ORAU Team

Dose Reconstruction Project for NIOSH

Technical Basis Document for the Nevada Test Site – 

Occupational Internal Dose

Document Number:

ORAUT-TKBS-0008-5 

Effective Date: 07/20/2006

Revision No.: 00 PC-1 

Controlled Copy No.: ________

Page 1 of 82

Subject Expert: Eugene M. Rollins

Document Owner

Approval: Signature on File Date: 09/23/2004

Eugene M. Rollins, TBD Team Leader

Approval: Signature on File Date: 09/22/2004 Judson L. Kenoyer, Task 3 Manager

Concurrence: Signature on File Date: 09/22/2004 Richard E. Toohey, Project Director

Approval: Signature on File Date: 09/30/2004

James W. Neton, Associate Director for Science

Supersedes:

None

TABLE OF CONTENTS

Section

Record of Issue/Revisions ...................................................................................................................

Page

5

Acronyms and Abbreviations ...............................................................................................................6

5.1 INTRODUCTION .....................................................................................................................8

5.2 In Vitro Minimum Detectable Activities and Counting Methods............................................... 12

5.2.1 In Vitro Urine and Fecal Analysis .................................................................................... 13

5.2.2 In Vitro Methods for Individual Radionuclides.................................................................. 13

5.2.2.1 In Vitro Bioassay for Iodine...................................................................................... 14

5.2.2.2 In Vitro Bioassay for Americium............................................................................... 15

5.2.2.3 In Vitro Bioassay for Plutonium................................................................................ 15

5.2.2.4 In Vitro Bioassay for Tritium..................................................................................... 16

5.2.2.5 In Vitro Bioassay for Uranium .................................................................................. 17

5.2.2.6 In Vitro Analysis for Strontium.................................................................................. 19

5.2.2.7 In Vitro Analysis for Thorium.................................................................................... 19

5.2.2.8 In Vitro Analysis for Radium..................................................................................... 20

5.2.2.9 In Vitro Gross Fission Product Analysis ................................................................... 20

5.2.2.10 In Vitro Analysis for Gamma Emitters ...................................................................... 22

5.2.3 Correcting for Urinalysis Volume..................................................................................... 22

5.2.4 Fecal Sample Analysis.................................................................................................... 22

5.3 In Vivo MDAS, Counting Methods, and Reporting Practices .................................................. 23

5.3.1 Whole-Body Counting..................................................................................................... 24

5.3.2 Chest Counting...............................................................................................................28

5.3.3 Thyroid Counting ............................................................................................................ 32

5.3.4 Wound Counting .............................................................................................................32
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5.4 Personal Air Sampling Data ................................................................................................... 32

5.5 Interferences and Uncertainty ................................................................................................ 34

5.5.1 Contamination of Samples.............................................................................................. 34

5.5.2 Uncertainty .....................................................................................................................35

5.5.3 Less-Than Values........................................................................................................... 36

5.5.4 Determination of Worker Exposure ................................................................................. 36

References ........................................................................................................................................38

Attachment 5D Occupational Internal Dose for Monitored Workers..................................................42

Acronyms and Abbreviations .............................................................................................................44

5D.1 Occupational Internal Dose.................................................................................................... 46

5D.2 Bioassay Codes and In Vitro Minimum Detectable Activities and Detection Levels ................ 46

5D.2.1 Codes Used in Bioassay Records................................................................................... 46

5D.2.2 In Vitro Analyses for Individual Radionuclides................................................................. 49

5D.3 In Vivo MDAS and Reporting Practices at NTS...................................................................... 53

5D.3.1 Whole-Body Counting..................................................................................................... 53

5D.3.2 Chest Counting...............................................................................................................54

5D.3.3 Thyroid Counts ...............................................................................................................54

5D.4 Other NTS Information ........................................................................................................... 55

5D.4.1 Radionuclides of Concern and Specific Bioassay Programs for NTS 

Facilities..........................................................................................................................55

5D.4.2 Incidents .........................................................................................................................63

5D.4.3 Respiratory Protection Practices at NTS......................................................................... 69

5D.4.4 Historical Practices and Contamination Levels................................................................ 70

5D.5 Reference Tables for Determining Internal Dose.................................................................... 72

References ........................................................................................................................................77

Glossary ............................................................................................................................................80
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LIST OF TABLES

Table

5-1 2003 target minimum detectable activities for in vitro bioassay sample 

analysis...............................................................................................................................

Page

13

5-2 Relative abundance of various radioiodides ........................................................................ 14

5-3 Minimum detectable activities for the Helgeson shadow shield ........................................... 24

5-4 Frequency (number of workers) of nuclide appearance and concentration in 

subjects counted before January 1967................................................................................ 25

5-5 Frequency (number of workers) of nuclide appearance and concentration in 

subjects counted from January to August 1967................................................................... 25

5-6 Preliminary calibration sources for 1977 through 1981........................................................ 26

5-7 Whole-body count sensitivities ............................................................................................ 26

5-8 Examples of whole-body count “alert” levels ....................................................................... 27

5-9 Mean body burdens of 137Cs from fallout in the United States ............................................. 29

5-10 Uranium L X-ray intensity for decay of 239Pu........................................................................ 30

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

(12-25 keV) .........................................................................................................................30

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

- (40-80 keV).......................................................................................................................31

5-13 Historic air sampling limits................................................................................................... 35

5D-1 Codes for analyte................................................................................................................ 46

5D-2 Codes for body parts........................................................................................................... 47

5D-3 Codes for radionuclides and sample types .......................................................................... 48

5D-4 Codes for sample types....................................................................................................... 49

5D-5 Codes for units....................................................................................................................49

5D-6 Limits of detection for urine and fecal analysis .................................................................... 50

5D-7 1993 whole-body counting MDAs........................................................................................ 53

5D-8 1993 MDAs for chest (lung) counting .................................................................................. 54

5D-9 MDAs for thyroid counts ...................................................................................................... 55

5D-10 Drill-back resuspension and mine back containment loss radionuclides for 

identification versus time after test ...................................................................................... 56

5D-11 Drill-back resuspension, reentry/mine back resuspension, and 

decontamination facility, isotopes of concern for dose versus time after test ....................... 57

5D-12 Decontamination facility, isotopes for identification versus time after test............................ 59

5D-13 Atmospheric weapons test areas, isotopes for identification and of concern for 

dose....................................................................................................................................59

5D-14 Low-level waste site (A-3), isotopes for identification and of concern for dose .................... 60

5D-15 Low level waste site (A-5), isotopes for identification and of concern for dose..................... 60

5D-16 Radiation instrument calibration facilities, isotopes for identification and of 

concern for dose ................................................................................................................. 60

5D-17 Radiation instrument calibration facilities, isotopes for identification and of 

concern for dose ................................................................................................................. 61

5D-18 Radiochemistry and counting laboratories, isotopes for identification and of 

concern for dose ................................................................................................................. 61

5D-19 Isotopes of concern for dose, summary list ......................................................................... 62

5D-20 Current nuclides of concern for NTS locations .................................................................... 62

5D-21 Releases from underground tests........................................................................................ 64

5D-22 Historical NTS respiratory protection action levels............................................................... 70

5D-23 NTS historical contamination limits...................................................................................... 72

5D-24 Solubility types for radionuclides found at NTS.................................................................... 73
Effective Date: 09/30/2004 Revision No. 00 Document No. ORAUT-TKBS-0008-5 Page 4 of 82

5D-25 Fission products up to 1 year old as identified in 1959 documentation ................................ 74

5D-26 Other common radionuclides not normally a part of fission products that might 

be present...........................................................................................................................74

5D-27 Radionuclide activity ratios at formation (immediately after detonation)............................... 75

5D-28 Specific activity of selected alpha emitters .......................................................................... 76
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RECORD OF ISSUE/REVISIONS

ISSUE 

AUTHORIZATION 

DATE

EFFECTIVE 

DATE REV. NO. DESCRIPTION

Draft 11/07/2003 00-A Technical Basis Document for the Nevada Test 

Site – Occupational Internal Dose. Initiated by 

Eugene M. Rollins.

Draft 07/13/2004 00-B Incorporates internal and NIOSH review 

comments. Initiated by Eugene M. Rollins.

09/30/2004 09/30/2004 00 First approved issue. Initiated by Eugene M. 

Rollins.

09/30/2004 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
Effective Date: 09/30/2004 Revision No. 00 Document No. ORAUT-TKBS-0008-5 Page 6 of 82

ACRONYMS AND ABBREVIATIONS

ANSI American National Standards Institute

BN Bechtel Nevada, Inc. 

BZ breathing zone

CAM continuous air monitor

cc cubic centimeter

CFR Code of Federal Regulations

cpm counts per minute

CWT chest wall thickness

DAC Derived Air Concentration

DL Decision Level

DOE U.S. Department of Energy

dpm disintegrations per minute

DU depleted uranium

EPA U.S. Environmental Protection Agency

FWHM Full-Width-Half-Maximum

g gram

GFP gross fission product

G-M Geiger-Mueller

HEPA high-energy particulate air (filter)

HEU highly enriched uranium

IMBA Integrated Modules for Bioassay Assessment

K potassium

keV kilo electron volts

Kr krypton

L liter

LANL Los Alamos National Laboratory

LASL Los Alamos Scientific Laboratory

LLD Lower Limit of Detection

LLI lower large intestine

LLNL Lawrence Livermore National Laboratory

LRL Lawrence Radiation Laboratory

m3 cubic meter

MDA Minimum Detectable Activity; Minimum Detectable Amount

mL milliliter

mm millimeter

MPBB Maximum Permissible Body Burden

MPC Maximum Permissible Concentration

NaI sodium iodide
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NAS National Academy of Sciences

NBS National Bureau of Standards

nCi nanocurie

NTS Nevada Test Site

pCi picocurie

RAS retrospective air sampler

REECo Reynolds Electrical & Engineering Company, Inc.

RSN Raytheon Services Nevada

TBD technical basis document

ULI upper large intestine

U.S.C. United States Code

WBC Whole-Body Counter

WEF Waste Examination Facility

μCi microcurie
Effective Date: 07/20/2006 Revision No. 00 PC-1 Document No. ORAUT-TKBS-0008-5 Page 8 of 82

5.1 INTRODUCTION

Technical basis documents (TBDs) and Site Profile Documents are not official determinations made 

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

documents that provide 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 the 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

As noted above, the statute 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 contains an exclusion with respect to the Naval Nuclear Propulsion Program, 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 dosimetry monitoring results are considered valid for use in 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:

] 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) define “performance of duty” for DOE employees 

with a covered cancer or restrict the “duty” to nuclear weapons work.

 

1 The U.S. Department of Labor is ultimately responsible under the EEOICPA for determining the POC. 
Effective Date: 07/20/2006 Revision No. 00 PC-1 Document No. ORAUT-TKBS-0008-5 Page 9 of 82

• Radiation from naturally occurring radon present in conventional structures

• Radiation from diagnostic X-rays received in the treatment of work-related injuries

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. The NTS was, and is, 

an outdoor testing and research facility rather than a manufacturing or processing site. When the 

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). 

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

1995, became responsible for onsite radiological safety activities in 1955 and began onsite laboratory 

analyses, including bioassay samples, in 1958. LANL and the military were the primary parties 

responsible for radiological safety oversight prior to 1955. 

LANL conducted bioassay procedures from 1955 to 1958. In 1958, REECo began onsite bioassay. 

By 1961, REECo was conducting routine bioassay for 3

H, 239Pu, gross fission products (GFP), and 

gamma emitters. Bechtel Nevada (BN), which became the site contractor in 1995, currently holds 

bioassay responsibility. In 2001, BN disbanded the onsite analytical laboratory and contracted the 

services to an outside laboratory.

In 1967, REECo established an onsite whole-body counting capability at NTS. 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 and size of the test. 

The information in this TBD applies only to the NTS. It does not apply to cases in which the 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 will be addressed separately.

Historic Monitoring Perspective

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

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

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

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

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

perspective for internal dose (REECo unknown a): 

“During 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.” 
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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. 

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). Potential intake from resuspension is discussed in Section 4.1.2 of this TBD, which addresses 

environmental doses. 

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 21, 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. 

Radionuclides of Concern

Technical Basis for Internal Dosimetry at the NTS (REECo 1993a) 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 2003). Subsequent technical basis documents by Bechtel Nevada, Inc. (BN 2000, 2003a) 

cover the current needs of NTS and are not comprehensive on the support of nuclear weapons 

testing. 

 

Each operation that occurred at NTS had a different set of radionuclides of concern at the time of the 

test and afterward. 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 5D, Section 5D.4.1.

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 2003).
Effective Date: 09/30/2004 Revision No. 00 Document No. ORAUT-TKBS-0008-5 Page 11 of 82

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. 

Bioassay Program Description

As noted in the Historical Perspective (REECo date unknown a), the dosimetry emphasis was on 

external monitoring. Early bioassays 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. Routine bioassays at NTS included quarterly 

urine samples, annual whole-body count, and new/termination whole-body counts (REECo 1993a). 

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:

• 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

Facility descriptions and their specific routine bioassay programs (REECo 1993a) are summarized in 

Attachment 5D, Section 5D.4.1. Bioassay samples might be taken to document the fact 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 performed through projects with a 

specified (often short) duration. Routine bioassay monitoring, therefore, is usually part of the project. 

A baseline sample is collected as necessary, and post-work samples are collected at the conclusion 

of the project. If a project requiring bioassay monitoring is expected to last longer than 6 months, 

bioassay sampling on a quarterly to semiannual frequency throughout the duration of the project is 

recommended. An exception to this could be a 3

H project with high exposure potential with a 

relatively high frequency of monitoring required (e.g., weekly to monthly), because 3

H is not detected 

by typical field monitoring instruments (BN 2003a).
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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 

non-detection (Type II error) while accepting a probability, alpha (Type I error), of erroneously 

deciding that a positive (non-zero) quantity of analyte is present in an appropriate blank sample. The 

MDA is computed using the same value of alpha as used for the DL [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 33 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 
Effective Date: 09/30/2004 Revision No. 00 Document No. ORAUT-TKBS-0008-5 Page 13 of 82

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

background.

Bioassay codes used in the records are in the tables in Attachment 5D, Section 5D.2.1. 

5.2.1

MDA Values for Urinalysis and Fecal Analysis

In Vitro Urine and Fecal Analysis

Table 5-1 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. Historic MDA values are included in Section 5.2.2 and Attachment 5D, 

Table 5D-6. 

Table 5-1. 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-230/232 (d) 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 

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. Common in vitro

radiobioassays included:

In Vitro Methods for Individual Radionuclides

• Tritium in urine by liquid scintillation beta counting

• Strontium in urine by beta counting

• Gross fission products 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
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• Gamma emitters in urine and feces by gamma spectroscopy

The following paragraphs discuss in vitro methods for specific radionuclides. The MDAs for these 

radionuclides are listed in Attachment 5D, Table 5D-6. The suggested priority for dose reconstructors 

with regard to MDAs and other detection thresholds is to use (1) the limit on dose report if available, 

(2) Table 5D-6 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). 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

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 isotope, iodine 

thyroid dose related to external gamma dose, an in vitro blood testing protocol, percent contribution to 

dose among isotopes, relative thyroid doses from inhalation of curie-per-cubic-meter concentrations of 

each isotope, and a comparison of external whole-body dose and internal dose to the thyroid. 

Following a nuclear detonation, several radioiodine isotopes are produced; those with mass numbers 

131 through 135 are of significance from the standpoint of exposure to personnel. Table 5-2, derived 

from data presented in Bolles and Ballou (1956) 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 any one iodine isotope are known, the activity of the other iodine isotopes can be 

determined. Dose reconstructors should consult access records in the “Other Monitoring” section of 

the DOE file to determine likely exposure time after the test.

Table 5-2. Relative abundance of various 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 -- -- -- --

The relative activity from each of these five radioiodine isotopes at early times after fission is a 

complex function of the ingrowth and decay of the various iodine isotopes and their fission chain 

precursors, and 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 the activity from the five radioiodine species. The peak 

activity from this radionuclide occurs at about three-fourths of an hour after detonation, after which 

ingrowth of activity is no longer dominant and radioactive decay of this short-lived species (T1/2 = 

53 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 isotopes in relation to 131I. At times less 

than about 1 hr after detonation, the relative activity from the higher iodines is very much larger than 

at later times, largely as a result of the rapid ingrowth of 131I, whose absolute activity peaks at about 

5 hr after detonation, thereafter declining approximately according to its characteristic 8.05-day half 
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life. Kathren (1964) contains a graphic representation of the relative activity from the five radioiodine 

isotopes of significance.

Dose reconstructors will need to consider iodine when the cancer identified with the claim is thyroid.

5.2.2.2 In Vitro Bioassay for Americium 

The LASL (1954) method for 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 low-background 

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 1958).

REECo documentation (REECo 1968-1991a) 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 states that 241Am is analyzed using 243Am as a tracer. From 1982 

to 1987, the detection limit was listed as 2 × 10-11 μCi/mL for 241Am in urine (REECo 1977-1987). 

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.” Attachment 5D, Table 5D-6 lists 

historical detection thresholds for various periods. Table 5-1 lists the current MDA. 

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). 

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 
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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). 

In 1961, REECo urine analysis 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 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 plutonium urine and fecal samples were analyzed by alpha spectrometry, 

which could not differentiate between 239Pu and 240Pu. Urine samples were also analyzed by gamma 

spectroscopy, which could differentiate between the two isotopes. 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 0.02 pCi/L. The 238Pu LLD was 0.01 pCi/L urine. Routine bioassay consisted of a 

quarterly urine analysis, and the special bioassay consisted of a combination of urine and fecal 

analysis plus a lung count. Attachment 5D, Table 5D-6 lists historical detection thresholds for various 

periods. The current MDA is listed in Table 5-1. 

5.2.2.4 In Vitro Bioassay for Tritium 

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

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

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

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

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

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

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

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

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

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

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

the fluid intake of the worker (LASL 1954). In 1958, the procedure was the same except the tolerance 

for tritium in urine used was 85 μCi/L (LASL 1958a). 
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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 to1987 (REECo 1977-1987).

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. The 

current MDA is listed in Table 5-1. Attachment 5D, Table 5D-6 lists historical detection thresholds for 

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

“EVERGREEN,” "MINT," or "T". 

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

2004). Tritium was reported separately in the dosimetry records. Dose reconstructors 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 of 0.355 rem. This dose is based on the MDA of 5 μCi/L 

used at NTS. Individuals who were involved in tunnel work with job classifications of miner, mucker 

(muck machine operator), “bull gang” (underground laborer), shifter, tunnel walker, dinky locomotive 

operator and who held a Q-level clearance should be assigned tritium dose. No worker whose 

employment history is intermittent (employment intervals < 5 months) could have obtained a Q 

clearance and, therefore, would not have been involved in tunnel reentry or emplacement of devices 

during events. Having a Q-level clearance and working in Area 1 or 12 is an indication to the dose 

reconstructor of the possibility of tritium exposure. All other workers should not be assigned tritium 

dose (Arent and Smith 2004). 

Because of the ubiquitous distribution of tritium in the body, all organ doses will be identical to the 

annual dose. For the Interactive RadioEpidemiological Program (IREP), assume that all internal 

doses are “chronic” exposure rates. The radiation type for tritium is “electron < 15 keV.” The dose 

distribution type is “constant.” The annual dose is put in the parameter 1 column. Parameters 2 and 

3 are not used in IREP calculations.

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). If the 238U:234U ratio
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does not indicate a likely DU or HEU intake and there are no field indicators that an intake occurred, 

the result might be judged unlikely to have come from an occupational intake and no dose is 

calculated (BN 2003a).

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 the NTS.

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

fluorescence (the principle 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 5D, Table 5D-6 lists historical 

detection thresholds for various periods. Table 5-1 lists current MDAs for uranium isotopes. 

BN (2003b) stated that approximately 60 nonoccupationally exposed adults residing in the 

southwestern portion of Nevada were sampled to determine the natural uranium background. The 

analyses showed that the distribution of uranium approximated lognormal but tended to be slightly 

skewed at high concentrations. Geometric means for 234U, 235U, and 238U were 0.025, 0.0067, and 

0.016dpm/L, respectively. The 99.9th percentile of the sample result distribution used to separate 

environmental from potential occupational exposures was observed to be 0.208 dpm/day for 234U, 

0.153 dpm/day for 238U, and at 0.446 dpm/day (212 ng/L or 0.2 pCi/sample) for total uranium. 

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 5D, Table 5D-6 lists historical detection thresholds 

for various periods. Table 5-1 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 stainless-
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steel plate and counted for alpha activity with a low-background proportional counter. Natural 

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

determination. Quantities of the order of 0.5 d/m of 230Th could be determined with this method. In 

1993, the 230Th fecal MDA was 0.01 pCi/g (REECo 1993a). Attachment 5D, Table 5D-6 lists historical 

detection thresholds for various periods. Table 5-1 lists the current MDA.

5.2.2.8 In Vitro Analysis for Radium

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

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

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

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

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

estimated that about 0.01% of the body radium [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. Table 5D-6 in Attachment 5D lists historical detection thresholds for various periods. 

Table 5-1 lists current MDA value.

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. 

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.) 

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.

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 request if the gross 

beta activity exceeded the following control limits: 
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• 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.

REECo (1961) stated that cesium was not recovered by this procedure (i.e., coprecipitating fission 

products on alkaline earth phosphates from an alkaline solution and counting the sample package for 

gross beta activity), 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 a

considerable saving of time and money, according to REECo records. 

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

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

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

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

sensitivity was listed as about 10 pCi per sample (REECo 1961). From 1982 to 1987, the detection 

limit for GFP (beta) in urine was 1 × 10-10 μCi/mL (REECo 1977-1987). 

REECo (1993a) states 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). The specific 

MDA for 137Cs in urine was listed as 100 pCi/L (REECo 1993a).

Current guidance (BN 2003a) includes the following information for nonoccupational exposure. 

Cesium-137 is a product of fallout from weapons testing. Detection of background levels depends on 

the individual’s diet, the primary source of 137Cs being game meat (elk, deer, etc.). As with 90Sr, 

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

reconstructors need not consider results less than the derived screening level for investigation of 

intake or assignment of dose. If a 137Cs result is above the derived screening level, but there are no 

field indicators that an intake might have occurred, it is advisable to question the individual regarding 

the possibility of recent consumption of wild game meat. This information can support a decision on 

whether an occupational intake has occurred. Cesium-137 is a radionuclide of concern at NTS 

because of the nature of the work performed. It is appropriate to address only 137Cs levels above the 

annual reported levels in NCRP (1987a) by evaluating the intake using the Integrated Modules for 

Bioassay Assessment (IMBA) computer program (see Table 5-9). Attachment 5D, Table 5D-25 

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

Dose reconstructors should select fresh or aged fission products based on what is most claimantfavorable for the cancer diagnosed. The bioassay records do not indicate fresh or aged. Dose 

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

They should review the “Other Monitoring” section of the case file for additional data, such as access 

logs and radiological control technician log books, and review test information to determine if they 

need to consider any fission products (Table 5D-25) unique to the test based on the cancer location. 

They might need to request additional records to determine aged or fresh fission products.
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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 as needed based on the results of the initial pulse height analysis. 

Reportedly, almost any gamma component 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).

All urine samples were counted by gamma spectroscopy in 1993 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. Table 5D-6 in Attachment 5D lists historical detection thresholds for various periods. Table 5-1 

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.

Correcting for Urinalysis Volume

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. 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. 

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).

5.2.4

Fecal sample analyses were performed at NTS 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. 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 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, a week for 

example, 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. 
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5.3 IN VIVO MDAS, COUNTING METHODS, AND REPORTING PRACTICES 

During atmospheric testing (1951 to 1962), in vivo methods of analysis were in place but were 

secondary to other measurements. NTS did, however, have a bioassay program that included wholebody, thyroid, lung, and wound 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). It is assumed 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 REECo employees have been retrieved from other locations for inclusion in the REECo 

(now BN) record archives. Other reports/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 Whole-Body Counter (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.

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. 

Since 1967, whole-body counting has been performed at NTS. This is when 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). This facility was transferred to NTS REECo 

dosimetry in 1974. Routine counting was performed for drillers, miners, and radiation monitoring 

personnel beginning in mid-1975. In 1977, NTS considered adding lung counting for low-energy Xray 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 D, Section 5D.3.1 for whole-body counting MDAs. 

On the basis of this information, it is assumed 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 1968-1991a). 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) 
Effective Date: 09/30/2004 Revision No. 00 Document No. ORAUT-TKBS-0008-5 Page 24 of 82

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

is defined as:

MDA = [C1 + C2 * (∑c+wYi)

0.5] 

 c-w 

 (T1 * E * Y * k)

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

Whole-body counting is not identified in the analysis capability at NTS from 1951 to 1973 but, as 

noted above, did occur prior to 1967. Helgeson Nuclear Services performed the first whole body 

counting at the 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

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. Other 

radionuclides that dose reconstructors might need to address were calculated based on this formula 

and are listed in Table 5D-7.

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)

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

A draft summary report provided by NTS indicated the completion of more than 300 whole-body 

counts 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. 
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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.

Table 5-4. Frequency (number of workers) of nuclide 

appearance and concentration in subjects counted before 

January 1967. 

Nuclide

Body burden

< 0.02 μCi < 0.05 μCi < 0.1 μCi < 0.5 μCi

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- × 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.

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 [Reserved]. Table 5-6 lists the sources used for this calibration.

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. 
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Table 5-6. Preliminary calibration sources for 1977 

through 1981. 

Radionuclide

Energy 

(keV)

Branching ratio 

gamma/disintegration

Cr-51 320 0.09

Mn-54 835 1.00

Zn-65 1,115 0.49

Zn-65 511 0.34

Y-88 898 0.91

Y-88 1,836 1.00

Source: REECo (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 1968-1991a).

Sensitivity was defined as activity that is detectable with a relative 2 sigma error of +20%. 

Background counts were 1,000 seconds. At a background count of 45 counts per second, the 

sensitivities for potassium and cesium are listed in Table 5-7. 

Table 5-7. Whole-body count sensitivities. 

Radionuclide

For 20-minute 

count (μCi)

For 40-minute 

count (μCi)

K-40(a) 0.104 0.091

Cs-137(b) 0.011 0.010

a. Source: REECo (1968-1991a). 

b. Potassium activity in standard man is about 0.12 μCi.

c. 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 printout of the whole-body count. 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. 

Documentation indicates that from 1978 through 1980, whole-body counting used a 20.3- by 10.1-cmthick 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 (12-160) 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. 

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. 
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Table 5-8. Examples of whole-body count “alert” levels. 

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

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- 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-a-year calibration used 

a Bomab water phantom constructed of polyethylene that weighed 70 kg when filled with water and 

National Bureau of Standards (NBS)-traceable solutions of four monoenergetic gamma ray emitters. 

Phoswich detector system calibration used a tissue- and bone-equivalent torso phantom designed by 

LLNL. Tissue-equivalent plastic dosed with 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:

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. 
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The suggested priority for dose reconstructors with regard to MDAs and other detection thresholds is 

to use (1) the limit on dose report if available, (2) Attachment 5D, Table 5D-7 values, or (3) an 

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

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 whole-body count results. 

Table 5-9 lists those burdens. If whole-body count results show detection of only 40K and 137Cs and 

the 137Cs result is less than the values listed in Table 5-9, the 137Cs results can be assumed to be due 

to fallout. There are two exceptions:

• If other fission or activation product radionuclides are present in either the whole-body count or 

a recent urinalysis for 90Sr or any of the iodines, it is claimant-favorable to assume the 137Cs is 

from occupational sources. 

• Depending on the level of 137Cs in the whole-body count and the worker job description, the 

exposure might be based on the job.

5.3.2

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 

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

. 

Chest Counting

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 crude and could easily have led to errors of a factor of 2 or 

more. 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 dose reconstructors to determine the CWT; they must also estimate the 

relative fraction of each tissue. 

The Livermore torso phantom was fabricated to provide calibration information for transuranic 

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 with the thickest overlays. There is some set-to-set variation.

The International Calibration of Detectors Systems for the Measurement of Low Energy Photon 

Emitters In Vivo (IAEA, 1991) contains background and efficiency information for the Phoswich 

detectors, using the Livermore phantom. In this comparison, 239Pu detection was based on uranium LX-rays, with a total intensity, Ix, of 0.0457 X-rays per disintegration. These measurements can be 
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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 claimant-favorable 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).

used to estimate the expected range of MDA values. Table 5-10 lists data for the measurements with 

239Pu-loaded lungs. These are the results of one participating facility with higher than average 

backgrounds and lower than average reported detection efficiencies. However, these values reflect 

those that are could have been encountered in a routine measurement program. Table 5-11 lists 

background and efficiency data for 239Pu counting. 

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

lower than those for 239Pu counting because:

• The reported counting efficiencies are about 40 – 200 times that for 239Pu. 
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Table 5-10. Uranium L X-ray intensity 

for decay of 239Pu. 

X-ray Energy - keV

X-ray Intensity – Xrays 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). 

CWTcmNormalized efficiency

counts/cm2/106 photons

Efficiency*

counts/photon

Efficiency**

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

40.3 0.84 0.00042 0.0426 102

50% adipose – 50% 

muscle

2.15 2.41 0.00120 0.1212 36

28.0 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

* For 500 cm2 detector

** L-xray intensity = 0.0457 L-xrays/disintegration

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

whole-body counting facility from pre-World War II steel. However, safety reports from the 

atmospheric testing era mention chest/lung counting. 

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 
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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

Efficiency*

counts/photon

Efficiency**

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

40.3 116 0.058 46.2 0.14

50% adipose – 50% 

muscle

2.15 142 0.0710 56.6 0.12

28.0 131 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

* For 500 cm2 detector

** 59.5 keV gamma intensity = 0.359

• The photon intensity for the 241Am gamma ray is 7.8 times that for the uranium X-rays 

emanating from the plutonium decay.

while maintaining proximity to the lung and bronchial region, thereby optimizing the number of 

photons measured. 

The chest wall is a variable shield and the CWT must be determined for accurate quantitative 

radioactivity measurements for low-energy gamma rays. The differing densities and thicknesses of 

muscles and adipose can result in significant differences in attenuation corrections. 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

CWT = 1.973 (W/H) – 2.0038

where:

CWT = chest wall thickness in centimeters

W = weight in pounds

H = height in inches

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). The 

MDAs for chest lung counts are listed in Attachment 5D, Table 5D-8. 

The current NTS contractor states that direct measurements of plutonium in the body, typically via 

lung counts, are capable of detecting only very large intakes or extended chronic exposures. 

Because such occurrences are rare, maintaining routine lung counting capability was not costeffective (BN 2003a).
Effective Date: 09/30/2004 Revision No. 00 Document No. ORAUT-TKBS-0008-5 Page 32 of 82

5.3.3

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

after detonation. 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. 

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). Section 5.2.2.1 includes information on the ratios 

of the various iodines in the period immediately following a shot. Dose reconstructors should use this 

for individuals directly involved in incidents such as Baneberry or reentry activities as indicated in the 

“other monitoring” section of the records provided by DOE.

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 Ge 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 whole-body counting 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

Air sampling records were considered workplace monitoring records rather than intake monitoring 

records, and were not maintained in the same collection as bioassay records. Air sampling records 

were associated with each facility and were 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 
Effective Date: 09/30/2004 Revision No. 00 Document No. ORAUT-TKBS-0008-5 Page 33 of 82

between air sample concentrations in given rooms or work locations and a specific person would be 

difficult. 

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). 

RASs were used at the decontamination facility (Decon Pad), during post-test drilling, tunnel reentry, 

and other special operations. REECo (1993a) describes specifics for collection and detection of air 

samples. Air samples were counted at the REECo laboratory in Mercury (A-23), the REECo Decon 

Pad in Area 6, and at the LANL laboratory in Mercury. Continuous air monitors (CAMs) were used at 

the Treatability Test Facility and on the drill rig floors during LANL drill-backs. The CAMs detected 

radioiodines, tritium, plutonium, or other radionuclides of concern. For iodine air sampling, bioassay 

records indicate if the air samples were collected with a charcoal canister or a filter (Arent and Smith

2004).

The Operation Plumbbob On-Site Rad Safety Report (REECo 1957) stated that 1945 air samples 

were collected for each atmospheric and underground test. As a specific example, the appendix 

describing decontamination of an A-9 balloon site where air samples were evaluated indicated the 

highest concentrations of airborne contamination existed during the loader operation (5.3 × 104 d/m 

per m3 at a sampling station approximately 100 feet downwind). Respirators were required above 

22.2 × 104 d/m per m3 beta/gamma concentration in air. More air sample results are available in that 

report.

Operation Hardtack Phase II On-Site Rad-Safe Report (REECo 1958) listed air sampling reporting 

levels for airborne particulate 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.

A 1959 memorandum entitled “Operational Guides for Above-Ground Drill Sites into Ground Zero 

Areas” (REECo 1959) stated that if significant beta-gamma concentrations are indicated, a 1-hr 

Staplex sample will be collected and nasal swabs will be collected from each person possibly 

exposed. 

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. This report stated that nonradioactive material was sufficient in the 

“dump” to prevent significant radiation levels from accumulating. Debris beta-gamma ratios during the 

period (1 to 7 months after the detonation) were variable from 2:1 to 10:1. 

In addition, REECo (1960) stated that airborne radioactivity in the mining work area was reduced by 

the liberal application of water and by natural water seepage. 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 that high concentrations of airborne radioactive material 
Effective Date: 09/30/2004 Revision No. 00 Document No. ORAUT-TKBS-0008-5 Page 34 of 82

were encountered and measurements were made difficult by accompanying high levels of radonthoron daughter products.

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 performed dry, and cored material was removed by a large 

vacuum system. In dry drilling operations, air activity was reduced by filtering the vacuum system 

exhaust, 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. Handling of “dried” core samples often resulted in significant airborne 

contaminants. Air hoods and respirators were used to prevent worker exposure to internal radiation.

“Operation Guides for Tunnel Areas” (REECo 1962) mentioned workplace area samples, radon/

thoron check methods, and nasal swabs. Operation Storax On-Site Radiological Safety Report 

(REECo 1964) mentioned air sample collection on the drill platform (high and low volume) at 

breathing level. “Dynamic Environmental Sampling Program” (REECo 1963) discussed air sampling 

during drill-backs.

Re-entry Problems Associated with Radiation from Underground Nuclear Detonations (Brown 1963) 

discussed radiological conditions for atmospheric versus underground testing. This paper stated that 

the entire post-test work environment was characterized by inhalation exposure problems, 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 report stated that coarse (>12 μm) and 

fine (<1 μm) particulate was ejected to several thousand feet.

Table 5-13 lists examples of historic limits for air samples for specific years from 1950 to 1987 

compiled from references (REECo date unknown b, Geiger and Whittaker 1961, 1985). Threshold 

levels include tolerance, MPC, reporting, exposure, detection, alert, and sensitivity levels. Respiratory 

protection and contamination control procedures are outlined in Attachment 5D, Section 5D.4.3. 

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 claimant-favorable to assume the result 

is real. 
Effective Date: 09/30/2004 Revision No. 00 Document No. ORAUT-TKBS-0008-5 Page 35 of 82

Table 5-13. Historic air sampling limits. 

Year Limit Note

1957 MPC in air above background of mixture of 

unknown radionuclides was 1 E-09 μCi/ml (beta and 

gamma) and 5 E-12 μCi/mL (alpha)

When radionuclides were known, MPC was 3 times 

above values based on exposures of 40-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

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

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, use the assumption provided in Internal Dose Reconstruction Implementation Guide

(NIOSH 2002); 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)]. If actual standard 

Uncertainty
Effective Date: 09/30/2004 Revision No. 00 Document No. ORAUT-TKBS-0008-5 Page 36 of 82

deviations or other indications of error are reported with a bioassay measurement result, use the 

reported value. 

REECo (1993a) cites NCRP (1987a) and Traub and Robinson (1987) for discussion of uncertainties. 

Uncertainties in intake estimates were used to select the best model and action, which should be 

taken to reduce errors to the extent reasonably achievable and give the most accurate intake estimate 

possible.

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 

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.

Less-Than Values

5.5.4

NTS was unique in the DOE weapons complex. Devices were assembled from materials produced 

elsewhere. Once assembled, the devices were detonated or in some cases during the safety tests,

blown up. The source term was specific to each testing category (atmospheric, tunnel, etc.) and in 

some cases quantities of specific radionuclides and ratios remain classified. The open literature 

contains no information about the ratios of plutonium and uranium for these tests. 

Determination of 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 being identified as contaminated; therefore, a 

worker who was not suspected of contamination would not have received the initial screening 

provided by nasal swab. Nasal swabs were taken on a regular basis, particularly for anyone involved 

in decontamination activities. If contamination was indicated, the individual received bioassay (usually 

urinalysis). 

• For an initial internal evaluation of a radiation worker diagnosed with cancer in a nonmetabolic 

organ, use the guidance in Maximum Internal Dose Estimates for Certain DOE Complex 

Claims (ORAU 2004a). 

• For a radiation worker with thyroid cancer whose records indicate that there was involvement

in reentry after a test or in venting or working drill-backs prior to 1963, evaluate iodines based 

on the information in Section 5.2.2.1. A reentry time after the test can be estimated from the 

access records and used in this determination.

• For a radiation worker with other metabolic cancers, review Attachment 5D to determine what 

other radionuclides would affect the development of the cancer and perform the evaluation 

based on the radionuclides that provided 90% of the dose. Attachment 5D provides 

information on the radionuclides present during various test categories.

• For a nonradiation worker (i.e., unmonitored for external and internal radiation exposure), 

base the internal evaluation on the assigned dose developed from Technical Basis Document 

for the Nevada Test Site – Occupational Environmental Dose (ORAU 2004b). Nonradiation 
Effective Date: 09/30/2004 Revision No. 00 Document No. ORAUT-TKBS-0008-5 Page 37 of 82

workers include but are not limited to office workers, most laborers after the atmospheric 

testing, janitors, cement workers, housekeepers, and cafeteria workers. 

• [Reserved – Guidance for monitored workers with no or few bioassay measurements is under 

development. Example from Hanford TBD follows.]

I. Workers with No Confirmed Intakes

A. Special consideration for Pu Am and Th – If intake suspected, but not confirmed, DR can use 

more sensitive urine results from later time to determine worst case intake at time when analysis was 

not as sensitive. Use MDA from later urine analysis or if there are many sample all showing no 

detection, then use 0.5 x MDA.

B. Worst Case Chronic Intakes – For workers with many results over a long time, but no confirmed 

intakes, a maximum chronic intake can be determined using the MDA of the last sample as the upper 

bound of excretion assuming chronic intake for the entire exposure period. The MDA (not the 

decision level) should be used.

II. Unmonitored Workers (robust radiological safety program – not much chance of missing a large 

intake.)

A. No bioassay record and no evidence of dosimeter issue, then internal dose ~ environmental intake 

only 

B. If wore dosimeter, then the internal dose is less than a monitored worker with no bioassay result 

above reporting level. Estimate upper bound using radionuclides of concern and MDAs.
Effective Date: 09/30/2004 Revision No. 00 Document No. ORAUT-TKBS-0008-5 Page 38 of 82

REFERENCES

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

New York, New York.

Arent, L., and C. Smith, 2003, “Personal Communications with Martha DeMarre, U.S. Department of 

Energy, Nevada Operations Office,” memorandum-to-file, NIOSH Dose Reconstruction Team.

Arent, L., and C. Smith, 2004, “Personal Communications with Martha DeMarre, U.S. Department of 

Energy, Nevada Operations Office,” memorandum-to-file, NIOSH Dose Reconstruction Team.

BN (Bechtel National, Inc.), 2000, Technical Basis for Internal Dosimetry at the Nevada Test Site, DEAC08-96NV11718, Mercury, Nevada.

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

TBD-E211-002, Mercury, Nevada.

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. 

Bolles, N. C., and N. E. Ballou, 1956, Report USNRDL-456, U.S. Naval Radiological Defense 

Laboratory, San Francisco.

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

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

DOE (U.S. Department of Energy), 1989, “Radiation Protection for Occupational Workers,” Order 

5480.11, Washington, D.C.

Geiger, E. L., and E. L. Whittaker, 1961, Analytical Procedures of the Radiological Safety Laboratory, 

REECo, Mercury, Nevada.

Glasstone, S., 1971, Public Safety & Underground Nuclear Detonations, U.S. Atomic Energy 

Commission, TID-25708, Washington, D.C.

Helgeson (Helgeson Scientific Service), 1967, [Reserved], information downloaded 8-8-04 from 

http://www.helge.com/unit301.htm. 

Holland, J. Z., 1964, “Physical Origin and Dispersion of Radioiodine – A Review,” In: Biology of 

Radioiodine (L. K. Bustad, Ed.), Pergamon Press, Oxford, pp. 15-23.

IAEA (International Atomic Energy Agency), 1991, “International Calibration of Detectors Systems for 

the Measurement of Low Energy Photon Emitters In Vivo,” private communication – Working 

Material Document.

Kathren, R. L., 1964, Activity and Thyroid Dose from Radioiodines, Nucleonics Data Sheet.

LASL (Los Alamos Scientific Laboratory), 1954, Analytical Procedures of the Industrial Hygiene 

Group, LA-1858, University of California, Los Alamos, New Mexico.
Effective Date: 09/30/2004 Revision No. 00 Document No. ORAUT-TKBS-0008-5 Page 39 of 82

LASL (Los Alamos Scientific Laboratory), 1958a, Analytical Procedures of the Industrial Hygiene 

Group, LA-1858 (2nd. Ed.), University of California, Los Alamos, New Mexico.

LASL (Los Alamos Scientific Laboratory), 1958b, Los Alamos Handbook of Radiation Monitoring, LA1835, University of California, Los Alamos, New Mexico.

LBNL (Lawrence Berkeley National Laboratory), 2004, WWW Table of Radioactive Isotopes, 

downloaded 6-1-04 from http://ie.lbl.gov/toi/

NAS (National Academy of Sciences), 2003, A Review of the Dose Reconstruction Program of the 

Defense Threat Reduction Agency, Washington, D.C. 

NCRP (National Council on Radiation Protection and Measurement), 1987a, Use of Bioassay 

Procedures for Assessment of Internal Radionuclide Deposition, Report 87, Washington, D.C.

NCRP (National Council on Radiation Protection and Measurements) 1987b, Exposure of the 

Population in the United States and Canada from Natural Background Radiation, Report 94, 

Washington, D.C.

NIOSH (National Institute for Occupational Safety and Health), 2002, Internal Dose Reconstruction 

Implementation Guidelines, Rev. 1, OCAS-IG-001, Office of Compensation Analysis and 

Support, Cincinnati, Ohio.

NRDL (Naval Radiological Defense Laboratory), 1968, Hazards to Personnel Re-Entering the NTS 

following Nuclear Reactor Tests, NRDL-TR-66-149, San Francisco, California.

ORAU (Oak Ridge Associated Universities), 2004a, Technical Information Bulletin, Maximum Internal 

Dose Estimates for Certain DOE Complex Claims, ORAU-TIB-0002, Rev. 1, Oak Ridge, 

Tennessee.

ORAU (Oak Ridge Associated Universities), 2004b, Technical Basis Document for the Nevada Test 

Site – Occupational Environmental Dose, ORAU-TKBS-0008-4, Rev. 0, Oak Ridge, 

Tennessee.

Pan Am (Pan American Airways), 1967, Memorandum dated August 4, 1967, and draft report, “Report 

on State of the Art of Whole-Body Counting,” Jackass Flats, Nevada.

REECo (Reynolds Electrical & Engineering Company, Inc.), 1957, Operation Plumbbob On-Site RadSafety Report, Mercury, Nevada.

REECo (Reynolds Electrical & Engineering Company, Inc.), 1958, Operation Hardtack Phase II OnSite Rad-Safe Report, OTO 58-5, Mercury, Nevada.

REECo (Reynolds Electrical & Engineering Company, Inc.), 1959, “Operational Guides for AboveGround Drill Sites into Ground Zero Areas,” Information Bulletin #24, Mercury, Nevada.

REECo (Reynolds Electrical & Engineering Company, Inc.), 1960, Radiological Safety for 

Underground Nuclear Explosions, RRS-60-3, Mercury, Nevada.

REECo (Reynolds Electrical & Engineering Company, Inc.), 1961, Radiological Safety Division 

Standard Operating Procedure, Mercury, Nevada.
Effective Date: 09/30/2004 Revision No. 00 Document No. ORAUT-TKBS-0008-5 Page 40 of 82

REECo (Reynolds Electrical & Engineering Company, Inc.), 1962, “Operation Guides for Tunnel 

Areas,” Information Bulletin #20, Mercury, Nevada.

REECo (Reynolds Electrical & Engineering Company, Inc.), 1963, “Dynamic Environmental Sampling 

Program,” Information Bulletin #92, Mercury, Nevada.

REECo (Reynolds Electrical & Engineering Company, Inc.), 1964, Operation Storax On-Site 

Radiological Safety Report, NVO-162-14, Mercury, Nevada. 

REECo (Reynolds Electrical & Engineering Company, Inc.), 1968-1991a, Radiological Safety Division, 

Standard Operating Procedures, 1968-1991, Chemical and Radiological Analysis, Mercury, 

Nevada.

REECo (Reynolds Electrical & Engineering Company, Inc.), 1968-1991b, Radiological Safety Division, 

Standard Operating Procedures, 1968-1991, Instrumental Radiological Analysis and 

Laboratory Instruments and Equipment, Mercury, Nevada.

REECo (Reynolds Electrical & Engineering Company, Inc.), 1977-1987, Environmental Sciences 

Standard Procedure, Appendix A, “Spectrometer Calibrations and Detection Limits,” Mercury, 

Nevada.

REECo (Reynolds Electrical & Engineering Company, Inc.), 1985, Environmental Sciences 

Department, Radioanalytical Laboratory Procedures, Section 7, “Determination of 

Radiokrypton and Radioxenon in Air,” Mercury, Nevada.

REECo (Reynolds Electrical & Engineering Company, Inc.), 1993a, Technical Basis for Internal 

Dosimetry at the Nevada Test Site, DOE/NV/10630-64, Mercury, Nevada.

REECo (Reynolds Electrical & Engineering Company, Inc.), 1993b, list by year of instruments, 

dosimetry (mostly external), rad levels, exposure criteria, respiratory protection, internal 

contamination, and calibration information, From the O drive, file titled “NTS General 

Information 1953 – 1973.” 

REECo (Reynolds Electrical & Engineering Company, Inc.), date unknown a, Historical Perspective 

Notes, Internal Doses, Mercury, Nevada.

REECo (Reynolds Electrical & Engineering Company, Inc.), date unknown b, General Radiation 

Protection Considerations, CIC 317286, Mercury, Nevada. 

REECo (Reynolds Electrical & Engineering Company, Inc.), date unknown c, Iodine 1960-1963, 

Mercury, Nevada.

Rich, B.L., Hinnefeld, S.L., Lagerquist, C.R., Mansfield, W.G., Munson, L.H., Wagner, E.R., Vallario, 

E.J., Health Physics Manual of Good Practices for Uranium Facilities, EGG-2530, UC-41, June 

1988.

Smith, F., D. W. Boddy, and M. Goldman, 1952, “Biological Injury from Particle Inhalation, JANGLE 

Project 2.7," WT-372, in Operation JANGLE, Biological Hazards, WT-396, National Institutes 

of Health. Washington, D.C., June.
Effective Date: 09/30/2004 Revision No. 00 Document No. ORAUT-TKBS-0008-5 Page 41 of 82

Taplin, G. V., O. M. Meredith, Jr., and H. Kade 1958, Evaluation of the Acute Inhalation Hazard from 

Radioactive Fall-out Materials by Analysis of Results from Field Operations and Controlled 

Inhalation Studies in the Laboratory, WT-1172, Project 37.3, performed by University of 

California at Los Angeles for Civil Effects Test Group. 

Teasdale, C., 1985, Whole Body Counting Facility, Nevada Test Site, RRS-85-22, Reynolds Electrical 

& Engineering Company, Inc., Mercury Nevada.

Traub, R. J., and A. V. Robinson, 1987, “The Sources of Uncertainties Associated with Internal Dose 

Calculations,” Proceedings of the DOE Workshop on Radiobioassay and Internal Dosimetry, 

PNL-SA-14043.
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ATTACHMENT 5D

OCCUPATIONAL INTERNAL DOSE FOR MONITORED WORKERS 

TABLE OF CONTENTS

Section

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

Page

44

5D.1 Occupational Internal Dose.................................................................................................... 46

5D.2 Bioassay Codes and In Vitro Minimum Detectable Activities and Detection Levels ................ 46

5D.2.1 Codes Used in Bioassay Records............................................................................. 46

5D.2.2 In Vitro Analyses for Individual Radionuclides........................................................... 49

5D.3 In Vivo MDAS and Reporting Practices at NTS...................................................................... 53

5D.3.1 Whole-Body Counting............................................................................................... 53

5D.3.2 Chest Counting......................................................................................................... 54

5D.3.3 Thyroid Counts ......................................................................................................... 54

5D.4 Other NTS Information ........................................................................................................... 55

5D.4.1 Radionuclides of Concern and Specific Bioassay Programs for NTS 

 Facilities.................................................................................................................... 55

5D.4.2 Incidents ...................................................................................................................63

5D.4.3 Respiratory Protection Practices at NTS ................................................................... 69

5D.4.4 Historical Practices and Contamination Levels.......................................................... 70

5D.5 Reference Tables for Determining Internal Dose.................................................................... 72

References ........................................................................................................................................77

Glossary ............................................................................................................................................80
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LIST OF TABLES

Table

5D-1 Codes for analyte................................................................................................................

Page

46

5D-2 Codes for body parts........................................................................................................... 47

5D-3 Codes for radionuclides and sample types.......................................................................... 48

5D-4 Codes for sample types....................................................................................................... 49

5D-5 Codes for units....................................................................................................................49

5D-6 Limits of detection for urine and fecal analysis .................................................................... 50

5D-7 1993 whole-body counting MDAs........................................................................................ 53

5D-8 1993 MDAs for chest (lung) counting .................................................................................. 54

5D-9 MDAs for thyroid counts ...................................................................................................... 55

5D-10 Drill-back resuspension and mine back containment loss radionuclides for 

identification versus time after test ...................................................................................... 56

5D-11 Drill-back resuspension, reentry/mine back resuspension, and 

decontamination facility, isotopes of concern for dose versus time after test ....................... 57

5D-12 Decontamination facility, isotopes for identification versus time after test............................ 59

5D-13 Atmospheric weapons test areas, isotopes for identification and of concern for 

dose....................................................................................................................................59

5D-14 Low-level waste site (A-3), isotopes for identification and of concern for dose .................... 60

5D-15 Low level waste site (A-5), isotopes for identification and of concern for dose..................... 60

5D-16 Radiation instrument calibration facilities, isotopes for identification and of 

concern for dose ................................................................................................................. 60

5D-17 Radiation instrument calibration facilities, isotopes for identification and of 

concern for dose ................................................................................................................. 61

5D-18 Radiochemistry and counting laboratories, isotopes for identification and of 

concern for dose ................................................................................................................. 61

5D-19 Isotopes of concern for dose, summary list ......................................................................... 62

5D-20 Current nuclides of concern for NTS locations .................................................................... 62

5D-20 Releases from underground tests........................................................................................ 64

5D-22 Historical NTS respiratory protection action levels............................................................... 70

5D-23 NTS historical contamination limits...................................................................................... 72

5D-24 Solubility types for radionuclides found at NTS.................................................................... 73

5D-25 Fission products up to 1 year old as identified in 1959 documentation ................................ 74

5D-26 Other common radionuclides not normally a part of fission products that might 

be present...........................................................................................................................74

5D-27 Radionuclide activity ratios at formation (immediately after detonation)............................... 75

5D-28 Specific activity of selected alpha emitters .......................................................................... 76
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ACRONYMS AND ABBREVIATIONS

AMAD activity-median aerodynamic diameter

CA contamination area

cc cubic centimeter

Ci curie

cpm counts per minutes

DOD U.S. Department of Defense

DOE U.S. Department of Energy

dpm disintegration per minute

DPP drift protection plug

DU depleted uranium

EG&G Edgerton, Germeshausen, and Grier Corporation

g gram

GFP Gross Fission Product

H time of detonation 

HEPA high-energy particulate air (filter)

HEU highly enriched uranium

H&N Holmes & Narver

hr hour

ICRP International Commission on Radiation Protection

IMBA Integrated Modules for Bioassay Assessment

keV kilo (thousand) 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

LOS line of sight

LRL Lawrence Radiation Laboratory

m meter

MDA Minimum Detectable Activity; Minimum Detectable Amount

MDI Minimal Detectable Intake

mL milliliter

mrem millirem

MSA Mine Safety Appliances Company

nCi nanocurie 

NTS Nevada Test Site

OBP overburden plug
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OCAS Office of Compensation Analysis and Support

P purge or controlled ventilation

pCi picocurie

PPE personnel protection equipment

R release 

RCT radiological control technician

REECo Reynolds Electrical & Engineering Company, Inc.

RSN Raytheon Services Nevada

SNL Sandia National Laboratories

TTR Tonopah Test Range

WBC Whole-Body Counter

WEF Waste Examination Facility

μCi microcurie
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5D.1 OCCUPATIONAL INTERNAL DOSE

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.

5D.2 BIOASSAY CODES AND IN VITRO MINIMUM DETECTABLE ACTIVITIES AND 

DETECTION LEVELS

5D.2.1

Employer codes and job titles for NTS contractors [Reynolds Electrical & Engineering Company 

(REECo), Edgerton, Germeshausen, and Greer Corporation (EG&G), Holmes & Narver (H&N), and 

Raytheon Services Nevada (RSN)] are available to dose reconstructors on the O Drive maintained by 

the Office of Compensation Analysis and Support (OCAS) of the National Institute for Occupational 

Safety and Health. The computerized bioassay records contain the codes listed in Tables 5D-1 

through 5D-5 (DeMarre 2003).

Codes Used in Bioassay Records

Table 5D-1. Codes for analyte. 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)

These codes were not identified with a particular period. The measurements of alpha, beta, gamma, 

and gross fission products 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. If swipes were taken while working on a safety test 

(post test), they were used as an indicator that plutonium might be present in the sample. For 

gamma, when a multichannel analyzer was used, specific isotopes could be cited if they were 

detectable. For example, one could occasionally detect ruthenium.

The main concerns were tritium, iodines, and plutonium (Arent and Smith 2003). 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 WholeBody Counter (WBS) at this 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 
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uranium, “turballoy” for natural uranium, and “D-38” for depleted uranium. “LT” or “-“ in the bioassay 

records means less than the detection limit. The code “-99” means not detected (Arent and Smith

2004). Table 5D-2 lists body part codes and abbreviations that are specific to different periods, as 

listed.

Table 5D-2. Codes for body parts. 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

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”.
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Table 5D-3 lists codes for internal radionuclides and sample types 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 5D-3. Codes for radionuclides and sample types.

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

These codes 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 

2004). 

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 

2004). 

Table 5D-4 lists codes for the type of sample to be analyzed.

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. Dose reconstructors should review data from the original forms, not the microfiche “deadbio”

index.
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Table 5D-4. Codes for sample types. 

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 5D-5. Codes for units. 

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

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

Code Unit Description

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

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

68 μg/g micrograms per gram

69 μg/L micrograms per liter

70 μg micrograms U (Reg: KPA analysis)

71 μS/cm microseimens per centimeter

72 mS/cm milliseimens per centimeter

5D.2.2

Table 5D-6 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
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Table 5D-6. 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 1988(c)-1999(d) 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-present(d) 200 - REECo 1993a

Na-24 1970-present(d

) 100(e)

- REECo 1993a

Mn-54 1970-present(d) 100- REECo 1993a

Co-57 1970-present(d) 100- REECo 1993a

Fe-59 1970-present(d) 600- REECo 1993a

Co-60 1970-present(d) 50- REECo 1993a

Sr-85 1970-present(d) 200- REECo 1993a

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

Sr-90 1970-1999(d) 0.8(g)- REECo 1993a

Sr-90 2000-present1- BN 2000, 2003

Zr-95 1970-present(d) 600 - REECo 1993a

Nb-95 1970-present(d) 100- REECo 1993a

Tc-99m 1970-present(d) 100- REECo 1993a

Rh-101 1970-present(d) 100 - REECo 1993a

Rh-102 1970-present(d) 100- REECo 1993a

Rh-102m 1970-present(d) 200- REECo 1993a

Ru-103 1970-present(d) 200- REECo 1993a

Ru-106 1970-present(d) 1000 - REECo 1993a

Sb-122 1970-present(d) 100- REECo 1993a

Sb-124 1970-present(d) 100- REECo 1993a

Sb-125 1970-present(d) 300 - REECo 1993a

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

I-131 1970-present(d) 100- REECo 1993a

I-132 1970-present(d) 90- REECo 1993a

Te-132 1970-present(d) 100- REECo 1993a

Ba-133 1970-present(d) 200- REECo 1993a

Ba-133m 1970-present(d) 100 or 400(h) - REECo 1993a

I-133 1970-present(d) 100 - REECo 1993a

Cs-134 1970-present(d) 500- REECo 1993a

I-135 1970-present(d) 200- REECo 1993a

Cs-136 1970-present(d) 100 - REECo 1993a

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

Cs-137 1970-present(d) 100- REECo 1993a

Ce-139 1970-present(d) 100- REECo 1993a

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

Ba-140 1970-present(d) 500- REECo 1993a

La-140 1970-present(d) 200- REECo 1993a

Ce-141 1970-present(d) 200 - REECo 1993a

Ce-143 1970-present(d) 200- REECo 1993a

Ce-144 1970-present(d) 800- REECo 1993a

Nd-147 1970-present(d)400 - REECo 1993a

Eu-152 1970-present(d) 300- REECo 1993a

Eu-154 1970-present(d) 200- REECo 1993a
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Table 5D-6 (Continued). Limits of detection for urine and fecal analysis.

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

Eu-155 1970-present(d) 500 - REECo 1993a

Yb-169 1970-present(d) 600 - REECo 1993a

Ta-182 1970-present(d) 300 - REECo 1993a

W-187 1970-present(d) 300 - REECo 1993a

Ir-192 1970-present(d) 100 - REECo 1993a

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

Fission products- GFP 1988(c)-present(d) 3(j) - 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) 1988(c)-present 100 - BN 2000, 2003

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

Ra-226 1970-present(d) 300 0.4 or 0.04(h) REECo 1993a

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

Th-228 1970-present(d) - 0.01 REECo 1993a

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

Th-230 1970-present(d) - 0.01 REECo 1993a

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

Th-232 1993-1999(d) - 0.01 REECo 1993a

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

U-234 1970-present(d) 0.02 - REECo 1993a

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

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

U-235 1970-present(d) 0.01 0.008 REECo 1993a

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

U-238 1970-1999(d) 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/L(o) - REECo 1993a

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

Pu-238 1988(c)-1999(d) 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-239(p) 1954-1957(d) 2 dpm/24 hr (detection limit)(q) - LASL 1954

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

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

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

Pu-239 1988(c)-2000(d) 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-241(t) 1954-1957(d) 2 dpm/sample (detection limit)(u) - LASL 1954

Am-241(t) 1958-1981(d) 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 1988(c)-1999(d) 0.03(w) 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. 
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Table 5D-6 (Continued). Limits of detection for urine and fecal analysis.

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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5D.3 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.

5D.3.1

Consistent with industry-wide improvements in this area, a variety of hardware and software 

applications have been used for whole-body counting. In 1993, individuals in the routine bioassay 

program received employee, annual, and termination whole-body counts. Employees who received 

routine bioassay included radiation protection technicians 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. 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; use these when they are available. Table 

5D-7 lists default MDAs for various radionuclides. 

Whole-Body Counting

Table 5D-7. Whole-body counting MDAs. 

Radionuclide

1967

MDA 

(nCi)

1993

MDA

(nCi)

Na-22 3.0 0.7

Na-24 3.0 1.1

Sc-46 3.0 1.1

Mn-54 3.0 1.2

Fe-59 5.3 2.3

Co-57 3.5 1.4

Co-60 1.3 0.9

Sr-85 3.0 1.8

Zr-95 2.1 1.9

Nb-95 3.0 1.1

Mo-99 23 10(a)

Tc-99m 3.4 7.1

Ru-103 3.4 1.2

Ru-106 30 11

Rh-104 4.1 1.4(b)

Rh-102m 6.5 7.9(b)

Rh-102 3.2 1.0

Sb-122 4.2 2.1

Sb-124 3.1 1.6

Sb-125 10 3.6

Radionuclide

1967

MDA 

(nCi)

1993

MDA

(nCi)

Te-132 3.4 1.5(c)

I-131 3.7 1.6(c)

I-132 3.0 1.4(c)

I-133 3.5 1.4(c)

I-135 10 1.4(c)

Cs-134 3.1 1.5

Cs-136 3.0 1.4

Cs-137 3.0 1.7

Ba-133m NA 1.1

Ba-133 5 1.8

Ba-140 12 4.8

La-140 3.1 0.4

Ce-139 3.7 1.4(b)

Ce-141 6.2 2.6(b)

Ce-143 7.1 1.4(b)

Ce-144 NA 12(b)

Nd-147 NA 5.5(b)

Eu-152 11 4.2(b)

Eu-154 7.4 3.2(b)

Eu-155 14 4.1(b)

Radionuclide

1967

MDA 

(nCi)

1993

MDA

(nCi)

Yb-169 NA 0.2

Lu-174m NA 3.7(b)

Lu-174 NA 2.5(b)

Ta-182 NA 3.3(b)

W-181 NA 4.6(b)

W-187 13 4.6

Ir-192 3.6 0.9

Ra-226 & progeny 50(4)(b,d)

Ac-227 57 28(b,e) Th-228 75 33(b,f)

Pa-231 NA 6.7(b)

Th-232 NA 5(b,g)

U-233 NA 13(b)

U-235 5.6 3.4(b)

U-237 14 4.8(b)

Np-237 8.3 2.8(b,h)

Np-239 13 5.6(b)

Am-241 NA 4.3(b)

Am-234 NA 2.8(b) 243Cm 13 4.9(b)

Source: REECo (1993a) 

a. Only bioassay method identified for this radionuclide.

b. In conjunction with lung count.

c. In conjunction with thyroid count.

d. 50 nCi for Ra-226; 4 nCi based on Pb-241 in equilibrium with Ra-226. 

e. Based on Th-227 in equilibrium with Ac-227.

f. Based on Ra-224 in equilibrium with Th-228.

g. Based on Ac-228 in equilibrium with Th-232.

h. Protactinium-233 is used to measure Np-237 by whole-body count and the MDA for this procedure is 2.8 nCi of Pa-233 which, in 

equilibrium with Np-237, also represents 2.8 nCi of Np-237. 

i. “NA” means there was no gamma with energy greater than 100 keV on which to base the estimate.
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The 137Cs whole-body count sensitivity for 1968 to 1991 was 0.011 μCi for a 20-minute count and 

0.010 μCi for a 40-minute count. The Maximum Permissible Body Burden of 137

Cs in standard man is 

30 μCi.

5D.3.2

Based on the 1993 internal technical basis document (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 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.

Chest Counting

MDAs were generated individually for counts by processing software, with typical values listed in 

Table 5D-8.

Table 5D-8. 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.

5D.3.3

Based on the 1993 internal technical basis document (REECo 1993), thyroid counts occurred as soon 

as practicable after a suspected radioiodine uptake. Processing software generated individual MDAs 

for counts, with typical values listed in Table 5D-9.

Thyroid Counts
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Table 5D-9. 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) 

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 

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

mean as soon as possible.

Table 5D-21 lists tests and incidents for which dose reconstructors 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 dose reconstructor 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 after a test. From these records, the dose reconstructor 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 5-2. Without 

access records, base internal exposure on environmental levels. Compare the radionuclides listed in 

Table 5D-9 to the lists in Section 5D.4 and decide which radionuclides are most significant to dose 

based on the cancer location. 

5D.4 OTHER NTS INFORMATION 

5D.4.1

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 radionuclides of 

concern were determined in the following way:

Radionuclides of Concern and Specific Bioassay Programs for NTS Facilities

• 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.
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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.

Drill-backs (Yucca Flat and Pahute Mesa) 

Drill-back operations took place within days or weeks of an underground nuclear weapons test. 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 1963, engineering devices were used to prevent the 

escape of radioactive gases and particulates; however, there was a potential for release in some 

instances. The pathways for release and personnel exposure were (1) loss of containment during 

drilling or coring, and (2) resuspension of particulate fission or activation products during the coring 

operations. 

Radionuclides for identification of a problem from containment loss and dose concern were 131I, 133I, 

and 137Cs. Radionuclides for identification of a problem from resuspension and used to trigger 

searches in bioassays for the radionuclides of dose concern for resuspension are listed in Table 5D10. Radionuclides of dose concern from drill-back resuspension are listed in Table 5D-11.

Table 5D-10. Drill-back resuspension and mine back 

containment loss radionuclides for identification versus 

time after test. 

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

Mn-54 Mn-54 Mn-54 Mn-54 Co-60

Fe-59 Fe-59 Fe-59 Co-57 Sr-90

Co-57 Co-57 Co-57 Co-60 Sb-125

Co-60 Co-60 Co-60 Sr-90 Ba-133

Y-91 Sr-89 Sr-89 Y-91 Cs-137

Zr-95 Sr-90 Sr-90 Zr-95

Zr-97 Y-91 Y-91 Nd-95

Mo-99 Zr-95 Zr-95 Ru-106

Tc-99m Nd-95 Nd-95 Sb-125

Ru-103 Mo-99 Ru-103 Ba-133

Ru-106 Ru-103 Ru-106 Cs-135

Sb-124 Ru-106 Sb-124 Ce-139

Sb-125 Sb-124 Sb-125 Ce-144

I-131 Sb-125 Ce-139 Pm-147

I-133 I-131 Ce-141 Ta-182

I-135 Te-132 Ce-143

Cs-137 Ba-140 Ce-144

Ba-140 La-140 Ta-182

La-141 Ce-141

Ce-143 Ce-144

Ce-144 Nd-147

Nd-147 Ta-182

Pm-149

Ta-182
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Table 5D-11. Drill-back resuspension, reentry/mine back 

resuspension, and decontamination facility, isotopes of 

concern for dose versus time after test. 

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

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 isotope identification. The primary isotopes used for 

identification of an intake were 131I, 133I, and 137Cs and those listed in Table 5D-10. Annual wholebody 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

For the years before the establishment of engineering controls, the dose reconstructor will need to 

evaluate the potential for exposure, based on review of the access records provided in the claim file 

provided by DOE. 

Reentry and Mine Back (A-1& A-12)

Reentry and mine back operations were similar to those for drill-backs. The major differences were 

that the operations took place in a confined underground environment, the time after a test that the 

cavity was entered, the horizontal method of drilling and, in some cases, the cavity was opened and 

personnel entered. 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 
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when the line-of-sight (LOS) pipe was opened to remove experiment equipment and samples. 

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 5D-10. 

Radionuclides of dose concern from resuspension are listed in Table 5D-11. 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 done 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. 

Routine Tunnel Operations (A-1 and A-12)

Tritium was the isotope of dose concern; routine bioassay was to collect quarterly urine samples 

which were processed for 3

H. (No bioassay information was given in REECo 1993a.) See Section 

5.2.2.4 for a discussion on assigning tritium dose.

Decontamination Facility (A-6)

Isotopes used for identification of a problem at this facility are listed in Table 5D-12. Isotopes of dose 

concern are listed in Table 5D-11. Routine bioassay was to collect quarterly urine samples and 

conduct annual whole-body counts. Urine samples were gamma counted and analyzed for GFP, 

238Pu, 239Pu, 240Pu, and 241Am. 

Test Treatability Facility (A-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. Isotopes 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. 

Atmospheric Weapon Safety Tests (A-3 and A-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. 
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Table 5D-12. 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

Atmospheric Weapons Test Areas (all areas except A-22, 23, and 27 due to test location or 

airborne dispersion)

Isotopes for identification of a problem and dose concern are listed in Table 5D-13. 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 5D-13. Atmospheric weapons test areas, 

isotopes for identification and of concern for dose. 

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

Low-Level Waste Site (A-3) 

Isotopes for identification of a problem and dose concern are listed in Table 5D-14. Routine bioassay 

was to collect quarterly urine samples and conduct annual whole-body counts. Urine samples were 

gamma analyzed for GFP, plutonium, and americium. 
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Table 5D-14. Low-level waste site (A-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

Low-Level Waste Site (A-5)

Isotopes for identification of a problem and dose concern are listed in Table 5D-15. 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 5D-15. Low level waste site (A-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

Radiation Instrument Calibration Facilities 

Isotopes for identification of a problem and dose concern are listed in Table 5D-16. 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. 

Table 5D-16. 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

Radiography Operations 

The isotope of concern was 92Ir. Bioassay (urine/whole-body count) was conducted if a loss of source 

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

follow-up actions taken accordingly.

Well Logging Operations

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 
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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. 

Device Assembly Facilities (A-27) 

Isotopes for identification of a problem and dose concern were 3

H, 235U, 239Pu, and 240Pu. No 

bioassay information was listed.

Nuclear Rocket Development Areas (A-25) 

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. 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. 

Radioactive Source Storage Areas

Isotopes for identification of a problem and dose concern are listed in Table 5D-17. 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 5D-17. 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

Radiochemistry and Counting Laboratories 

Isotopes for identification of a problem and dose concern are listed in Table 5D-18. 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 5D-18. 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

Radionuclides of dose concern for NTS are summarized in Table 5D-19. Table 5D-20 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.
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Table 5D-19. Isotopes 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 5D-20. 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 Ru-102m

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

Radiation instrument calibration facilities H-3 Co-60 Sr-90 Cs-137

Ra-226 U-235 Pu-238 Pu-239

Am-241

Radiography operations Co-60 Cs-137 Ir-192

Well logging operations Co-60 Cs-137 Ra-226/Be-7 

Pu-238/Be-7 Am-241/Be-7

Nuclear Rocket Development Area (Area 25) Co-60 Sr-90 Cs-137 Nb-95(a)

U-234 U-235 U-238

Legacy biokinetic test areas (Test Cell A, Area 25) Am-241

Radioactive source storage areas Na-22 Mn-54 Co-57 Co-60

Ni-63 Sr-90 Cd-109 Ba-133

Cs-137 Eu-152 Eu-154 Ir-192

Ra-226 Th-228 U-235 Pu-238

Pu-239 Am-241

Radiochemistry and counting laboratories H-3 Na-24 Co-57 Co-60

Sr-85 Y-88 Y-90 Sr-90

Cd-109 Sn-113 Eu-152 Cs-137

Ce-139 Ce-144 Hg-203 Ra-226

Th-232 U-234 U-238 Pu-239

Pu-242 Am-241 Am-243 Cm-244

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
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5D.4.2

Based on guidance from NTS, the dose reconstructor should do the following when researching a 

claim: 

Incidents

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

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.

Establish if the claimant is linked to a release by job title or work location. 

In addition, the dose reconstructor might find the information in Table 5D-21 [from The Containment of 

Underground Nuclear Explosions (OTA 1989)] and Schoengold, DeMarre, and Kirkwood (1996) useful 

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: 

• The Report of the Test Manager for 56 Project – NTS, Part VI, Chapter 1, “Special Incidents,” 

describes an overexposure during a plutonium dispersion test (AEC 1956). 

• 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 

alpha-emitting 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 (air supplied respiratory and full anticontamination clothing (clothing that might have 

consisted of coveralls, hoods, booties, and gloves), nasal swabs taken).

• A memo for B Tunnel, U12b (AEC 1961), 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, 
Effective Date: 

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Table 5D-21. Releases from underground tests.

Test name Date Type Purpose & yield Sponsor Location

Isotopes 

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, 144CeTest/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 

analysis(f): 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 failure(a)

Test release at R+12 

hours: 360(b) 

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 failure(a)

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.
Effective Date: 

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Table 5D-21 (Continued). Releases from underground tests.

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.

Mini Jade 5/26/83 Tunnel Weapons effects

<20 kt

LANL

DOD 

U12n.12 Xe-133, Xe133m Controlled 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.
Effective Date: 

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Table 5D-21 (Continued). Releases from underground tests.

Test name Date Type Purpose & yield Sponsor Location

Radionuclides

identified in 

release Release type & notes Release (Ci) Release summary

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 failure(a)

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.

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)
Effective Date: 

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Table 5D-21 (Continued). Releases from underground tests.

Test name Date Type Purpose & yield Sponsor Location

Radionuclides

identified in 

release Release type & notes Release (Ci) Release summary

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 1970-1988(d) 5.5 E+03

Table 5D-21 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).

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). 
Effective Date: 09/30/2004 Revision No. 00 Document No. ORAUT-TKBS-0008-5 Page 68 of 82

• 108 miners approached the quarterly exposure limit in 1961 due to tritium exposure in U12b. 

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. Urinalysis results showed that 280 workers received 50 

mrem or more, and the maximum internal dosage was 2,435 mrem. Appendix D of this report, 

“Typical Detailed Safety Support Plan for Underground Nuclear Tests,” describes radiological 

support to post-shot 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 NTS-SOP-0524 limits. Two post-shot 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) describes an iodine incident resulting in doses of 27 and 31 rem 

to the thyroid (131I and 133I, respectively) for two employees involved in post-shot drilling. 

These levels were above permissible exposure limits. Six employees received measurable 

thyroid radiation exposures. A Type B Investigation was conducted (REECo 1965b).

• Extensive records for the 1970 Baneberry test are available on the OCAS O Drive. 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 the 

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 the 

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.

• A report entitled Evaluation of Protection, Bioassay, and Dose Assessment Programs for 

Internal Radiation Exposures at the 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 (1) 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; (2) the exposure of one worker in June 1995 to 239Pu aerosols 

believed to have been generated while handling and characterizing high-specific-activity 

fragments found during the Double Track Site soils characterization project; and (3) 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. 
Effective Date: 09/30/2004 Revision No. 00 Document No. ORAUT-TKBS-0008-5 Page 69 of 82

• French and Skrable noted that the bioassay data from the E-Tunnel workers were limited in 

both the quantity and quality needed for making reasonable accurate estimates of their intakes 

and doses. 

• French and Skrable estimated the worker at the Double Track Site (239Pu aerosol) had an 

effective ingestion intake of 3,260 pCi from nonrespirable aerosols with a committed effective 

dose of 0.2 mrem, and an inhalation intake of 360 pCi of 1 μm AMAD class Y aerosols with a 

committed effective dose of 90 mrem. The activity of 239Pu in fecal samples of the 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 (French and Skrable 1995). 

• For the tritium exposure, French and Skrable concluded that considerable bioassay and air 

sampling data confirmed that the exposures were minimal with little dose consequence, and 

no further dose estimates were provided. In 1995, the previous contractor (EG&G) discovered 

contamination in the same basement as a result of routine bioassay samples from workers. 

The source of contamination was later found to have resulted from a modification of a sealedtube neutron generator by a scientist employed by that contractor. This modification included 

breaching the sealed tube that contained three tritide disks (130 Ci tritium total). According to 

the DOE NTS Annual Site Environmental Report for Calendar Year 1995 (Black and 

Townsend 1996), the maximum release into the building was 123 millicuries. 

5D.4.3

The following paragraphs provide examples of NTS respiratory protection practices. Historical 

accounts of respirator usage are included to give the dose reconstructor 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 56 was an atmospheric safety experiment that studied plutonium dispersal in NTS Area A-11 

from 1955 to 1956. Documentation (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 personnel protection equipment 

(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 in 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 
Effective Date: 09/30/2004 Revision No. 00 Document No. ORAUT-TKBS-0008-5 Page 70 of 82

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.”

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, 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 (Ranier 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 will 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 5D-22 lists NTS respiratory protection levels for selected years (REECo 1993b).

Table 5D-22. 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
Effective Date: 09/30/2004 Revision No. 00 Document No. ORAUT-TKBS-0008-5 Page 71 of 82

Nasal swabs 4,230

Facial swipes 109

Air samples 1,945

Urine kits issued 73

Respirators issued 10,755

5D.4.4

The Plumbbob On-Site Radiological Safety Report (REECo 1957) is an example of the magnitude of 

early radiological support; it states the following samples were collected:

Historical Practices and Contamination Levels 

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 5D-23 lists NTS contamination limits for selected years. Contamination limits were often 

used as indicators of when bioassay samples were to be taken.
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5D.5 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. For determining 

dose, assume that acute exposure occurred within one work shift. 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 “other monitoring” section of 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 5D-21). 

Note: If the file provided by DOE does not include “other monitoring,” dose reconstructors should 

base the internal dose assigned to the worker on the environmental ambient. 

The following tables provide information about the characteristics of the source term at NTS. The 

dose reconstructor needs to keep in mind 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 

dose reconstructor might have to address. The dose reconstructor should use professional judgment 

when including or deleting specific radionuclides in the source term based on the test category with 

which the worker was associated and the cancer type. 

Table 5D-23. 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 the coveralls are > 1 × 103 dpm/55 cm2.

Decon personnel at >:

• Outer clothes: 500 cpm/55 cm2 (alpha) or 7 mR/hr (beta + gamma)

• Shoes: 500 cpm/55 cm2 (alpha) or 7 mR/hr (gamma)

• Skin or underclothing: 100 cpm/55 cm2 (alpha) or 1 mR/hr (gamma)

• Equipment: <7 mR/hr (gamma) or 500 cpm/55 cm2 (fixed alpha)

Respiratory devices: < 1 mR/hr (beta + gamma) or 100 cpm/55 cm2 (alpha fixed and removable)

Alpha contamination >5,000 cpm/55 cm2 requires field decon.

Access permits required to enter areas with beta-gamma levels >10 mR/hr or with 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 (beta-gamma) 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 
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Table 5D-24. Solubility types for radionuclides found at NTS. 

Radionuclides of concern

Solubility 

type

Americium-241/243 M

Cesium-137 F

Curium-244 M

Iodine-131/133 F

Plutonium-239 M or S

Radium-226 M

Strontium-85/90 F or S

Strontium-90/Yttrium-90 F or S

Thorium-228/232 F or S

Tritium F

Uranium-238 F, M, or S

Radionuclides of concern

Solubility 

type

Other radionuclides identified at NTS

Antimony-125 F or M

Cesium-134 F

Cerium-141/144 M

Colbalt-60 M or S

Europium-152/154/155 M

Iron-59 F

Manganese-54 F

Polonium-210 M

Niobium-95 M or S

Ruthenium-106 F or S

Zirconium-95 F or M
Effective Date: 09/30/2004 Revision No. 00 Document No. ORAUT-TKBS-0008-5 Page 74 of 82

Table 5D-25. Fission products up to 1 year old as 

identified in 1959 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

Rhodium-106

1.0 year

30 seconds

0.039

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

Barium-137m

33 years

2.6 months

0.51

-

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

Cerium-143 33 hour 1.39 – 0.71

Cerium-144

Praseodymium-144

282 day

17.5 minutes

0.3 – 0.17

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

Table 5D-26. 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
Effective Date: 

09/30/2004 Revision No. 00 Document No. ORAUT-TKBS-0008-5 Page 75 of 82

Table 5D-27. Radionuclide activity ratios at formation (immediately after detonation). 

Numerator H-3 Sr-89 Sr-90 Y-91 Mo-99 Ru-103 Ru-106 I-131 Cs-135 Cs-136 Cs-137

H-3 1

Sr-89 1 1.47E+2 7.95E-1 2.45E-1 3.86E-1 7.59 1.41E-1 7.87E+6 1.90E+1 1.01E+2

Sr-90 6.82E-3 1 5.42E-3 1.67E-4 2.64E-3 5.18E-2 9.64E-4 5.37E+4 1.30E-1 6.92E-1

Y-91 1.26 1.84E+2 1 3.08E-2 4.85E-1 9.54 1.78E-1 9.90E+6 2.39E+1 1.27E+1

Mo-99 4.07E+1 5.97E+3 3.25E+1 1 1.57E+1 3.09E+2 5.75 3.21E+8 7.75E+2 4.14E+3

Ru-103 2.59 3.80E+2 2.06 6.33E-2 1 1.97E+1 3.66E+1 2.04E+7 4.92E+1 2.63E+2

Ru-106 1.32E-1 1.93E+1 1.05E-1 3.23E-3 5.09E-2 1 1.86E-2 1.04E+6 2.51 1.34E+1

I-131 7.08 1.04E+3 5.63 1.73E-1 2.73 5.37E+1 1 5.57E+7 1.35E+2 7.18E+2

Cs-135 1.27E-7 1.86E-5 1.01E-7 3.12E-9 4.90E-8 9.64E-7 1.79E-8 1 2.41E-6 1.29E-5

Cs-1.6 5.26E-2 7.71 4.18E-2 1.29E-3 2.03E-2 6.10E-2 7.43E-3 4.14E+5 1 5.34

Cs-137 9.86E-3 1.45 7.85E-3 2.42E-4 3.81E-3 7.48E-2 1.39E-3 7.76E+4 1.87E-1 1
Effective Date: 09/30/2004 Revision No. 00 Document No. ORAUT-TKBS-0008-5 Page 76 of 82

Table 5D-28. Specific activity of selected 

alpha emitters. 

Radionuclide Specific activity (mCi/mg)

Th-230 1.94E-2

U-232 2.14E+1 

U-233 9.47E-3 

U-234 6.18E-3

U-235 2.20E-6

U-236 6.34E-5

U-238 3.33E-7

Pu-236 5.31E+2

Pu-238 1.74E+1

Pu-239 6.13E-2

Pu-240 2.26E-1

Pu-242 3.90E-3

Am-241 3.24

Am-243 1.85E-1
Effective Date: 09/30/2004 Revision No. 00 Document No. ORAUT-TKBS-0008-5 Page 77 of 82

REFERENCES

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

Albuquerque, New Mexico. 

AEC (U.S. Atomic Energy Commission), 1961a, “Employees, NTS, Mercury, Nevada – Personnel 

Exposures,” memorandum, Washington, D.C. 

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

Radiological Safety, Chapter 0524, NTSO-0524-01, Washington, D.C.

Arent, L., and C. Smith, 2003, “Personal Communications with Martha DeMarre, U.S. Department of 

Energy, Nevada Operations Office,” memorandum-to-file, NIOSH Dose Reconstruction Team.

Arent, L., and C. Smith, 2004, “Personal Communications with Martha DeMarre, U.S. Department of 

Energy, Nevada Operations Office,” memorandum-to-file, NIOSH Dose Reconstruction Team.

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.

BN (Bechtel Nevada, Inc.), 2000, Technical Basis for Internal Dosimetry at the Nevada Test Site, DEAC08-96NV11718, Mercury, Nevada. 

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

TBD-E211-002, Mercury, Nevada. 

DNA (Defense Nuclear Agency), 1979, Compilation of Local Fallout Data from Test Detonations 1945-

1962 Extracted from DASA 1251 Volume 1 Continental U.S. Tests, DNA 1251-1 EX, U.S. 

Department of Defense, Washington, D.C. 

DNA (Defense Nuclear Agency), 1981, Plumbbob Series 1957, DNA 6005F, U.S. Department of 

Defense, Washington, D.C. 

DNA (Defense Nuclear Agency), 1982, Operation Hardtack II 1958, DNA 6026FDNA 6026F, U.S. 

Department of Defense, Washington, D.C. 

DeMarre, M., 2003, E-mails to L. Arent (ORAU Dose Reconstruction Team), Tables of Codes, 

including “analyze-codes.xls,” “body-part-codes.xls,” “internal-nuclide-codes.xls,” and 

“sample-type-codes.xls,” U.S. Department of Energy, Nevada Operations Office, Las Vegas, 

Nevada, September 17. 

DeMarre, M., 2004, E-mail to L. Arent (ORAU Dose Reconstruction Team), Measurements of Alpha, 

Beta, Gamma, and Gross Fission Products at NTS (June 11) and NTS Code Questions 

(August 9), U.S. Department of Energy, Nevada Operations Office, Las Vegas, Nevada.

DOE (U.S. Department of Energy), 1962, “Procedures to Limit Radiation Exposure of Miners at NTS,” 

Nevada Operations Office Memorandum, Mercury, Nevada. 

DOE (U.S. Department of Energy), 1970, Test Manager’s Rad Safe Advisor- Status Report, 

Baneberry, Nevada Operations Office, Mercury, Nevada. 
Effective Date: 09/30/2004 Revision No. 00 Document No. ORAUT-TKBS-0008-5 Page 78 of 82

DOE (U.S. Department of Energy), 2000, United States Nuclear Tests, July 1945 through September 

1992, DOE/NV-209, Las Vegas, Nevada, downloaded from 

http://www.nv.doe.gov/news&pubs/publications/historyreports/pdfs/DOENV209_REV15.pdf. 

French, C. S., and K. W. Skrable, 1995, Evaluation of Protection, Bioassay, and Dose Assessment 

Programs for Internal Radiation Exposures at the NTS Particularly as Related to Three 

Exposure Situations. 

Geiger, E. L. and E. L. Whittaker, 1961, Analytical Procedures of the Radiological Safety Laboratory, 

REECo, Mercury, Nevada.

ICRP (International Commission on Radiation Protection), 1960, in Health Physics 3, Report of ICRP 

Committee II on Permissible Dose for Internal Radiation (1959). 

LASL (Los Alamos Scientific Laboratory), 1951, Operation Ranger, Volume 4, Program Reports - 

Gross Weapons Measurements, WT-201, University of California, Los Alamos, New Mexico. 

LASL (Los Alamos Scientific Laboratory), 1954, Analytical Procedures of the Industrial Hygiene 

Group, LA-1858, University of California, Los Alamos, New Mexico.

LASL (Los Alamos Scientific Laboratory), 1958a, Los Alamos Handbook of Radiation Monitoring, LA1835, University of California, Los Alamos, New Mexico. 

LASL (Los Alamos Scientific Laboratory), 1958b, Analytical Procedures of the Industrial Hygiene 

Group, LA-1858 (2nd. Ed.), University of California, Los Alamos, New Mexico.

LRL (Livermore Radiation Laboratory), 1957a “Saturn Tunnel Re-Entry,” memorandum, University of 

California, Livermore, California. 

LRL (Livermore Radiation Laboratory), 1957b, “Entry of Saturn Tunnel, August, 30 1957,” 

memorandum, University of California, Livermore, California. 

LRL (Livermore Radiation Laboratory), 1961, General Re-Entry Procedure for Underground Nuclear 

Events, CN-294, University of California, Livermore, California. 

LRL (Livermore Radiation Laboratory), 1965, Palanquin Reentry & Recovery Safety Procedure, 

University of California, Livermore, California. 

NCI (National Cancer Institute), 1997, Estimated Exposures and Thyroid Doses Received by the 

American People from Iodine-131 in Fallout Following Nevada Atmospheric Nuclear Bomb 

Tests, Publication No. 97-4264, Washington, D.C.

OTA (Office of Technology Assessment, U.S. Congress), 1989, The Containment of Underground 

Nuclear Explosions, OTA-ISC-414, U.S. Government Printing Office, Washington, D.C. 

REECo (Reynolds Electrical & Engineering Company, Inc.), 1957, Operation Plumbbob On-Site RadSafety Report, Mercury, Nevada. 

REECo (Reynolds Electrical & Engineering Company, Inc.), 1958, “Decontamination of Tunnel U12f,” 

memorandum, Mercury, Nevada. 
Effective Date: 09/30/2004 Revision No. 00 Document No. ORAUT-TKBS-0008-5 Page 79 of 82

REECo (Reynolds Electrical & Engineering Company, Inc.), 1960, Radiological Safety for 

Underground Nuclear Explosions, RRS-60-3, Mercury, Nevada. 

REECo (Reynolds Electrical & Engineering Company, Inc.), 1961, Radiological Safety Division 

Standard Operating Procedure, Mercury, Nevada.

REECo (Reynolds Electrical & Engineering Company, Inc.), 1964, Operation Storax On-Site 

Radiological Safety Report, NVO-162-14, Mercury, Nevada. 

REECo (Reynolds Electrical & Engineering Company, Inc.), 1965a, Operation Whetstone, Onsite 

Radiological Safety Report, July 1964 through June 1965, NVO-410-24, Mercury, Nevada. 

REECo (Reynolds Electrical & Engineering Company, Inc.), 1965b, Historical Radiation Records, 

Operation Whetstone (July 1964 – June 1965), #20 Merlin Event U3CT (02/16/65), Mercury, 

Nevada. 

REECo (Reynolds Electrical & Engineering Company, Inc.), 1973, Onsite Environmental Sciences 

Activities During the Baneberry Event, NVO-410-29, Mercury, Nevada. 

REECo (Reynolds Electrical & Engineering Company, Inc.), 1977-1987, Environmental Sciences 

Standard Procedure, Appendix A, Spectrometer Calibrations and Detection Limits, Mercury, 

Nevada.

REECo (Reynolds Electrical & Engineering Company, Inc.), 1993a, Technical Basis for Internal 

Dosimetry at the Nevada Test Site, DOE/NV/10630-64, Mercury, Nevada. 

REECo (Reynolds Electrical & Engineering Company, Inc.), 1993b, list by year of instruments, 

dosimetry (mostly external), rad levels, exposure criteria, respiratory protection, internal 

contamination, and calibration information; from the OCAS O Drive, file “NTS General 

Information 1951 – 1973.” 

Schoengold, C. R., M. E. DeMarre, and E. M. Kirkwood, 1996, Radiological Effluents Released from 

U.S. Continental Tests 1961-1992, DOE/NV-317, Bechtel Nevada, Las Vegas, Nevada.

Wilcox, F. W., and J.,S. Coogan, 1959, Atomic Weapon Accidents – Alpha Emitter Dispersal and 

Decontamination, REECo, Mercury, Nevada.
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GLOSSARY

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. 

alert level (circa 1980s)

The result that calls the attention of the observer to the fact that further action 

pertaining 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. 

Alert levels for specific analyses follow:

gamma

When the gamma count rate of a sample reaches 5,000 cpm, the sample is transferred 

to a gamma spectrometer for further analysis. Otherwise, a 137Cs– equivalent activity 

concentration (μCi/cc) is reported. 

Each positive radioiodine result is 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 indicates that thyroid counting of the individual 

is to be considered.

The alert level for routine quarterly urine samples is “any detectable” for 91Y, 103Ru, 106Ru106Rh, 141Ce, 

144Ce144Pr, and 154Eu.

Alert levels for all other gamma emitters are activity concentrations indicative of body burdens that are 

10% of the maximum permissible body burdens (MPBB) 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 is 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 is 50 mrem.

GFP - alert level was 10-7 μCi/cc.

Pu (or other transuranic elements) and Sr – any detectable level is the alert level.

controllable area

An area in which remedial actions are feasible. Criteria for a controllable area as 

defined by DOE are “... areas where trained rad-safe monitors are available, where 

communications are effective (where the exposure of each individual can be 

documented), where people can be expected to comply with recommended remedial 

actions, and where remedial actions against uptake of radionuclides in the food chain 

are practicable.” This equates to a zone of approximately 125 miles from the test 

control point. 
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controlled tunnel purging

An intentional release to allow recovery of experimental data and equipment or reuse 

of part of the tunnel system. 

late-time seeps

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. 

lower limits of detectability (LLD, circa 1980s) 

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

Minimum Detectable Activity (or Amount) (MDA)

is the estimated standard error for the net sample 

activity (before 1993). REECo 1993a states the LLD is a value selected above the 

MDA to reduce the probability of reporting false positive results.

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

sample (BN 2003).

operational release

Small consequential releases that occur when core or gas samples are collected or 

when the drill-back hole is sealed. 

Radex Areas

In a 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. In a 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 over the entire period of weapons testing, the first 

entry into an area or tunnel immediately (as soon as safety limitations will permit) 

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 reporting level for the period 1993-2003 is defined as the minimum level of a 

bioassay measurement result, which requires the measurement lab to provide prompt 

notification to Dosimetry (REECo 1993a). 

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. 
Effective Date: 09/30/2004 Revision No. 00 Document No. ORAUT-TKBS-0008-5 Page 82 of 82

tolerance level

The term for maximum permissible exposure before MPD was coined.

venting

Prompt, massive, uncontrolled releases of radioactive material. Ventings are 

characterized as active releases under pressure, such as when radioactive material is 

driven out of the ground by steam or gas.