA (pCi/g)a Source

La-140 1970–presentd 200 - REECo 1993a

Ce-141 1970–presentd 200 - REECo 1993a

Ce-143 1970–presentd 200 - REECo 1993a

Ce-144 1970–presentd 800 - REECo 1993a

Nd-147 1970–presentd 400 - REECo 1993a

Eu-152 1970–presentd 300 - REECo 1993a

Eu-154 1970–presentd 200 - REECo 1993a

Eu-155 1970–presentd 500 - REECo 1993a

Yb-169 1970–presentd 600 - REECo 1993a

Ta-182 1970–presentd 300 - REECo 1993a

W-187 1970–presentd 300 - REECo 1993a

Ir-192 1970–presentd 100 - REECo 1993a

Fission products- GFP (beta) 1977–1987 1E-10 μCi/mL (detection limit)b - REECo 1977–1987

Fission products- GFP 1988c–presentd 3j - REECo 1993a

Gross alpha 1968–present - 1E-07 μCi/g (detection limit)k REECo 1993b

Gamma 1977–1987 5E-8 μCi/mL (detection limit)b - REECo 1977–1987

Gamma (Cs-137) 1988c–present 100 - BN 2000, 2003

Ra-226 1958–1969d 0.88 dpm/sample (detection level)l - LASL 1958a

Ra-226 1970–presentd 300 0.4 or 0.04h REECo 1993a

Ra-226 2000–present 0.1 pCi/sample - ? BN 2000, 2003

Th-228 1970–presentd - 0.01 REECo 1993a

Th-230 1954–1992d 0.5 dpm/sample (detection level) m - LASL 1954, REECo 1993b

Th-230 1970–presentd - 0.01 REECo 1993a

Th-230 2003–present pCi/sample 0.04 pCi/sample BN 2003

Th-232 1993–1999d - 0.01 REECo 1993a

Th-232 2000–present 0.02 pCi/sample 0.05 pCi/sample BN 2000, 2003

U-234 1970–presentd 0.02 - REECo 1993a

U-234 2000–present 0.04 pCi/sample 0.04 pCi/sample BN 2000, 2003

U-235n 1961–1969 0.03 dpm/sample (limit of sensitivity) - REECo 1961, 1993b

U-235 1970–presentd 0.01 0.008 REECo 1993a

U-235 2000–present 0.04 pCi/sample 0.04 pCi/sample BN 2000, 2003

U-238 1970–1999d 0.02 0.008 REECo 1993a

U-238 2000–present 0.04 pCi/sample 0.04 pCi/sample BN 2000, BN 2003

Elemental U 1970–present 5 μg/Lo - REECo 1993a

Pu-238 1982–1987 2E-10 μCi/mL (detection limit)b - REECo 1977–1987

Pu-238 1988c–1999d 0.01 - REECo 1993a

Pu-238 2000–present 0.006 pCi/sample 0.03 pCi/sample BN 2000, 2003

Np-239 1970–present 400 - REECo 1993a

Pu-239p 1954–1957d 2 dpm/24 hr (detection limit)q - LASL 1954

Pu-239p 1958–1960d 0.05 dpm/sample (detection limit)r LASL 1958a, REECo 1993b

Pu-239p 1961–1976d 0.005 dpm/sample (limit of sensitivity) - REECo 1961, 1993b

Pu-239 1977–1987 5E-11 μCi/mL (detection limit)b - REECo 1977–1987

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Radionuclide Period Urine MDA (pCi/L)a Fecal MDA (pCi/g)a Source

Pu-239 1988c-2000d 0.01 (alpha spec)s; 50 (gamma spec)s 0.004 (alpha spec) REECo 1993a

Pu-239 2000–present 0.006 pCi/sample 0.03 pCi/sample BN 2000, 2003

Pu-240 1970–present 0.01 0.004 REECo 1993a

Pu-240 2000–present 0.006 - BN 2000, 2003

Am-241t 1954–1957d 2 dpm/sample (detection limit)u - LASL 1954

Am-241t 1958–1981d 0.5 d/m-24 hr sample (LLD)v - LASL 1958a, REECo 1961, 1993b

Am-241 1982–1987 2E-11 μCi/mL (detection limit)b - REECo 1977–1987

Am-241 1988c–1999d 0.03w 0.03 REECo 1993a

Am-241 2000-2002 0.008 pCi/sample 0.03 pCi/sample BN 2000

Am-241 2003–present 0.006 pCi/sample 0.03pCi/sample BN 2003

Cm-244 2000–present 0.008 pCi/sample - BN 2000, 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 x 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 REECo (1993a).

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. LASL 1954 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. LASL (1958a) states the detection limit at the 99% confidence level is approximately 0.05 dpm/sample.

s. Can’t 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. REECo (1993b) 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%.) LASL (1958b) 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..

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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)

ratio of kβ at the 95% confidence level (0.645) to the value of kβ at the 99% confidence level (2.326 –

0.71). Helgeson reported values for 60Co, 65Zn, 95Zr95Nb, and 137Cs; the values for the remaining 

radionuclides were derived from the Helgeson 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. × 3-in. NaI 

crystal, so values for 1967 to 1977 using a 3-in. × 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 decease 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. × 10-in. NaI detector. 

The efficiency reported for this detector is determined from E = 0.0089 × e-0.000175Eγ, where Eγ is in 

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 sec (21.7 min), 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.

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

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

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0

5

10

15

20

25

30

0 500 1000 1500 2000 2500

Photon energy - keV

Eff. - cpm/nCi

8x4

11.5x4

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. × 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. × 4-in. detector value by using the ratio of detector face areas: 

Bkg11.5x4 = Bkg8x4 × 2.07 = 279 cpm

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

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Radionuclide

Primary photon 

energiesb (keV)

Yγ

Photons per 

disintegration

Estimated MDAs (nCi)

1967c 1968–1977d 1978–1980e 1981–May 1993f

June

1993g

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

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

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Radionuclide

Primary photon 

energiesb (keV)

Yγ

Photons per 

disintegration

Estimated MDAs (nCi)

1967c 1968–1977d 1978–1980e 1981–May 1993f

June

1993g

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

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. Helgeson Nuclear Services shadow shield, with 8-in. × 4-in. NaI detector, 8-min count (Helgeson 1967).

d. Pan Am or REECo shadow shield with 3-in. × 3-in. NaI detector, 20-min count (Teasdale 1985).

e. REECo shadow shield with 8-in. × 4-in. NaI detector, 1,300-sec. count time.

f. REECo scanning bed with 11.5-in. × 4-in. NaI detector, 1,200-sec. count time (Teasdale 1985).

g. REECo germanium with chair geometry (REECo 1993a).

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 Np237, 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-min count and 0.010 μCi for a 40-min count. The MPBB of 137Cs in 

standard man is 30 μCi.

A.2.2

Based on REECo (1993a), lung counts occurred as soon as practicable after a suspected intake of 

thorium, uranium, or a transuranic. The first mention of lung counts in REECo procedures was in July 

Chest Counting

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1983 (REECo 1977–1987). There is no earlier mention of chest counting in REECo documentation; 

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-236 0.04

Ce-144 0.2 U-238 (f) 9(0.4 or 0.04)

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

Source: REECo (1993a)

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 Ra226). 

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 for Ra-226. When 

determined from Th-234, the MDA for U-238/Th-234 is 0.4 or 0.04 nCi assuming 

equilibrium [both values are listed in REECo (1993a); 0.04nCi is the value listed in Table 

4.1].

g. Based on Am-241 measurement.

A.2.3

Based on REECo (1993a), 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.

Thyroid Counts

REECo (1993a) 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 

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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: REECo (1993a)

promptly conducted for anyone suspected of being exposed to radioiodines. “Prompt” is interpreted to 

mean as soon as possible.

Table A-26 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 5-13. 

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.4 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

The information in the following paragraphs is from the 1993 NTS Technical Basis (REECo 1993a), 

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:

Radionuclides of Concern and Specific Bioassay Programs for NTS Facilities

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• 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: 

• Americium-241: 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

• Strontium-90: Urine

• Uranium-238: Lung count

• Plutonium-238: Urine, feces, lung count, wound count

• Plutonium-239: Urine, feces, lung count, nasal swabs, wound count

A.3.1.1 Drill-backs (Yucca Flat and Pahute Mes a, 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 hr after 

a test and the duration of the survey was typically about 1 hr 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 used to prevent the 

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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.

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 whole-body 

counts were conducted on a routine basis for specific radiation workers, and on a special basis 

following 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

Table A-12. Drill-back resuspension and mine back 

containment loss radionuclides for identification versus 

time after test. (REECo 1993a)

1 day 10 days 100 days 365 days 10,000 days

Na24 Na24a

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

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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 SCA comment 

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 prior to 1964. ORAUT (2007b) describes drill-back 

operations including potential exposures.

Table A-13. Drill-back resuspension, reentry/mine back 

resuspension, and decontamination facility, isotopes of 

concern for dose versus time after test. (REECo 1993a)

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

I-133 Te-132 Ce-144

I-135 Ce-141

Ce-143 Ce-144

Ce-144

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 hr of a test and the duration of the survey was typically about 1 hr 

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 

official end of the reentry phase. Teams with different missions, but with some overlapping duties, 

were involved in the underground reentry, including:

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• Work Teams (miners and a Rad-Safe Monitor/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, RadSafe Monitors/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 personnel protection 

equipment (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 Rad-Safe Monitors/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 line-of-sight (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 cement-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).

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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 following 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. (REECo 1993a 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 Tes t Treatability Facility (Area 25)

The Test Treatability Facility was a pilot project to bench-test technologies to be used for 

decontamination of soils containing transuranic 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. 

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A.3.1.6 Atmos pheric Weapon Safety Tes ts (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. 

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.3.1.7 Atmos pheric Weapons Tes t Areas (all Areas except Area 22, Area 23, and Area 27 

due to tes t location or airborne dis pers ion)

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

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A.3.1.8 Low-Level Was te 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 Was te 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, Pu, and Am.

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 Ins trument 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. 

A.3.1.11 Radiography Operations (Area 23)

The isotope of concern was 92Ir. In addition to 192Ir, BN (2000) lists 60Co and 137Cs as radionuclides of 

concern for this operation. Bioassay (urine/whole-body count) was conducted if a loss of source 

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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

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/whole-body count) was conducted if there was a loss of source 

containment detected by swipes or other analysis. Urine samples were gamma counted or processed 

for Pu depending on the source type. Additional “Nuclides of Concern” listed in BN (2003) include the 

following well logging sources: 241Am7

Be, 238Pu7Be, and 226Ra7Be. 

A.3.1.13 Nuclear Explos ive and Device As s embly 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 REECo (1993a) or BN 

(2000, 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 whole-body 

counts. 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 whole-body counts 

and/or urine sample collection and analysis were conducted. Specific bioassays and analysis 

depended on the source leaking.

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

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A.3.1.16 Radiochemis try 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 whole-body 

counts only for those personnel routinely handling uncontained radioactive materials. Urine samples 

were gamma counted, 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 tes ting)

BN (2003) included some facilities and radionuclides of concern that REECo (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 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.

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A.3.1.18 Summary lis ts 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 20Hg 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 Am-241

Y-90 I-131 Eu-155 U-233 Am-243

Zr-95 Ba-133 Yb-169 U-234 Cm-244

Nb-95

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

NTS facility and area Radionuclides of concern

Big Explosive Experimental Facility (BEEF) U-234 U-238

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 munition 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

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

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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

Waste Examination Facility (WEF) U-233 Pu-238 Pu-239 Pu-240

Am-241 Am-243 Cm-244 Cf-252

Site Monitoring Services and RAMATROL H-3 Sr/Y-90 Cs-137 U-234 

U-235 U-238 Pu-239 Am-241

Source: BN 2003a

a. BN (2000) also lists H-3 as a radionuclide of concern.

b. Identified by A. Ogurek (2004) as a radionuclide of concern

A.3.1.19 Atmos pheric Tes ting 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 (HHS 2006a and 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, 242Cu, 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:

A.3.2

Based on guidance from NTS, the DR should do the following when researching a claim: 

Incidents

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

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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

* indicates 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

1. Determine that the individual was working at NTS during a given operation. 

2. Consult Compilation of Local Fallout Data from Test Detonations 1945–1962 (DNA 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. (Refer to ORAUT 2008c and Section A.4.1). 

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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): 

1. Escape of radioactive gas from the Tower 2-A cab contaminated the working area, resulting in 

12 workers being exposed.

2. One worker removed his respirator while working in an area highly contaminated with alphaemitting material.

3. Several workers were exposed while removing and cutting a cable highly contaminated with 

alpha-emitting material.

4. Four workers without respirators entered a tunnel that was highly contaminated with alphaemitting materials.

• The Hardtack II report (DNA 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 respiratory, the taking of nasal swabs, and full anticontamination clothing 

(clothing that might have consisted of coveralls, hoods, booties, and gloves). 

• A memo for B Tunnel, U12b (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 (REECo 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 

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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 Weapons-related

1.85 kt

LLNL U12k.01 40K, 95Zr, 95Nb, 103Ru, 105Ru, 131I, 133I, 135I, 132Te, 140Ba/140La, 141Ce, 144CeTest/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 Weapons-related

4.5 kt

LLNL U9m 95Zr/95Nb, 103Ru, 106Ru, 105Rh, 131I, 133I, 135I, 132Te, 140Ba/140La, 141Ce, 144Ce

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.

Des Moines 6/13/62 Tunnel Weapons-related

2.9 kt

LLNL U12j.01 103Ru, 106Ru/106Rh, 131I, 133I, 135I, 132Te, 140Ba/140LaTest/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 Weapons-related

10 kt

LLNL U8d Gross fission 

products

Isotopic 

analysise

: 99Mo, 131I, 132I, 132Te

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: 133Xe, 135Xe

Controlled: 131I, 133I, 135I

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.

Diagonal Line 11/24/71 Shaft Weapons effects

<20 kt

LLNL

DoD

U11g 85mKr, 87Kr, 88Kr, 131I, 132I, 133I, 135I, 131mXe, 133Xe, 133mXe, 135XeTest 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 135Xe (80-85%), 85mKr, 87Kr, 88Kr, 131mXe, 133Xe, and 133mXe, with trace 

quantities of 131I, 132I, 133I, and 135I detected. Minor 

levels of radioactivity were detected offsite by aircraft 

only.

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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, Xe133mControlled 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% Xe133; 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.

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 Weapons-related

1.07 kt

LLNL U2eq Kr-85m, Kr-87, 

Kr-88, Xe-133, 

Xe-133, Xe135, 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.

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. Prior to 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 

(Simultaneous 

with Diamond 

Ace Test)

9/23/82 Tunnel Weapons effects

<20 kt

LLNL

DOD

U12n.15 Kr-85m, Kr-88, 

Xe-133, Xe133, 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. Prior to 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.

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Test name Date Type Purpose & yield Sponsor Location

Radionuclides

identified in 

release Release type & notes Release (Ci) Release summary

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. Prior to 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.

Midas Myth 2/15/84 Tunnel 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 Weapons-related

<20 kt

LLNL U2ev Kr-85m, Kr-87, 

Kr-88, Xe-133,

Xe-133, Xe135, 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.

Kappeli 7/25/84 Shaft Weapons-related

20-150 kt

LLNL U20am Kr-85 Late-time seep (release 

detected onsite only)

Natural, late-time 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 Kr-85

11/25/84 - 7/25/85: 3.6 Ci of Kr-85

7/25/85 - 7/25/86: 5.0 Ci of Kr-85

Tierra 12/15/84 Shaft Weapons-related

20-150 kt

LLNL U19ac Xe-133, Xe131m, Kr-85, 

Ar-37 

Late-time seep (release 

detected onsite only)

Natural, late-time 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, Xe133, Xe-135Controlled 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. Prior to 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.

Mill Yard 10/9/85 Tunnel Weapons

effects

<20 kt

LANL

DOD

U12n.20 Xe-133, Xe135, Xe-135mControlled 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.

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Test name Date Type Purpose & yield Sponsor Location

Radionuclides

identified in 

release Release type & notes Release (Ci) Release summary

Diamond Beech 10/9/85 Tunnel Weapons

effects

<20 kt

LLNL

DOD

U12n.19 Xe-133, Xe133m, Xe-135Controlled 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.

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 Weapons-related

20-150 kt

LLNL U19an Xe-133, Kr-85 Late-time seepage 

(onsite only)

Natural, late-time 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 Weapons-related

20-150 kt

LLNL U20ap Xe-133, Xe133m, Xe131m, 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 E-02

Xe-135: 9.8 E-04

Natural, late-time 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 Ghost 6/20/87 Tunnel Weapons

effects

<20 kt

LANL

DOD

U12t.09 Kr-85 Controlled tunnel purge

(onsite only)c

Controlled release at time 

of release: 3.0

Kr-85: 3.0

Activity was contained in cavity 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 (see Schoengold, 

DeMarre, and Kirkwood 1996) and are not necessarily the isotopes of concern for workers.

a. Containment failures are normalized to 12 hr after the test.

b. The camphor failure includes Ci-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 (DOE 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 Ρ+12).

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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 

Ρ+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 Ρ+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 Ρ+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 Ρ+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 Ρ+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 Ρ+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 Ρ+12).

Notes: The total release, at the time of release, was 36,000 Ci; at Ρ+12, the total calculated activity was 33,000 Ci. The total release associated with MIGHTY OAK was assumed to be all 133Xe, 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). 

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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 (REECo 1965a) for 1964 and 1965 describes 

exposure to internal radiation up to NTSO-0524-01 limits (AEC 1961). Two post-test 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. Historical Radiation Records for 

Operation Whetstone (July 1964 – June 1965), #20 Merlin Event, U3CT (REECo 1965b) 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 1965b).

• 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. Following 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: 

• 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 (AMAD) class Y aerosols with a committed 

effective dose of 90 mrem.

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• 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 sealed-tube 

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

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. 

Respiratory Protection Practices at NTS

The Operation Ranger Report (LASL 1951) 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 contamination area (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 (LRL 1957a) 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 memo lists measured contamination levels on PPE (up to 106 alpha cpm) and

includes individual nasal swipes results (sanitized). A second memorandum (LRL 1957b) details the 

August 30, 1957, reentry into U12c during which team members wore respirators and Mine Safety 

Appliances (MSA) Company air 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” (REECo 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 Willson 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.”

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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 d/m/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 d/m (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-hr 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 (REECo 1993b).

A.3.4

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.

Historical Practices and Contamination Levels 

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

Year Respiratory protection action levels

1957 No respirator < 9.6 x 103 dpm/m3 hrs (alpha) per quarter year

No respirator (short periods) < 100 dpm/m3 (alpha)

Respirator required > 100 dpm/m3 (alpha)

No respirator < 2.2 x 104 dpm/m3 (beta + gamma)

Respirator required > 2.2 x 104 dpm/m3 (beta + gamma)

Note: Respirators using filters have an efficiency up to 99.9% for particles as small as 0.3 micron 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 (Pu239 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.

4 x 10-11 μCi/cc (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

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The Defense Nuclear Agency (DNA) report on the 1957 Plumbbob Operations (DNA 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 (REECo 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 d/m (beta, total both nostrils) or 2 × 102 d/m (alpha, total both 

nostrils) were requested to submit urine samples. These samples were processed in the laboratory to 

determine body burdens of gross fission products or alpha emitters.

Procedure to Limit Radiological Exposures of Miners (DOE 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.

• 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 > 1 × 103 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)

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Year Contamination limits

• 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: REECo 1993b.

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. 

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. 

NOTE: The ICRP 30 (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 (10–100 days), and Y (greater than 100 days). (REECo 1993a 

defines D as days, M as weeks, and Y as years.)

The ICRP 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 

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Table A-29. Inhalation classes for radionuclides 

found at NTS. Radionuclide(s) REECo 1993a 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 W (a) W, Y W, Y

Pu-239 W (a) W, Y W, Y

Pu-240 W (a) - -

Pu-241 W (a) - -

Pu-242 W (a) - -

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

Ta-182 Y - -

Te-132 D - -

Th-228 W - -

Th-230 W - -

Th-232 W - W, Y

U-233 W - -

U-234 (b) W, Y (c) D, W or Y D, W or Y

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Radionuclide(s) REECo 1993a BN 1998 BN 2003 

U-235 (b) W, Y D, W or Y D, W or Y

U-238 (b) 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 (REECo 

1993a).

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, it this compound cannot be confirmed then 

Type D(current Type F) should be assumed. Based on NTS 

specific information both Type D, W, and Y are potentially present 

at NTS (REECo 1993a), therefore the solubility Type that 

maximizes dose should be used for noncompensable cases.

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.

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

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

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Radionuclide Half-life Beta energy (MeV)

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: RHTS 1954

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 2008a

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