ORAU Team

NIOSH Dose Reconstruction Project

Technical Basis Document for the Rocky Flats Plant

– Occupational Internal Dose

Document Number:

ORAUT-TKBS-0011-5 

Effective Date: 12/13/2005

Revision No.: 00 PC-1 

Controlled Copy No.: ______

Page 1 of 103

Subject Expert: Roger B. Falk

Document Owner

Approval: _Signature on File_____________Date: 01/16/2004

Robert Meyer, TBD Team Leader

Approval: _Signature on File_____________Date: 01/16/2004 Judson L. Kenoyer, Task 3 Manager

Concurrence: _Signature on File__________Date: 01/16/2004

Richard E. Toohey, Project Director

Approval: _Signature on File_____________Date: 01/16/2004

James W. Neton, OCAS Health Science Administrator

Supersedes:

None

TABLE OF CONTENTS

Section

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

Page

5

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

5.1 Introduction................................................................................................................................. 7

5.2 Source Term...............................................................................................................................8

5.2.1 Plutonium ...........................................................................................................................8

5.2.1.1 Isotopic Composition ................................................................................................. 8

5.2.1.2 Pu Solubility and Particle Size ...................................................................................9

5.2.2 Americium ........................................................................................................................10

5.2.2.1 Isotopic Composition ............................................................................................... 10

5.2.2.2 Am Solubility and Particle Size ................................................................................ 10

5.2.3 Enriched Uranium............................................................................................................. 10

5.2.3.1 Isotopic Composition ............................................................................................... 10

5.2.4 Depleted Uranium ............................................................................................................ 11

5.2.4.1 Isotopic Composition ............................................................................................... 11

5.3 Bioassay Data...........................................................................................................................11

5.3.1 Urinalysis Data .................................................................................................................11

5.3.1.1 Plutonium Urinalysis ................................................................................................ 11

5.3.1.1.1 Methods, Units, Isotopes, and Interferences..................................................... 11

5.3.1.1.2 Plutonium Reporting Levels, Minimum Detectable Activities, 

and Uncertainties ............................................................................................. 12

5.3.1.2 Americium Urinalysis ............................................................................................... 14

5.3.1.2.1 Methods, Units, Isotopes, and Interferences..................................................... 14
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5.3.1.2.2 Americium Reporting Levels, Minimum Detectable 

Activities, and Uncertainties.............................................................................. 14

5.3.1.3 Enriched Uranium Urinalysis.................................................................................... 15

5.3.1.3.1 Methods, Units, Isotopes, and Interferences..................................................... 15

5.3.1.3.2 EU Reporting Levels, Minimum Detectable Activities, and 

Uncertainties .................................................................................................... 15

5.3.1.4 Depleted Uranium Urinalysis ................................................................................... 16

5.3.1.4.1 Methods, Units, Isotopes, and Interferences..................................................... 16

5.3.1.4.2 DU Reporting Levels, Minimum Detectable Activities, and 

Uncertainties .................................................................................................... 16

5.3.1.5 Gross Alpha Urinalysis ............................................................................................ 17

5.3.1.5.1 Methods, Units, Isotopes, and Interferences..................................................... 17

5.3.1.5.2 Reporting Levels, Minimum Detectable Activities, and 

Uncertainties .................................................................................................... 17

5.3.1.6 Tritium .....................................................................................................................18

5.3.2 Lung Count Data .............................................................................................................. 18

5.3.2.1 Americium/Plutonium............................................................................................... 19

5.3.2.1.1 Methods, Units, Isotopes, and Interferences..................................................... 19

5.3.2.1.2 Reporting Levels, Minimum Detectable Activities, and 

Uncertainties .................................................................................................... 19

5.3.2.2 Thorium/Depleted Uranium...................................................................................... 21

5.3.2.2.1 Methods, Units, Isotopes, and Interferences..................................................... 21

5.3.2.2.2 Reporting Levels, Minimum Detectable Activities, and 

Uncertainties .................................................................................................... 22

5.3.3 Other Bioassay Data ........................................................................................................ 23

5.3.3.1 Wound Count Data .................................................................................................. 23

5.3.3.2 Nasal Smears and Fecal Samples........................................................................... 23

5.4 Records and Reports ................................................................................................................24

5.4.1 Urinalysis Records and Reports ....................................................................................... 24

5.4.2 Lung Count Records and Reports .................................................................................... 27

5.5 Internal Dosimetry for Unmonitored Workers ............................................................................ 30

5.5.1 Internal Dosimetry for Workers Not Involved in an Incident .............................................. 31

5.5.2 Internal Dosimetry for Workers Involved in an Incident..................................................... 31

References ........................................................................................................................................32

Glossary ............................................................................................................................................34

Attachment A Minimum Detectable Activity for Urinalysis Methods at Rocky Flats ........................... 36

A.1 Introduction...............................................................................................................................37

A.2 MDA Methodology .................................................................................................................... 37

A.3 History of Methods .................................................................................................................... 38

Reference..........................................................................................................................................54
Effective Date: 01/12/2004 Revision No. 00 Document No. ORAUT-TKBS-0011-5 Page 3 of 103

Attachment B Minimum Detectable Activity for In Vivo Lung Counts at Rocky Flats.......................... 55

B.1 Introduction...............................................................................................................................56

B.2 MDA Methodology .................................................................................................................... 56

B.3 History of Counting Systems and Procedures ........................................................................... 57

B.4 Assessment of MDA .................................................................................................................60

B.5 Values of the Variables, 1969 → for NaI(Tl) and Phoswich Detector Systems .......................... 62

B.6 Values of the Variables, 1976 → for Ortec Germanium Detector Systems................................ 63

B.7 Values of the Variables, 1978 → for PGT I Germanium Detector Systems ............................... 64

B.8 Values of the Variables, 1979 → for PGT II Germanium Detector Systems .............................. 65

B.9 Values of the Variables, 1985 → for PGT Organ Pipe Germanium Detector Systems............... 66

B.10 Values of the Variables, 1985 → for EG&G Organ Pipe Germanium Detector Systems............ 66

B.11 MDA for Rocky Flats Plutonium ................................................................................................ 67

References ........................................................................................................................................69

Attachment C Examples of Records and Reports Used at Rocky Flats............................................. 72

(See List of Figures)

LIST OF TABLES

Table

5.2.1.1-1 Weight percent and fraction of alpha activity for weapons-grade

plutonium......................................................................................................................

Page

9

5.2.1.1-2 Weight percent and fraction of alpha activity for ZPPR plutonium................................. 9

5.2.3.1-1 Weight percent and fraction of alpha activity for enriched 

uranium ......................................................................................................................10

5.2.4.1-1 Weight percent and fraction of alpha activity for depleted 

uranium ......................................................................................................................11

5.3.1.1.2-1 Median and extreme MDA values for plutonium ......................................................... 13

5.3.1.2.2-1 Median and extreme MDA values for americium......................................................... 14

5.3.1.3.2 Median and extreme MDAs for enriched uranium....................................................... 15

5.3.1.4.2-1 Median and extreme MDAs for depleted uranium from 1952 to 

April 1964 ................................................................................................................... 17

5.3.1.4.2-2 Median and extreme MDAs for depleted uranium from May 1964 

to the present ............................................................................................................. 17

5.3.1.5.2-2 Median and extreme MDAs for gross alpha measurements........................................ 18

5.3.2.1.2-1 Summary of MDAs for 241Am ...................................................................................... 20
Effective Date: 01/12/2004 Revision No. 00 Document No. ORAUT-TKBS-0011-5 Page 4 of 103

LIST OF FIGURES

Figure

C-1 Urinalysis Record Card and Health Sciences Data System – Urinalysis Detail (1) 

(first activity date on the HSDS portion: 10-29-54) ................................................................

Page

73

C-2 Urinalysis Record Card and Health Sciences Data System – Urinalysis Detail (2) 

(first activity date on the HSDS portion 8-19-53) .................................................................... 74

C-3 Urinalysis Record Card and Health Sciences Data System – Urinalysis Detail (3) 

(first activity date on the HSDS portion: 1-6-58) .................................................................... 75

C-4 Health Sciences Data System – Urinalysis Detail (1) (first activity date 9-17-58).................... 76

C-5 Health Sciences Data System – Urinalysis Detail (2) (first activity date 3-19-73).................... 77

C-6 Analytical Report – Bioassay Analysis Data 3-15-93.............................................................. 78

C-7 Analytical Report – Bioassay Analysis Data 10-28-93 ............................................................ 79

C-8 Form 1 – Sample Results 1-29-96.......................................................................................... 80

C-9 Rocky Flats Environmental Technology Site (1) 8-27-96 (analytes: U238, U235, 

U234).....................................................................................................................................81

C-10 Rocky Flats Environmental Technology Site (1) 8-8-96 (analyte: Pu239).............................. 82

C-11 Form 1 – Sample Results – Quanterra, Richland 7-31-98 ...................................................... 83

C-12 General Engineering Laboratories, Inc. 6-28-99..................................................................... 84

C-13 Health Sciences Urinalysis Record (with tritium, fecal and nasal smear results) .................... 85

C-14 Health Physics – Body Counter Information 12-8-65.............................................................. 86

C-15 Health Physics – Body Counter Information 5-16-68.............................................................. 87

C-16 Health Physics – Body Counter Information 8-26-68.............................................................. 88

C-17 Health Physics – Body Counter Information 9-16-70.............................................................. 89

C-18 Radiation Dosimetry – Body Count Results 10-3-74............................................................... 90

C-19 Radiation Dosimetry – Body Count Results 5-30-75............................................................... 91

C-20 Radiation Dosimetry – Body Count Results 1-9-78................................................................. 92

C-21 Body Counter Results 12-8-81 ............................................................................................... 93

C-22 Radiation Dose Assessment – Body Count Results 7-22-83.................................................. 94

C-23 Radiation Dose Assessment – Body Count Results 5-18-83.................................................. 95

C-24 Radiation Dose Assessment – Body Count Results 2-21-84.................................................. 96

C-25 Radiation Dose Assessment – Body Count Results 3-22-84.................................................. 97

C-26 Radiation Dose Assessment – Body Count Results 10-10-85 ................................................ 98

C-27 Radiation Dose Assessment – Body Count Results 3-6-89.................................................... 99

C-28 Internal Dosimetry – Lung Count Results 11-23-93.............................................................. 100

C-29 ABACOS-Plus 3-6-96........................................................................................................... 101

C-30 ABACOS-Plus 11-15-01....................................................................................................... 102

C-31 ABACOS-Plus 6-14-01......................................................................................................... 103
Effective Date: 12/13/2005 Revision No. 00 PC-1 Document No. ORAUT-TKBS-0011-5 Page 5 of 103

RECORD OF ISSUE/REVISIONS

ISSUE 

AUTHORIZATION 

DATE

EFFECTIVE 

DATE REV. NO. DESCRIPTION

01/12/2004 01/12/2004 00 New document to establish technical basis 

document for the Rocky Flats Plant - 

occupational internal dose. First approved issue. 

Initiated by Robert Meyer.

01/12/2004 12/13/2005 00 PC-1 Page change initiated to incorporate recent 

direction from NIOSH to include DOL review 

comments on page 7 in Section 5.1. Change 

made to Table 5.3.1.4.2-2 on page 17 in Section 

5.3. No sections were deleted. Retraining is 

not required. Initiated by Robert Meyer.

Approval:

Robert Meyer, TBD Team Leader

Signature on File________________11/22/2005 

Judson L. Kenoyer, Task 3 Manager

Signature on File ____________ 11/22/2005 

Kate Kimpan, Project Director

Signature on File___________ 12/07/2005 

James W. Neton, Associate Director for Science

Signature on File_____________ 12/13/2005 
Effective Date: 01/12/2004 Revision No. 00 Document No. ORAUT-TKBS-0011-5 Page 6 of 103

ACRONYMS AND ABBREVIATIONS

cm centimeter

cm2 square centimeter

cpm counts per minute

CWT chest wall thickness

dpm disintegrations per minute

DTPA diethylenetriaminepentaacetate

EDTA ethylenediaminetetraacetate

hr hour

keV kilovolts-electron, 1,000 electron volts

L X-rays low-energy x-rays produced during radioactive decay

LLL Lawrence Livermore laboratory

ml milliliter

MLT minutes live time

MPLB maximum permissible lung burden

nCi nanocurie

pCi picocurie

PHA pulse height analysis

ppm parts per million

RCG radioactive concentration guides

RF Rocky Flats

ROI region of interest

TBP tributyl phosphate

TOPO trioxyl phosphene oxide

TTA thenoyl trifluro acetone

μCi microcurie

μg microgram

ZPPR zero-power physics reactor
Effective Date: 11/22/2005 Revision No. 00 PC-1 Document No. ORAUT-TKBS-0011-5 Page 7 of 103

5.1 INTRODUCTION

Technical basis documents and site profile documents are general working documents that provide 

guidance concerning 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 National Institute for Occupational Safety and Health 

(NIOSH) in the completion of the individual work required for each dose reconstruction.

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

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

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

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

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

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

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

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

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

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

facility encompassed by EEOICPA.

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

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

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

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

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

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

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

determined in accordance with the [probability of causation] 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.

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. §Section

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 radiation exposures 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 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.

This section discusses Rocky Flats internal dosimetry data for the dose reconstructor, including 

guidance for the appropriate use of that information. 

Workers at Rocky Flats had the potential to receive intakes of plutonium, americium, enriched 

uranium, depleted uranium, and tritium, as well as miscellaneous other radionuclides. Section 5.2 

describes the available source term information, including isotopic composition, solubility, and particle 
Effective Date: 01/12/2004 Revision No. 00 Document No. ORAUT-TKBS-0011-5 Page 8 of 103

size. Site-specific internal dosimetry information for other radionuclides, such as thorium, curium and 

neptunium, is rare or not available. 

The primary modes of intake would have been chronic or acute inhalation or through breaks in the 

skin (wounds). The primary bioassay data are the urine data, the activity of the radionuclide of 

interest that is excreted in the urine following an inhalation or wound intake, and the lung count data, 

the activity of the radionuclide present in the lungs after an inhalation intake. Section 5.3 discusses 

these two data sets in detail, including the history, sensitivity, and pertinent nuances of the methods 

and data. 

The internal exposure record for a worker consists of records of the bioassay data and reports of 

involvement in incidents, accidents, or special situations. Section 5.4 describes samples of these 

records and reports with explanations of the aspects important to dosimetry.

Workers could have received intakes not monitored by bioassay, or the records could be missing or 

incomplete. Section 5.5 discusses internal dosimetry for those instances.

5.2 SOURCE TERM

5.2.1

5.2.1.1 Isotopic Composition

Plutonium

Three aspects of the isotopic composition of plutonium are important to internal dose reconstruction:

1. The percent by weight of 241Pu, to calculate the ingrowth of 241Am for the lung count data

2. The fraction of the activity for each alpha-emitting plutonium isotope, to account for the dose 

contributed by unmeasured isotopes

3. The ratio of the activity of 241Pu to the alpha activity of the other plutonium isotopes, to 

calculate the intake of 241Pu from intakes from bioassay data for 239Pu and 240Pu.

For weapons-grade plutonium, which was present at Rocky Flats throughout most of its 1952 to 1989 

production history, the ratio of the activity of 241Pu to the alpha activity of the other plutonium isotopes 

is 5.1, and the 240Pu content is about 6% by weight. Table 5.2.1.1-1 shows the weight percent and 

fraction of alpha activity for each isotope. 

The ZPPR special project in the mid-1960s involved reactor-grade plutonium. The ratio of the activity 

of 241Pu to the alpha activity of the other plutonium isotopes is 32. Table 5.2.1.1-2 lists the weight 

percent and alpha activity fraction for each isotope. Reports of accidents or incidents involving ZPPR 

plutonium generally note “ZPPR’ or “ZPPR material”, especially on the lung count reports.

The dose reconstruction should account for the activity of 241Am in the plutonium mixture. The 

concentration of the 241Am is variable depending on the time since the plutonium was purified and 

whether the mixture involves waste or by-product (separated 241Am) from the purification of aged 

plutonium. Starting in 1969, ppm 241Am was measured for the plutonium mixture involved in 

significant possible inhalation incidents and is generally recorded on lung count reports for workers 

involved in those incidents. A nominal amount, 100 or 1000 parts per million (ppm) by mass, of 241Am 

should be assumed with the plutonium if no other data are available. If the plutonium intake for 

weapons grade plutonium is assessed for 239,240Pu , the activity of 241Am in the intake mixture is 
Effective Date: 01/12/2004 Revision No. 00 Document No. ORAUT-TKBS-0011-5 Page 9 of 103

calculated by multiplying the 239,240Pu activity by [48.2 x ppm 241Am/ (1x106 - ppm 241Am)]. For ZPPR 

plutonium, the 239,240Pu activity is multiplied by [44.6 x ppm 241Am/ (1x106

 - ppm 241Am)] to obtain the 

activity of 241Am in the intake mixture. 

Table 5.2.1.1-1. Weight percent and fraction of alpha 

activity for weapons-grade plutonium.

a

 

Isotope Percent by weight

Fraction of 

alpha activity

238Pu 0.01 0.023

239Pu 93.79 0.80

240Pu 5.80 0.18

241Pu 0.36 b – 242Pu 0.03 Negligible

a. Source: Final Environmental Impact Statement, Rocky Flats 

Plant Site (DOE 1980, Volume 1, Table 2.7.2-2, p. 2-170). 

Values are the average for Rocky Flats plutonium from July 1976 

to July 1, 1978. This isotopic composition is also typical of 

plutonium metal processed at Rocky Flats through 1990 (James 

1990).

b. The percent by weight of 241Pu for the period 1959 through 1977 

was 0.49, with a range of 0.35 to 0.65 (RFETS 2002, p. 5.1).

Table 5.2.1.1-2. Weight percent and fraction of alpha 

activity for ZPPR plutonium.

a

 

Isotope Percent by weight

Fraction of 

alpha activity

239Pu 87.6 0.70

240Pu 10.0 0.30

241Pu 2.4 –

a. These ZPPR values are based on extracted data in a working file, 

from an undocumented source.

5.2.1.2 Pu Solubility and Particle Size

Most plutonium in metal-working operations and involved in fires is insoluble (i.e., type S). 

Exceptions, such as plutonium metal associated with solvents such as carbon tetrachloride, may be 

assumed to be more soluble (type M) if this is what the data show or it is more claimant favorable. 

The plutonium fire on October 15, 1965, in Buildings 776 and 777 is a special case. The plutonium 

was avidly retained in the lungs of exposed workers with relatively low transfer to the urine, exhibiting

highly insoluble (super type S) characteristics. 

Plutonium in chemical-processing operations can be either soluble (type M), insoluble (type S), or a 

mixture of solubilities. A claimant favorable approach is to assume insoluble plutonium if the 

qualifying cancer is of the respiratory system and to assume soluble plutonium for all other cases. 

Lung count data in conjunction with urine data may help to determine absorption type. 

In general, particle size and distributions are not available for work areas or incidents at Rocky Flats. 

Therefore, dose reconstructions should use the default value of 5 μm.

One exception is the plutonium fire on October 15, 1965, in Buildings 776 and 777, for which Mann 

and Kirchner (1967) measured a mass median diameter of 0.3 μm with a geometric deviation of 1.83. 

A claimant favorable approach is to assume 0.3 μm for all plutonium fires unless the qualifying cancer 

involves the tissues of the extrathoracic regions. 
Effective Date: 01/12/2004 Revision No. 00 Document No. ORAUT-TKBS-0011-5 Page 10 of 103

5.2.2

5.2.2.1 Isotopic Composition

Americium

For the NIOSH dose reconstruction project, the assumption is that measured americium is 100% 

241Am. 

5.2.2.2 Am Solubility and Particle Size

Americium was present in two forms at Rocky Flats, as a purified by-product of plutonium recovery 

and as atoms formed by the nuclear transformation of 241Pu and imbedded in the matrix of the 

plutonium particle. As a purified by-product, ICRP 68 specifies americium inhalation absorption as 

type M. For imbedded atoms in the matrix of an inhaled plutonium particle, the dose reconstruction 

should use the solubility classification described for the plutonium particle in Section 5.2.1.2.

The dose reconstruction should assume a 5 um AMAD, except for fires a X.X um AMAD should be 

assumed for consistency with Section 5.2.1.3 above. 

5.2.3

5.2.3.1 Isotopic Composition

Enriched Uranium

Production at Rocky Flats involved enriched uranium from 1952 to 1963. Internal dose 

reconstructions would comprise mainly the percent of the alpha activity for 234U, but should also 

account for the dose contributed by unmeasured isotopes. Table 5.2.3.1-1 lists the weight percent 

and fraction of alpha activity for each isotope.

Table 5.2.3.1-1. Weight percent and fraction of alpha 

activity for enriched uranium.

a

 

Isotope Percent by weight

Fraction of 

alpha activity

234U 1.0 0.97

235U 93. 0.031

236U 0.39 0.0039

238U 5.4 0.00028

a. Source: DOE (1980, Volume 1, Table 2.7.2-4, p. 2-172). 

5.2.3.2 EU Solubility and Particle Size

Operations for enriched uranium paralleled those for plutonium, including chemical processing and 

metalworking. Compounds of uranium are generally more soluble than those of plutonium, and 

solubility classification is uncertain. The International Commission on Radiological Protection assigns 

compounds UO3 (yellow cake), UF4, and UCl4 to inhalation type M and compounds UO2 and U3O8 to 

inhalation type S (ICRP66 1994; ICRP30 1979).

In many cases, the compound of uranium involved in an intake is not identified. There are no Rocky 

Flats site-specific data for enriched uranium. In general, intakes in chemical processing areas 

occurred as type M or as mixtures of type M and type S. Reconstructions should use the most 

claimant favorable mixture. For intakes that occur in metal-working areas, the claimant favorable

assumption is type S if the qualifying cancer site is in the respiratory system. For other cancer sites, 

the more claimant favorable assumption is type M. Reconstructions should use the default value of 5 

μm. 
Effective Date: 01/12/2004 Revision No. 00 Document No. ORAUT-TKBS-0011-5 Page 11 of 103

5.2.4

5.2.4.1 Isotopic Composition

Depleted Uranium

Depleted uranium (DU) was present at Rocky Flats throughout its production history. U-238 will 

account for the majority of DU internal dose, but the total uranium alpha activity should be included in 

the dose reconstruction.

Table 5.2.4.1-1. Weight percent and fraction of alpha 

activity for depleted uranium.

a

Isotope Percent by weight

Fraction of 

alpha activity

234U 0.00058 0.097

235U 0.23 0.013

238U 99.77 0.890

a. These values are derived from data in Table 2.7.2-4, page 2-172 

in DOE (1980, Volume 1). 

5.2.4.2 DU Solubility and Particle Size

Operations with depleted uranium involved metal working, including casting, forming, and melting. 

Likely compounds are UO3 and U3O8. The solubility classification is ambiguous, falling somewhere 

between type S and type M (RFETS 1998a, Section 6.1; HPS 1995; Lawrence 1984). If the qualifying 

cancer site is in the respiratory system, use the claimant favorable assumption of type S. For other 

cancer sites, the more claimant favorable assumption is type M. 

Site-specific data for particle size of uranium is not available. Reconstructions should use the default 

value of 5 μm AMAD. 

5.3 BIOASSAY DATA

The primary data for intake assessment at Rocky Flats are the urinalysis data and the lung count 

data. Other bioassay data, such as wound count data, fecal sample data, and nasal smear data, 

were obtained in special situations but generally were not used to quantify intakes.

5.3.1

Attachment 5A, Minimum Detectable Activity for Urinalysis Methods at Rocky Flats, discusses the 

history of the methods, reporting and recording levels, and sensitivities of the methods as they 

evolved and were implemented at Rocky Flats. This section summarizes, supplements, and expands 

the information in Attachment 5A.

Urinalysis Data

5.3.1.1 Plutonium Urinalysis

5.3.1.1.1 Methods, Units, Isotopes, and Interferences

Through 1989, the units of the results are disintegrations per minute per a 24-hour excretion period 

(dpm/24-hr). After 1989, the units of the results are disintegrations per minute per sample 

(dpm/sample), regardless of the sample volume or excretion period. Assume a 24-hour excretion 

period unless the record indicates that the actual excretion period was different. 
Effective Date: 01/12/2004 Revision No. 00 Document No. ORAUT-TKBS-0011-5 Page 12 of 103

Through 1977, samples were counted using an air proportional detector system that did not have 

sufficient resolution to separate the alpha energies for the plutonium alpha-emitting isotopes. Starting 

in 1973, an alpha pulse height analysis (PHA) system with surface barrier detectors was phased in 

and had completely replaced the air proportional detector system by 1978. The plutonium urine 

results provided by the air proportional detector system include activity from 238Pu, 239Pu, and 240Pu. 

Plutonium urine results for samples counted by the PHA system include only 239Pu and 240Pu results. 

Intake assessments are simpler and more claimant favorable if the dose reconstruction assumes 

239Pu and 240Pu for all plutonium urine results unless the worker was involved in a special situation 

involving pure 238Pu. If the intake is assessed using 239Pu and 240Pu data, the 238Pu component of the 

intake is obtained by multiplying the 239Pu and 240Pu intake by 1.0264. 

Interferences were likely in the period 1952 through 1962 because of a lack of specificity of the 

chemical procedure to isolate only the plutonium in the extract. Plutonium results would likely include

some americium and thorium activity. In addition, for gross alpha analyses assigned to plutonium 

through 1973, the result could include some contribution from uranium. However, it is claimant 

favorable to disregard such interferences and take the plutonium results at face value unless a value 

can be determined to be an outlier.

From 1963 to 1977, the ion exchange method significantly reduced interferences from americium, 

uranium, and thorium. As the PHA system was phased in starting in 1973, the possibility of 

interferences was further reduced. After 1977, these interferences were not a significant issue for 

plutonium urine results because all samples were counted on the PHA system. 

Another source of interference was contamination of the tracer (

236Pu or 242Pu) by the analyte 

isotopes, 239Pu and 240Pu, an infrequent occurrence.

Chelation (EDTA or DTPA) treatments cause enhanced excretion of plutonium in the urine. Urine 

data within 90 days of a chelation injection have historically been excluded from calculations of

intakes or depositions of plutonium. Information in the medical or dosimetry records should allow the 

dose reconstructor to discern chelation treatments, which generally followed a significant and 

documented incident. In the urine data reports for the health sciences data system, urine data 

affected by chelation were flagged with a code 1. Code 1 was also used to flag urine data that did not 

pass quality standards. The dose reconstructor should be wary of any urine result flagged with a 

code 1, and in general should not use these data in dose reconstruction. 

5.3.1.1.2 Plutonium Reporting Levels, Minimum Detectable Activities, and Uncertainties

The minimum reporting level for plutonium through 1961 was 0.88 dpm/24-hr sample (this was 10% of 

the RF tolerance level). For 1962 through April 6, 1970, the minimum reporting level was 0.20 

dpm/24-hr sample. Results less than the reporting level were reported as zero or background (or 

some abbreviation, e.g., BK). For some workers, results initially reported as background were 

superceded by the report of the actual result, if the actual result was ≥0.00 dpm/24-hr sample. After 

April 6, 1970, all results ≥0.00 dpm/24-hr sample were reported. Negative results were reported as 

zero through 1989. After 1989, the actual negative value was reported. Starting in approximately 

1990, urine results were not normalized to a 24-hr sample. Instead, the results are dpm/sample, 

regardless of the sample volume.

The minimum detectable activity (MDA) for plutonium is presented here for the median conditions. By 

definition of the median value, half of the sample-specific MDAs are lower than the median value, and 

half are higher. In most cases the dose reconstructor is not likely to have sufficient data to determine 

the sample-specific MDA, so the median values should be used. 
Effective Date: 01/12/2004 Revision No. 00 Document No. ORAUT-TKBS-0011-5 Page 13 of 103

Table 5.3.1.1.2-1 lists the MDA values for plutonium. The values for 1952 through 1977 are based on 

examination of urinalysis data logs for the period 1952 – 1971 (see Attachment 5A). The MDA value 

for 1971 was extrapolated through 1977. The MDA value for 1978 through 1989 is based on matrix 

blank data for the routine plutonium urinalysis program for the period August 1, 1990, through 

September 27, 1991, using blank values with a sample-specific recovery in the range of 0.10 to 1.10. 

This range of recoveries mimics the range used from 1978 to 1989 for a valid analysis of routine 

samples. For 1990 to 1992, the blank values with a sample-specific recovery in the range from 0.35 

to 1.10 were used to determine the MDA value. For 1993 to the present, the value of the MDA is 

equal to the sample-specific MDA of 0.020 dpm/sample contractually required in the Rocky Flats 

Environmental Technology Site (RFETS) bioassay statement of work (RFETS 1998b) for any 

laboratory processing the sample, although the required MDA was not consistently achieved by the 

on-site laboratory. Note that, the value of the sample-specific MDA is included in the urinalysis data 

reports starting in 1990.

Table 5.3.1.1.2-1. Median MDA values for 

plutonium .a,b 

Period dpm/24-hr sample

1952 – 1953 0.57

1954 – 1962 0.51

1963 0.44

1964 – 1977 0.54

1978 – 1989 0.24

1990 – 1992 0.24

1993 – 0.020

a. Note: The unit of the MDA values starting in 1990 is 

dpm/sample.

b. Sample-specific MDA values, if found in the record 

starting 1990, should be used instead of the generic 

MDA values in this table. 

Some urine samples may have been processed by an off-site commercial laboratory before 1993. 

The reports for those samples may have the sample-specific MDAs. If these are not available, the 

MDA listed in Table 5.3.1.1.2-1 should be used. 

Some periods contain transitions that improved the detection of plutonium. For example, from 1964 to 

1977, electrodeposition of the plutonium replaced evaporation of the extract on the planchet. In 

addition, starting in 1973 with four detectors, plutonium samples were processed with an internal 

standard and were counted on a PHA system to establish the sample-specific recovery. The count 

time was also increased to 720 minutes. Because of the difficulty of determining which improvements 

apply to each sample, the MDAs in Table 5.3.1.1.2-1 do not account for the improvement until the 

transition was completed for all samples (claimant favorable). 

The uncertainty of the result was not quantified and reported in the record until approximately 1980. 

The reported value was the two-sigma standard error and included only uncertainties of counting 

statistics, adjusted by the sample-specific recovery. Starting in approximately 1986, contributions 

from other sources of uncertainty were included, and the reported value was the one-sigma standard 

error. To estimate the uncertainty for results without a reported uncertainty, a reasonable approach is 

to divide the median MDA value by 3.3, where 3.3 is the sum of kα and kβ, and kα = kβ = 1.645 (see 

Attachment 5A).
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5.3.1.2 Americium Urinalysis

5.3.1.2.1 Methods, Units, Isotopes, and Interferences

Attachment 5A describes the methods through 1971. After 1971, the method for americium paralleled 

that for plutonium. The isotope is 241Am.

The units of the results are disintegrations per minute per a 24-hr excretion period (dpm/24-hr sample) 

through 1989. After 1989, the units of the results are disintegrations per minute per sample 

(dpm/sample), regardless of the sample volume or excretion period. 

The main interference is thorium, specifically 228Th, which has two alphas with energies similar to 

those of 241Am and has chemical properties similar to americium. If the chemical extraction procedure 

for americium was not run precisely, thorium would be eluted from the ion exchange column with the 

americium. When the extract was counted, even for the PHA system, the 228Th could not be 

distinguished from the 241Am.

The plutonium to americium alpha activity ratio (239,240Pu dpm/24-hr sample divided by 241Am 

dpm/24-hr sample) for paired plutonium and americium urine results provides a credibility check. An 

alpha activity ratio less than 2.0 (corresponding to a parts per million (ppm) value for 241Am of 10,000 

or greater) is not credible unless the worker was involved with 1) separated 241Am (Line 1 in Building 

771, 2) the molten salt process in Building 776, 3) research and development projects involving pure 

americium), 4) with material from the ZPPR project, or 5) waste identified for those operations. 

The dose reconstructor should use the plutonium urine data instead of the 241Am urine data to assess 

intakes of weapons grade plutonium. The intake of the 241Am is then calculated from the value of the 

initial parts per million of 241Am measured or assumed for the plutonium mixture involved in the intake.

5.3.1.2.2 Americium Reporting Levels, Minimum Detectable Activities, and Uncertainties

The reporting levels for americium were ≥ 0.24 d/m/24 hours in 1963, ≥ 0.20 d/m/24 hours in 1964 

through 1967, and ≥ 0.30 d/m/24 hours in 1968 through 1971. Results less than the reporting level 

were reported as zero or background (or some abbreviation, e.g., BK). The reporting practice for the 

period 1972 through 1976 is to be determined. Until it is determined, reconstructions should assume 

that the reporting level for 1968 to 1971 was continued through 1976. Starting in 1977, all results 

≥0.00 dpm/24-hr sample were reported. Negative results were reported as zero through 1989. After 

1989, the actual negative value was reported. As for plutonium, urine results were not normalized to 

a 24-hr sample starting in about 1990. Instead, the results are dpm/sample, regardless of the sample 

volume.

The MDAs for americium (Table 5.3.1.2.2-1) were determined as described for plutonium (see Section 

5.5.1.1.2 and Attachment 5A), with the difference that the americium analyses started in 1963. 

The discussions of MDA and uncertainty for plutonium in Section 5.3.1.1.2 apply to americium.

Table 5.3.1.2.2-1. Median MDA values for 

americium.

a,b 

Period dpm/24-hr sample

1963 0.44

1964 – 1965 0.55

1965 – 1970 0.46

1971 – 1977 0.76
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1978 – 1989 0.31

1990 – 1992 0.30

1993 – 0.020

a. Note: The unit of the MDA values starting in 

1990 is dpm/sample.

b. Sample-specific MDA values, if found in the 

record starting 1990, should be used instead of 

the generic MDA values in this table. 

5.3.1.3 Enriched Uranium Urinalysis

5.3.1.3.1 Methods, Units, Isotopes, and Interferences

The units of the results are disintegrations per minute per a 24-hr excretion period (dpm/24-hr sample) 

for the entire period.

Since urine samples analyzed for enriched uranium were counted with the air proportional detectors, 

all of the alpha emitting isotopes of uranium are included in the result.

Site-specific information about possible interferences that may have occurred for the urinalysis 

methods for enriched uranium is not available. 

It is claimant favorable to assume that the result is all enriched uranium.

5.3.1.3.2 EU Reporting Levels, Minimum Detectable Activities, and Uncertainties

Table 5.3.1.3.2 lists the MDAs for enriched uranium. The reporting level for enriched uranium through 

1963 was ≥8.8 dpm/24-hr sample (10% of the RF tolerance level). From 1964 to 1971, the minimum 

reporting level ranged from 20 to 28 dpm/24-hr sample, depending on the volume of the sample, as 

observed from the urinalysis data logs for that period. Results less than the reporting level were 

reported as zero or background (or some abbreviation, e.g., BK). When urinalysis for enriched 

uranium was stopped at Rocky Flats is undetermined, although it likely occurred in the early 1970s.

Table 5.3.1.3.2. Median MDAs for enriched 

uranium. 

Period dpm/24-hr sample 

1952 – 1953 14

1954 – 1959 13

1960 – 1963 9.4

1964 – 1969 31

1970 – 1971 25

The MDAs for enriched uranium were determined as described for plutonium (see Section 5.5.1.1.2 

and Attachment 5A). 

Uncertainties for the enriched uranium urine results have not been quantified or reported. To estimate 

the uncertainty for results without a reported uncertainty, a reasonable approach is to divide the 

median MDA value by 3.3, where 3.3 is the sum of kα and kβ, and kα = kβ = 1.645(see Attachment 5A).
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5.3.1.4 Depleted Uranium Urinalysis

5.3.1.4.1 Methods, Units, Isotopes, and Interferences

Attachment 5A describes the uranium urinalysis methods through 1971. In 1972 through 1979, 

depleted uranium samples were chemically processed with the uranium-specific process, trioxyl 

phosphene oxide (TOPO) extraction procedure, and the extract (electrodeposited) was counted on 

the gas flow proportional counter. From 1980 to 1997, depleted uranium samples were processed 

with a tracer (232U or 236U) by ion exchange and alpha-counted with the alpha spectrometry system 

with surface barrier detectors in vacuum. The starting year of use of the tracer has not been 

determined. From 1997 to the present, depleted uranium samples were processed at an off-site, 

commercial laboratory according to provisions of the bioassay statement of work (RFETS 1998b).

The units for 1952 to April 1964 are micrograms uranium per 24-hr excretion period (μg/24-hr 

sample). The mass measurement is for all the isotopes of uranium. From May 1964 to 1989, the 

units are dpm/24-hr sample. After 1989, the units of the results are dpm/sample, regardless of the 

sample volume or excretion period. 

The urine data logs through 1971 do not identify the isotopes involved. However, it is reasonable to 

assume that all the uranium alpha-emitting isotopes were included in the air proportional detectors

measurement. For the 1980s, the logs have not been reviewed sufficiently to determine which 

isotopes were included in addition to 238U, which contributes 89% of the alpha activity. It is claimant 

favorable to assume that the reported urine result pertains only to 238U and to determine additional 

intakes for the other isotopes. In the 1990 reports, the urine data include the results separately for 

234U, 235U, and 238U.

The major interference is the contribution from natural uranium, which is ubiquitous, sometimes in 

concentrated pockets, in the terrain near the Rocky Flats site. No adjustments have been made to

the reported depleted uranium urine results for this background, which was highly variable. It is 

claimant favorable to assume that the depleted uranium urine results are from occupational intakes of 

depleted uranium unless isotopic data are available that indicate natural uranium (i.e., 234U activity is 

approximately equal to 238U activity).

5.3.1.4.2 DU Reporting Levels, Minimum Detectable Activities, and Uncertainties

The minimum reporting level for depleted uranium through April 1964 was 5.8 μg/24-hr sample (10% 

of the tolerance level). After April 1964 through 1971, the minimum reporting level was the same as 

for enriched uranium, 20 to 28 dpm/24-hr sample, depending on the volume of the sample. The 

reporting level for 1972 through 1979 (TOPO procedure) is to be determined. A claimant favorable

approach is to use the reporting level for 1964 through 1971. In the 1980s, all results ≥0.00 

dpm/24-hr sample were reported. Negative values were reported as 0.00 dpm/24-hr sample. In the 

1990s and after, all actual results, including negative values, were reported.

The MDAs for depleted uranium for fluorometric measurements were determined as described in 

Attachment 5A. Median MDAs for depleted uranium from 1952 to April 1964 are shown in Table 

5.3.1.4.2-1. For alpha-counting methods, the MDAs in the period 1964 through 1971 are the same as 

those for enriched uranium in Table 5.3.1.3.2. The MDAs for 1972 through 1979 are to be 

determined. The MDAs for 1980 to the present were derived in the same manner as described for 

plutonium but are based on 238U. 
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Table 5.3.1.4.2-1. Median MDAs for 

depleted uranium from 1952 to April 1964. 

Period μg/24-hr sample 

1952 – 1955 31

1955 – 1959 12

1960 – 1964 (April) 11

Table 5.3.1.4.2-2. Median MDAs for 

depleted uranium from May 1964 to the 

present.

a,b 

Period dpm/24-hr sample

1964 – 1969 31

1970 – 1971 25

1972 – 1979 25c

1980 – 1989 0.56

1990 – 1992 0.40

1993 – 0.10

a. The MDA value unit starting in 1990 is dpm/sample.

b. Sample-specific MDA values, if found for 1990+, should 

be used instead of generic MDA values. 

c. actual practice unknown; assume continuation of 

earlier practice.

The discussion of the uncertainty for plutonium in Section 5.3.1.1.2 applies to depleted uranium.

5.3.1.5 Gross Alpha Urinalysis

5.3.1.5.1 Methods, Units, Isotopes, and Interferences

Gross alpha measurement is a non-specific analysis used for workers who were potentially exposed 

to both uranium and plutonium in the same monitoring period. Workers who were potentially exposed 

to other alpha-emitting radionuclides, such as neptunium and curium, may also have been monitored 

for gross alpha. Urinalysis methods are discussed in Attachment 5A. The gross alpha method was 

discontinued in the early 1970s, likely 1973. The results are reported as dpm/24-hr sample of either 

enriched uranium (the default analyte through 1963) or plutonium (after 1963). Interferences are 

likely, because the methods were non-specific. Isotopes are all alpha-emitting isotopes of the analyte.

5.3.1.5.2 Reporting Levels, Minimum Detectable Activities, and Uncertainties

The reporting level for gross alpha through 1963 was ≥8.8 dpm/24-hr sample (10% of the RF 

tolerance level for enriched uranium). After 1963, the reporting level was ≥0.9 dpm/24-hr sample, 

credited to plutonium. (Gross alpha data are likely coded as G in the urine data reports.) 

Samples with results ≥0.9 dpm/24-hr sample typically, but not always, were counted using a PHA 

system to determine whether to credit the result to enriched uranium or plutonium, or a portion to 

both. The default condition, through 1963, was to credit the result to enriched uranium unless the 

PHA count indicated otherwise. After 1963 (and enriched uranium operations were phased out), the 

default condition was to credit the result to plutonium. In either case, the results should be considered 

to be upper bounds, because of the non-specificity of the analysis.

The MDA for gross alpha listed in Table 5.3.1.5.2-2 were determined as described in Attachment 5A. 
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Uncertainties for the gross alpha urine results have not been quantified or reported. To estimate the 

uncertainty for results without a reported uncertainty, a reasonable approach is to divide the median 

MDA value by 3.3, where 3.3 is the sum of kα and kβ and kα = kβ = 1.645 (see Attachment 5A). This 

uncertainty does not include the effect of interferences, which is a major issue for a non-specific 

analysis like gross alpha measurement.

Table 5.3.1.5.2-2. Median MDAs for gross 

alpha measurements.

a

 

Period dpm/24-hr sample

1952 1.0

1953 0.88

1954 – 1959 0.79

1960 – 1962 0.55

1963 0.55

1964 – 1971 0.69

5.3.1.6 Tritium

Workers were monitored for possible tritium exposures only for special projects or situations, starting 

in 1973. The methods have not been reviewed, but likely involved liquid scintillation measurements. 

The urine results are reported as pCi/liter of urine, and actual results were reported, generally with the 

standard deviation. It has not been determined whether the reported uncertainty in the 1970s to early 

1980s is one or two times the standard deviation. The sensitivity of the method was two to three 

orders of magnitude better than the significant level of about 1 μCi/liter. Although the actual MDA has 

not been quantified for the methods in the 1970s and 1980s, it likely is in the range of several hundred 

to several thousand pCi/liter. The current MDA for tritium is 600 pCi/liter (RFETS 1998c, p. 7-3).

5.3.2

In vivo lung counts have been performed at Rocky Flats since 1964 to determine the activity of 

plutonium in the lungs of workers who were exposed, or had the potential to be exposed, to airborne 

plutonium. The method of in vivo lung counts was to place one or more detectors over the chest of 

the subject and count the photons emitted from the plutonium mixture, if any, in the chest. Plutonium 

was not detected directly because of the low abundance of gamma photons and the severe 

attenuation of the more abundant L X-rays. Instead, the 59.5-keV gamma photon from 241Am was 

used to detect 241Am, which is present to some extent in all weapons-grade plutonium at Rocky Flats. 

The activity of plutonium was then calculated from the detected 241Am by measuring, calculating, or 

assuming the fraction of the 241Am in the plutonium mixture on the date of the lung count. At Rocky 

Flats, the fraction of the 241Am in the plutonium mixture has historically been characterized in terms of 

ppm by weight. Direct in vivo measurement of plutonium in the lungs, although investigated, was 

never implemented at Rocky Flats. 

Lung Count Data

The Rocky Flats lung counter also measured 234Th, via the 63-keV gamma (doublet) photon, to 

determine the activity of 238U in the lungs of workers exposed to depleted uranium. This 

measurement was made possible by the improved resolution of the germanium detectors 

(Ge detectors) that allowed baseline separation of the 59.5-keV gamma of 241Am from the 63-keV 

gamma doublet of 234Th. The activity of 238U was considered to be equal to that of the measured 

234Th, assuming equilibrium. 
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Attachment 5B, Minimum Detectable Activity for Urinalysis Methods at Rocky Flats, contains more 

detail. Section 5.4 and Attachment 5C discuss of the data and report forms. 

5.3.2.1 Americium/Plutonium

5.3.2.1.1 Methods, Units, Isotopes, and Interferences

Before April 1997, lung count data were not converted to a quantified amount or activity unless there 

was confirmation that the count was from an actual deposition in the lungs.. For unquantified results, 

the data are generally in units of counts per minute accompanied by a decision noted as normal, 

background, or some abbreviation of background. For quantified results through about 1968, the unit 

was micrograms of plutonium. In addition, the result was converted to a fraction of the maximum 

permissible lung burden (MPLB) using a plutonium specific activity of 0.07 μCi μg-1 and the MPLB of 

0.016 μCi (16 nCi) for the alpha-emitting isotopes of plutonium. Starting in about 1973, the activities 

of both plutonium (including all the alpha-emitting isotopes of weapons-grade plutonium) and 

americium (241Am) were recorded, in nanocuries. In addition, the activity of 241Am was stated as a 

fraction of the MPLB, using 14.7 nCi for the MPLB for 241Am. After 1989, the results were no longer 

stated as a fraction of the MPLB.

There are two sources of interferences to consider. The first is the 63-keV gamma doublet of 234Th 

from depleted uranium operations being mistaken for 241Am for lung counts with the NaI or phoswich 

detector systems. This interference was most troublesome for workers with residual lung depositions 

of plutonium and americium who subsequently worked in depleted uranium operations. The second 

interference is the contribution of count from 241Am not in the lungs, for example, contributions from 

contamination on the skin, from material being cleared from the upper respiratory system, or from 

ingested material. A positive detection of 241Am did not necessarily indicate an intake of the 

plutonium/americium mixture, especially for a lung count in response to an incident.

5.3.2.1.2 Reporting Levels, Minimum Detectable Activities, and Uncertainties

Reporting levels are not easily defined, because quantification was preceded by verification counts 

and professional judgments. In addition, before 1974, the practice was not to quantify a positive 

detection of 241Am unless the deposition could be associated with a known incident with a known ppm 

241Am. Affected workers were classified as positive unknowns or some variation. Starting in 1974, 

the practice was changed to quantify the plutonium depositions for positive unknowns by assuming a 

default value of 1,000 ppm 241Am on the date of the most likely intake or on the date of the first 

positive lung count. The ppm 241Am was then calculated for the date of the lung count to account for 

the ingrowth of 241Am from the nuclear transformation of 241Pu and the radioactive decay of the initial 

241Am. 

In general, this quantification was not applied retroactively to earlier positive lung counts. Once a lung 

deposition of plutonium had been quantified for a worker, the deposition continued to be quantified for 

all subsequent lung counts (except screening counts for new intakes), regardless of the result of the 

subsequent lung count (including negative values), until each of the last three results was less than 

the decision level for the count and the average of the last three results was within one standard 

deviation of 0.00 nCi plutonium.

The decision levels varied. From 1965 through 1968, the decision level was two times the uncertainly 

of the matched subject’s net count, when there was a decision level at all. Starting in 1969, for NaI 

and phoswich detector systems, the decision level was equal to three times the standard deviation of 

the net count rate for a set of lung counts for unexposed known cold subjects, based on the index 
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method (see Attachment 5B). Results between two and three sigma were noted but not always 

investigated. For the Ge detector systems, starting in 1976, the decision level (also called the cutoff) 

was equal to 1.645 times the standard deviation of the net count rate. The decision level, for 1995 

and later, was calculated by the ABACOS-Plus© software for a probability of a Type I (false positive) 

error of 5%. The decision level was used as a reporting level from 1995 through early 1997. 

Table 5.3.2.1.2-1 lists the MDAs for 241Am, which were calculated for the evolution of lung counting 

systems used at Rocky Flats as described in Attachment 5B. 

Table 5.3.2.1.2-1. Summary of MDAs for 241Am.

Period Detector system Index

MDA (nCi) for 241Am

Minimum system Standard system

Half time Full time Half time Full time

1964–1968 NaI(Tl) 4x4 0.90 1.7 1.5 1.3 1.2

1.35 2.8 2.5 2.1 1.9

1.80 4.6 4.1 3.5 3.2

1969 → NaI(Tl) 4x4 0.90 – – 0.80 0.76

1.35 – – 1.3 1.3

1.80 – – 2.2 2.0

1973 → Phoswich 0.90 – – 1.2 1.2

1.35 – – 2.0 2.0

1.80 – – 3.3 3.2

1976 –

1978

Ortec Arrays

(High-purity Ge)

0.90 0.26 0.18 0.20 0.14

1.35 0.48 0.32 0.37 0.25

1.80 0.86 0.59 0.66 0.45

1979 → Ortec Arrays

(High-purity Ge)

0.90 0.20 0.14 0.16 0.11

1.35 0.37 0.25 0.28 0.19

1.80 0.66 0.45 0.51 0.35

1978 → PGT I Arrays

(High-purity Ge)

0.90 0.22 0.15 0.17 0.12

1.35 0.40 0.27 0.31 0.21

1.80 0.71 0.49 0.55 0.38

1979 → PGT I Arrays

(High-purity Ge)

0.90 0.17 0.12 0.13 0.09

1.35 0.31 0.21 0.24 0.16

1.80 0.55 0.38 0.42 0.29

1979 → PGT II Arrays

(High-purity Ge)

0.90 0.22 0.15 0.17 0.12

1.35 0.40 0.28 0.31 0.21

1.80 0.74 0.50 0.57 0.39

1985 → PGT Organ Pipe Ge Detectors 0.90 – – 0.15 0.11

1.35 – – 0.26 0.18

1.80 – – 0.46 0.32

1991 → EG&G Organ Pipe Ge Detectors 0.90 – – 0.14 0.10

1.35 – – 0.26 0.18

1.80 – – 0.48 0.33

1995 → Ortec 2 Organ Pipe Ge 

Detectors

0.90 – – – 0.14

1.35 – – – 0.3

1.80 – – – 0.6

These values of MDAs are for three indices, representing the median and the approximate 5th- and 

95th-percentile body statures of Rocky Flats male workers. To obtain the worker-specific MDA, the 

dose reconstructor can calculate them using the information in Attachment 5B or interpolate (or 

extrapolate) from the values in Table 5.3.2.1.2-1. The worker-specific index is generally stated on 

lung count report forms from 1969 through 1994, and can be derived from the weight and height data

on report forms from 1995 and later. (The MDA values are reported on report forms from 1995 and 
Effective Date: 01/12/2004 Revision No. 00 Document No. ORAUT-TKBS-0011-5 Page 21 of 103

later, but the values are not worker-specific. The dose reconstructor should disregard these MDA 

values.) The default MDA would be for an index of 1.35, if height and weight data for the worker are 

not available.

The MDA for plutonium would be calculated by multiplying the worker-specific value of the MDA for 

241Am by the MDA conversion factor (Equation 5B-17 in Attachment 5B), which is based on the value 

of the ppm 241Am on the date of the lung count. The value of the ppm 241Am on the date of the lung 

count, accounting for ingrowth of 241Am from the nuclear transformation of 241Pu and the radioactive 

decay of the initial 241Am, is given by Equation 18 in Attachment 5B. The dose reconstructor needs to 

establish the date of the intake and the initial ppm 241Am. If that information is not apparent in the 

available records, a claimant favorable approach is to assume the initial ppm 241Am to be 100. 

The assumption of the intake date is not straightforward and should balance maximizing the plutonium 

lung deposition (intake date is close to the date of the lung count) and maximizing the accrued lung 

dose (intake date is far from the date of the lung count). In addition, the choice of intake date for the 

lung count data should be coordinated with that for the associated urine data. 

In addition, the dose reconstructor must choose the value of the initial fraction of 241Pu. At the Rocky 

Flats lung counter, 0.005 was historically used as the initial mass fraction of 241Pu and is a realistic 

choice for intakes that occurred in the 1950s through June 1976. The fraction 0.0036, based on the 

isotopic composition for Rocky Flats stream plutonium in the mid 1970s, should be used for intakes 

that occurred from July 1976 through 1989. For intakes incurred after 1989, the initial fraction of 241Pu 

should be reduced to account for the aging (radioactive decay) of the 241Pu.

The uncertainties of the results were reported for the net count per minute starting with the Ge 

detector systems in 1976. The uncertainty was reported at one standard deviation and included only 

the contribution from counting statistics. Starting in approximately 1981, the counting statistics 

uncertainty was also applied to the assessed activity and to the value of the fraction of the MPLB. 

With the advent of the ABACOS-Plus© software in 1995, the percent error at one standard deviation 

was reported for all identified nuclides. Beginning on October 11, 1999, a 30% systematic 

uncertainty, which included contributions of uncertainties in the chest wall thickness (CWT), the 

location of the activity in the lungs, the uncertainty in the ppm 241Am, and the influence of activity 

deposited in other organs, was included in the total propagated uncertainty (RFETS 2000, p. 3-18). 

The major uncertainty for the calculation of the plutonium lung deposition is the ppm 241Am in the 

plutonium in the lungs at the time of the lung count. Elements contributing to the uncertainty are the 

intake date, the value of the initial ppm 241Pu, the initial fraction of 241Pu, and the degree of association 

of the americium with the plutonium while in the lungs. An underlying assumption is that the 

americium remains associated with the plutonium particles in the lungs until the particles are 

dissolved or removed from the lungs. The degree of validity of this assumption has not been 

determined.

5.3.2.2 Thorium/Depleted Uranium

5.3.2.2.1 Methods, Units, Isotopes, and Interferences

The method to detect depleted uranium was to detect the 63-keV gamma (doublet) photon of 234Th 

and to calculate the activity of 238U, assuming equilibrium. This method was implemented manually 

for special cases in approximately 1978. Starting in 1983, the count data for the 63-keV doublet 

photon were routinely processed and reported. However, the activity of the 238U was calculated only 

for special cases and not routinely. A supplemental method, implemented in about 1989, detected the 
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93-keV gamma (doublet) photon of 234Th, and the count data were routinely processed and reported. 

This supplemental method was used mainly to reduce false positive results for the detection of 234Th, 

because detection of both doublet photons was required before detection of 234Th was considered. 

Starting in 1995, the activity of 238U was calculated and reported if the 63-keV peak (or sometimes the 

93-keV peak) was detected by the ABACOS-Plus© peak search software. If the peak was not 

detected, the activity of 238U was reported as less than the decision level (the activity of the decision 

level was reported). Starting in early 1997, the activity of 238U was reported, including negative 

results, even if a peak was not detected. In a similar manner, the activity of 235U was reported. 

Starting in about 1999, the activity of 238U was based solely on the 63-keV peak.

The main part of the data for the 63-keV doublet photon is in units of net counts per minute. To 

convert to activity (nanocuries) of 238U, the counts per minute is divided by the calibration factor for 

241Am (see Attachment 5B), normalized to the ratio of photon abundances [abundance of 59.5-keV 

gamma, 241Am, is 0.359; abundance of 63-keV doublet gamma, 234Th, is 0.0381 (Lederer and Shirley 

1978); ratio (59.5 keV gamma/63 keV doublet gamma) = 9.4]. That is, nCi 238U equals [(234Th 63-keV 

net cpm) divided by (241Am calibration factor)] multiplied by 9.4. To calculate the activity for depleted 

uranium, the 238U activity is divided by 0.89 (see Section 5.2.4.1). 

The interference is 238U in natural uranium. Unless there is an activity reported for 234U that is 

approximately equal to that reported for 238U, the dose reconstruction should use the claimant 

favorable assumption that the 238U activity is all from occupational exposure to depleted uranium.

5.3.2.2.2 Reporting Levels, Minimum Detectable Activities, and Uncertainties

Reporting levels were not generally used for depleted uranium until 1995 with the implementation of 

the ABACOS-Plus© software (see Section 5.3.2.2.1). Before 1995, the 238U activity was generally 

quantified only after verification of an intake.

The MDA for 238U has not been determined rigorously. However, the 238U worker-specific MDA can 

reasonably be expected to be a multiple of the 241Am worker-specific MDA because the detected 

photons (63 keV and 59.5 keV) are very close in energy. As described in Section 5.3.2.2.1 for using 

the calibration factor for 241Am to determine the 238U activity, the 238U worker-specific MDA can be 

obtained by multiplying the 241Am worker-specific MDA by 9.4. That result is divided by 0.89 to obtain 

the worker-specific MDA for depleted uranium. (As noted in Section 5.3.2.1.2 for americium and 

plutonium, MDA values are reported on report forms 1995 and later, but are not worker-specific. The 

dose reconstructor should disregard these MDA values.)

The major uncertainty is the assumption of equilibrium of the 234Th with the 238U before 1990, when 

depleted uranium was still being processed. Part of the process was to remove decay chain 

radionuclides, especially thorium, by heating the uranium ingot to drive the smaller atoms of thorium 

to the surface or top of the ingot, which was then cut off. The result was depleted uranium metal with 

a deficiency of 234Th for several weeks plus scrap depleted uranium with an excess of 234Th (superequilibrium). Super-equilibrium is claimant favorable. The effect of a deficiency of 234Th has not been 

assessed. 

The standard deviation of the net count rate is reported through 1995, but includes only the 

contribution of counting statistics. To estimate the uncertainty of a 238U or depleted uranium activity 

calculated from the net count rate, the dose reconstructor can divide the worker-specific MDA by 3.3.
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5.3.3

5.3.3.1 Wound Count Data

Other Bioassay Data

Wounds are defined as any break in the skin (e.g., cuts, punctures, abrasions, acid burns). Any 

wound that occurred in a work area involving plutonium was monitored for plutonium contamination, 

especially after the advent of the wound counter in 1957. Counting a blood sample or directly 

counting the wound site with an alpha detector were also methods used to monitor wounds to detect 

possible plutonium contamination. In Rocky Flats terminology in the 1950s and 1960s, wound counts 

were called gamma specs, and the wound counter was called a gamma spectrometer. Wounds 

occurring in uranium work areas were monitored selectively. The record could contain an incident 

report, a wound count data sheet, a medical decontamination report, and a medical treatment report, 

depending on the era and circumstances. 

The process was to attempt to decontaminate the wound in the building of the occurrence by washing 

and encouraging bleeding to flush any plutonium out of the wound. Then the worker was sent or 

escorted to the medical facility for a wound count and additional decontamination if the wound count 

was positive. The sequence of additional decontamination was washing with soap and water, 

washing with Clorox©, scrubbing with Clorox©, and excision.

Wound count information is largely irrelevant to dose reconstruction. The relevant items are the 

urinalysis data, the identification of the mode and date of intake, and whether there was residual 

plutonium at the wound site. If there was residual plutonium at the wound site, the dose reconstructor 

should consider an acute injection into the blood stream plus a possible long-term chronic injection. 

The profile of the urine data following the date of the wound provides guidance on the proportion of 

the acute and chronic components. If there was no detected residual plutonium at the wound site, 

there would have been an acute injection into the blood stream.

Residually positive uranium contamination in wounds was rare, if at all. It is reasonable and claimant 

favorable to assess any initially positive uranium wound as an acute injection.

5.3.3.2 Nasal Smears and Fecal Samples

Nasal smear (later called swab) and fecal sample data were occasionally obtained throughout Rocky 

Flats operations as supplemental data for workers with actual or suspected significant inhalation 

intakes. Through the 1980s, they were used subjectively to verify that an intake did occur and to 

estimate the possible magnitude of the intake. The data have also been used to determine or confirm 

the ppm 241Am in the inhaled plutonium mixture. Some obstacles to using nasal smear or fecal data 

to quantify an intake are unknown particle size distribution, unknown fraction of the plutonium 

captured by the nasal smear or fecal sample, inconsistent and largely undocumented sampling 

technique for nasal smears (which sometimes were called nose blows), and unknown counting 

efficiency (e.g., sample geometry and alpha absorption, especially in the 1950s and 1960s). Through 

1989, the requested fecal sample was the second voiding following the incident. In some cases, the 

second, third, and fourth voidings were requested. 

Starting in the 1990s, the nasal or mouth smears were used as a workplace indicator to identify 

potential intakes, and fecal sampling was used to confirm and evaluate suspected intakes (RFETS 

1998d, p. 2-2). 
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The reported MDAs (RFETS 1998d, pp. 2-7, 2-8) are:

• 20 dpm/sample, for (gross alpha, liquid scintillation) routine nasal samples 

• 0.20 dpm/sample, for fecal samples with a 21-day reporting time (plutonium alpha isotopic)

• 1.30 dpm/sample, for fecal samples with a 14-day reporting time (plutonium alpha isotopic)

• 2.6 dpm/sample, for fecal samples with a 7-day reporting time (plutonium alpha isotopic)

• 100 dpm/sample, for fecal samples with a 2-day reporting time (nonisotopic, rapid analysis)

These MDA values apply to samples starting approximately in 1993. MDA values for earlier years are 

not available.

 

5.4 RECORDS AND REPORTS

This section discusses the interpretation of the data and information on records and reports of 

bioassay data. Attachment 5C, Examples of Records and Reports Used at Rocky Flats, contains the 

figures described below.

5.4.1

Figures 5C-1 through 5C-3 are examples of the Urinalysis Record Card and the Health Sciences Data 

System – Urinalysis Detail report. The Urinalysis Record Card was the recording medium for the 

urinalysis data from 1952 through 1969 and is the primary record for urine data in this period. The 

urine data were manually entered on this card through 1969. These data were also entered into a 

database starting in about 1961. In about 1970, the Health Sciences Data System was implemented 

to record, process, and report urinalysis data and the derived fraction of the maximum permissible 

systemic burden. Examples of reports from this system are included in Figures 5C-1 through 5C-3 for 

instructional purposes.

Urinalysis Records and Reports

Interpretation of the Urinalysis Record Card

Urine results are presented in columns under the month for a given year (in the row). The top number 

is the day of the month (assumed to be the excretion day). The middle number is the sample result, 

either a number or BK (see Section 5.3.1.1.2). The bottom number is the technique code and refers 

to the codes in the header (see Attachment 5A, page 3). 

The unit of the result is given in the header. Sometimes the unit is written with the result (e.g., μg in 

Figure 5C-1). Be careful not to interpret μg as the number 49. 

The corresponding data on the Health Sciences Data System – Urinalysis Detail report should be the 

same as that on the Urinalysis Record Card. If not, the data on the Urinalysis Record Card should be 

taken as the correct data, with the exception noted in Section 5.3.1.1.2 (i.e., some plutonium results 

reported as BK on the card were re-reported with the actual result). On some cards, the dose 

reconstructor may observe the initially reported result crossed out and replaced by a lower value. The 

technical basis for that change has not been determined. In addition, that change generally was not 

applied to the data in the Health Sciences Data System. It is reasonable and claimant favorable to 

disregard the modified result. 

The analyte code for depleted uranium was sometimes transcribed incorrectly from the card to the 

urinalysis detail report as U (see Figure 5C-1) instead of D (see Figures 5C-2 and 5C-3) with the unit 

of dpm/24-hr sample instead of μg/24-hr sample. 
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Figures 5C-4 and 5C-5 are two versions of urinalysis reports from the Health Sciences Data System. 

Both versions report the data in the same way but with differences in the headers. A third version, 

Figure 5C-5 (the newer version), adds a column (the uncertainty of the result).

Interpretation of the Health Sciences Data System – Urinalysis Detail Report

The Activity Date is taken to be the date that the sample was excreted. However, the recorded date 

frequently was the date that the sample was received at the laboratory, especially for routine samples. 

(This applies also to the dates on the Urinalysis Record Card.)

ANAL is the code for the analyte:

P = plutonium

A = americium

U = enriched uranium (pre-1970, approximately)

U = depleted uranium (1970–1989, approximately)

D = depleted uranium (1952–1969, approximately)

G = gross alpha

NO CAL is a code used to flag the logic of the software.

0 = use normally in the calculation

1 = do not use in the calculation

2 = date of a new intake

Code 1 was used primarily for two situations to exclude a sample result from the systemic burden 

calculation: if the excretion of the analyte was enhanced by a chelation treatment or if the analysis of 

the sample did not meet quality standards (an invalid analysis or result). Sample results within 90 

days of a chelation treatment were generally (or should have been) coded as 1. Using Code 2 to flag 

the date of a new, significant intake was done inconsistently. In reports generated in the 1980s, an 

asterisk was used instead of a Code 2 to flag the date of a new intake. The dose reconstructor should 

disregard the Code 2 or asterisked entries. 

• ELAPSED DAYS is the number of days since the hire date. This data field is not likely to be of 

use.

• The EXPOSURE VALUE or DPM/24HR is the result of the urinalysis for the analyte. In 

general, the unit was dpm/24-hr sample, except for depleted uranium, from 1952 through April 

1964. 

• The column in parentheses is the uncertainty, starting in 1980. Any value or symbol in the 

parentheses before 1980 is only a placeholder and should be disregarded.

• The BODY BURDEN % or SYSTEM BURDEN is the calculated fraction of the maximum 

permissible systemic burden, calculated from Code 0 results for plutonium and for americium. 

This data field is not likely to be of use.

Figures 5C-6 and 5C-7 are examples of urinalysis reports from the on-site bioassay laboratory from 

1990 to the mid-1990s. Figure 5C-6 is for a special urine sample analyzed for plutonium, and Figure 

5C-7 is for a routine urine sample analyzed for plutonium. Both forms have the same format. The 

first three columns are self-explanatory.

• The Dec Level is the decision level Lc in unit of dpm/sample.
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• The Aspec is code for the alpha spectrometry quality. The Aspec codes are defined on the 

lower-left portion of the report. Aspec code 0 is analogous to the Code 0 previously used for 

urine data in the Health Sciences Data System. Codes 1, 3, and 4 indicate a failed analysis 

and disqualify the result.

• The DQO is the code for status of the data quality objectives for the results of the batch blank 

and control samples. The DQO codes are defined on the lower center portion of the report. 

DQOs, in theory, were assessed for the blank, accuracy, and precision. In practice, the DQO 

was usually assessed only for the blank. Hence, the notation ANN means that the blank was 

acceptable, the accuracy was not assessed, and the precision was not assessed. An F would 

indicate that the batch failed a data quality objective, and, if the batched failed, every sample 

in the batch was conditionally failed, pending further evaluation.

• The Batch Val is the overall validation of the result. V means valid, and I means invalid. Do 

not use a result that has an I validation code.

• The Analyte is self-explanatory.

• The Recovery is the fraction of the tracer recovered by the analysis.

• The Result (DPM) is the result of the sample in units of dpm/sample. The dose reconstructor 

should assume a 24-hr urine sample unless there is information that indicates otherwise.

• The Error is the uncertainty at one standard deviation.

Figure 5C-8 is an example of the urinalysis data report by Quanterra, a commercial off-site laboratory, 

starting in 1993. The form header information, except for the collection date and the matrix, is not 

useful. The collection date, if not the sample excretion date, should be replaced by the sample date 

written on the form. The result header is largely self-explanatory. 

• The Report Unit is the primary information is the result and its total error (at one standard 

deviation) in units of dpm/sample. 

• The decision level (Lc) and the sample-specific MDA are also stated. 

• The YIELD is the percent recovery of the tracer. 

• The RST/MDA is the ratio of the result and the sample-specific MDA. 

• The RST/CNTERR is the ratio of the result and the counting error. 

• The ANALYSIS DATE is the date the sample was analyzed, not the excretion date. 

• The ALIQUOT SIZE is the volume of the sample, in milliliters (ALQ UNIT). 

• The DETECTOR ID is self-explanatory. 

• The METHOD NUMBER references the document number of Quanterra’s analytical procedure 

used to process the sample.
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Figures 5C-9 and 5C-10 are examples of the analytical report of the on-site bioassay in the mid1990s. Most of the information is self-explanatory. Some points:

• The date sampled is the excretion date.

• The data can only be used if the Alpha Spec Condition Code is 0 and if the Data Validation 

Code is V. 

• The 234U activity is approximately equal to 238U activity in Figure 5C-9, and both results are 

greater than the decision level. As stated in Section 5.3.1.4.1, this is the classic pattern 

indicating natural uranium, and not an occupational intake of depleted uranium.

Figure 5C-11 is an updated version of urinalysis data report of Quanterra. The significant 

improvement is the validation of each result (QUAL = V). Use only results with a QUAL = V. 

Figure 5C-12 is the urinalysis data report for General Engineering Laboratories, Inc. (GEL). The 

header information is largely self-explanatory. 

• The Date Collected is the sample excretion date. The time (0600 military time) is also noted. 

0600 is used as a default end time of the 24-hr excretion period if the actual end time is not 

documented.

• The VF is the volume fraction, the fraction of the sample that was analyzed. VF = 1 indicates 

that the entire sample was analyzed.

• Use only data that has a Data Validation Code = V.

Figure 5C-13 is an example of data card used in the 1970s and 1980s to record data manually for 

tritium urine samples and for other samples such as fecal samples and nose smears. The unit of the 

tritium results is pCi/liter. The unit of the fecal sample and nasal smear results is dpm/sample.

There may be other versions of in vitro bioassay reports. In all cases, the important data are the 

excretion date, the analyte, the result in the proper units, and whether the result was valid.

5.4.2

Figure 5C-14 is an example of an early lung count report. The aftermath of the October 15, 1965, 

plutonium fire in Buildings 76 and 77 was the first extensive use of the lung counter to detect 

americium and plutonium depositions for Rocky Flats workers. 

Lung Count Records and Reports

• The in vivo lung counting system was called the Body Counter. In Rocky Flats terminology, 

the lung count was called a body count through 1989. Most claimants will likely use the term 

body count instead of lung count. The dose reconstructor should not mistake the Rocky Flats

body count for a whole body count, which was widely used at other facilities to detect intakes 

of fission products.

• The Time field was used either for the time of the day at the start of the count or for the length 

of the count. In this case, the length of the count was noted 40 MLT means 40 minutes live 

time).
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• The Minus Bkg + match notation indicates that the result is the net count rate after the room 

background count rate and the net count rate of a matched person was subtracted.

• The 1.4 LB notation is the calculated plutonium deposition in terms of the multiple of the MPLB 

of plutonium (1 MPLB = 16 nCi Pu alpha emitters).

• The Body Location is the position of the detector. In this case, the detectors were positioned 

over the right and left portions of the chest. In many early counts, one of the detectors was 

positioned over liver, gut, or below sternum instead of over one side of the chest. That data 

has little dosimetric use.

Figure 5C-15 is the August 1967 revision to the Health Physics Body Counter Information form. The 

change was to present the results after subtraction of the room background, Net (1) cpm, and after 

subtraction of matched subject net cpm, Net (2) cpm. In addition, the plutonium deposition was stated 

in terms of micrograms of plutonium.

Figure 5C-16 is the August 1968 revision to the Health Physics Body Counter Information form.

• The Net cpm is the subject’s total count rate minus the room background count rate. 

• The Predicted cpm replaced the net count rate of the matched subject. 

• The Result is the final net cpm. 

• In this example, there is no measurement for the right chest. The dose reconstructor should 

estimate the contribution for the right chest before using data from this count, because the 

lung data set generally includes contributions from both right and left lungs.

Figure 5C-17 is an example of a lung count with no result tabulated. This is an example of a positive, 

unknown case (see Section 5.3.2.1.2). In addition, note the tabulation of the index, which was used 

later to estimate the chest thickness. Sufficient information is presented here and in Attachment 5B to 

allow the dose reconstructor to calculate the plutonium and americium activities for this lung count, for 

any assumed or actual intake date.

Figure 5C-18 is the December 1973 revision to the previous form, with expanded information. 

• The ROOM is the designation of the counting chamber, A, B, or C, used for this count. 

• The RATIO field documented the ratio of the 241Am photopeak region of interest (ROI) and a 

background ROI around 100 keV. The ratio was used as a supplemental, subjective tool to 

improve detection of americium. A ratio of 1.20 or greater indicated likely detection of 

americium. 

• The ppm 241Am was used to record either the ppm 241Am for a new incident or, as in this case, 

the calculated value of the ppm 241Am (including ingrowth of 241Am) for a prior actual or 

assumed intake. 

The form included fields to record the activity and fraction of the MPLB for both plutonium and 

americium. (This lung count, now quantified, is for the same positive, unknown case as Figure 5C17).
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Figure 5C-19 is an example of the previous form for a count judged to be background. Data fields 

were added to capture data for measurements of the L X-ray (17 keV) ROI, especially for the 

phoswich detector system. Although that information was captured occasionally, the data were not 

used because of the instability of the predicted background cpm.

The previous lung count reports were for counts using the NaI detector system. Figure 5C-20 is an 

example of the lung count data for a Ge detector system. The data for the five to eight detectors of 

the Ge systems were multiplexed into a composite total count tabulated in the row for TOTAL CHEST. 

The standard deviation of the resultant counts per minute is based on counting statistics only. For 

workers with confirmed lung depositions, the calibration factors for plutonium and americium were 

generally written on the form, as in this case.

Figure 5C-21 is an example of the first computer report for the lung count results. The data are 

labeled appropriately. This report is for a worker with a confirmed deposition. The report for workers 

without a confirmed deposition does not report the calibration factors, the ppm Am, or the lung 

burden. Instead, it reports the cutoff, which is the decision level, and Normal if the DIFFERENCE is 

less than the cutoff.

Figure 5C-22 is an example of a computer report for the phoswich detector system, which was used 

as a backup screening system in the 1980s. Note the outcome statement, “RESULTS ARE 

NORMAL.” If the results were not normal, the subject would have been recounted with a Ge detector 

system. Because the phoswich system could not resolve the 60- and 63-keV photopeaks, they share 

a common ROI. Another feature is the tabulation of the total count for each pertinent ROI. ROI 3 is 

the total count for the 60-keV/63-keV ROI, and ROI 4 is the background count for the 60- and 63-keV 

photopeaks. ROI 4 was also used as the count for the 93-keV photopeak, and ROI 5 was its 

background. ROI 2 was likely the count in the L X-ray region, but it was not used.

Figures 5C-23 and 5C-24 are examples of the next generation of reports for the Ge detector systems. 

The innovation is the data capture in ten ROIs. In Figure 5C-24, the ROIs are labeled with the 

photopeak of interest. Although the data were captured, most of the data were not used, mainly 

because the relationship between the photopeak and its background was not established or was too 

variable. ROI 5, labeled BKG in Figure 5C-24, is the common background (divided by a factor) for 

both the 60- and 63-keV photopeaks.

Figure 5C-25 is an example of a report for a worker with a confirmed deposition. There are no new 

fields.

Figure 5C-26 is an example of a report for a worker with no detected deposition and illustrates a 

frequent problem with the L X-ray data, the problem of low-end electronic noise in one or more of the 

detectors. The dose reconstructor should disregard all L X-ray data (including the 13- and 17-keV 

ROIs).

Figure 5C-27 is an example of a report where data for the 93-keV photopeak is analyzed and 

presented.

Figure 5C-28 is an example of the next generation of reports. On this report, the ROI data for each 

detector is tabulated separately, as well as the sum. ADC #1 stands for analog-to-digital converter for 

detector #1, which, in this case, is an EG&G detector; and similarly for the other detectors. This 

report also does not report the results in terms of the fraction of the maximum permissible lung 

burden, an obsolete concept since 1989.
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Figure 5C-29 is an example of the lung count report generated by an early version of the ABACOSPlus© software, to mid-February 1997. Because this software is based on a peak search method, no 

ROI data are available. In addition, if a uranium or americium peak was not found, the activity was 

reported as less than the decision level. 

Figure 5C-30 is an example of the lung count report generated by the ABACOS-Plus© software, after 

mid-February 1997. The activities of 235U, 238U, and 241Am are calculated and reported, even if the 

peak was not detected or if the result was negative. The MDA values are for the average worker, as 

stated on the report. The MDA value for 238U is lower than the worker-specific decision level for this 

case. The worker-specific MDA should be at least twice the worker-specific decision level.

Figure 5C-31 is an example of the lung count report from the ABACOS-Plus© software for a worker 

with a confirmed deposition. The software calculated the deposition for the plutonium isotopes based 

on the intake date, noted in the header, and on the calculated ppm 241Am (including in-growth) based 

on the value of the initial ppm 241Am in the worker’s file. The % Error for 241Am is assigned to the 

plutonium isotopes. The basis of the decision level for the plutonium isotopes is not obvious, but is 

likely the decision level for detecting the L X-rays. In any case, the decision level value listed here 

does not apply and should be disregarded for the plutonium isotopes. The value of the ppm 241Am on 

the date of the count is not reported on lung count reports generated by the ABACOS-Plus© software. 

This value can be calculated using Equation 5B-18 in Attachment 5B, using the value of the initial 

ppm 241Am tabulated generally on one of the early lung count reports.

Much of the information from the ABACOS-Plus© software is not useful, including Count Rate, 

Detector Count Rate, Analysis Limits, and the total activity.

The dose reconstructor should note the intake date. If the intake date is different from the date for 

Count Started, the intake date is from the file for a worker with a confirmed deposition. Otherwise, the 

date of the lung count is used as the intake date.

The dose reconstructor should be aware that the lung counter detectors were also used for wound 

counts. Reports of wound measurements, including the calibration of the detector using americium 

and plutonium sources, look the same as the lung count reports except for some header information 

(name, employer, job code, reason, height, or weight). 

It is important to note that the activities calculated for plutonium for lung counts are based on a 

specific, actual, or assumed intake date and initial ppm 241Am. The plutonium values are valid and 

appropriate only for that intake data. If the dose reconstructor chooses to use another intake date or 

initial ppm 241Am, the dose reconstructor should recalculate the set of plutonium lung deposition 

activities based on the recalculated ppm 241Am for ingrowth. This is accomplished by multiplying the 

original activity of plutonium by the ratio (the original ppm 241Am on the date of the count divided by 

the new value of the ppm 241Am on the date of the count). The new value of the ppm 241Am on the 

date of the count can be calculated using Equation 5B-18 in Attachment 5B. The dose reconstructor 

also should adjust the activities for the discontinuity factors presented in Attachment 5B. Generally, 

use of the discontinuity factors is claimant favorable. 

5.5 INTERNAL DOSIMETRY FOR UNMONITORED WORKERS

This section presents guidance to the dose reconstructor to estimate internal doses to workers who 

worked in buildings involving plutonium, enriched uranium, or depleted uranium but have no 

appropriate bioassay data in their record. The dose reconstructor is the judge of the presence of 

appropriate bioassay data. For example, if a worker was potentially exposed to a radionuclide such 
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as plutonium, but had bioassay data for a different radionuclide such as enriched uranium, the worker 

is unmonitored for plutonium but is monitored for enriched uranium, if exposures to the different 

radionuclides are not linked.

5.5.1

[Please see Note to reviewers preceding the Table of Contents.] This section pertains to workers with 

no record of involvement in an incident and no citation of an incident in the claimant interview 

involving a radionuclide associated with the building(s) where they worked. This guidance suggests a 

reasonable approach, based on the air concentration limits in effect during the period of potential 

intake by the unmonitored worker.

Internal Dosimetry for Workers Not Involved in an Incident

The air concentration limits, stated in units of μCi/cm3, were [To be determined: we need a table of air 

concentration limits for the history of Rocky Flats, not only the official limits in AEC, ERDA, and DOE 

orders but also the values actually used. The old, in-house limits were called RCG, which, I believe, 

stands for radioactivity concentration guides. We need a history of the Rocky Flats RCG values for 

airborne radioactivity. We know some values. For example, the Rocky Flats limit for plutonium was 

8.8 dpm per cubic meter of air, in the 1950s, based on calculations from incident and supplied air 

involvement reports that listed both the dpm of the plutonium in the air sample and the % of the limit. 

We also know, from annual reports of the Radiation Safety Section at Rocky Flats (Radiation Safety 

1987), that the in-plant guides in the mid-1980s were: 4.4 dpm/m3 for soluble plutonium, 9.0 dpm/m3

for insoluble plutonium, and 70.0 dpm/m3 for uranium. Maybe the Rocky Flats in-plant guides did not 

change much over the first 38 years.]

The suggested approach is to estimate the time spent by the worker in a building involved in a 

radionuclide of interest and credit the worker with a chronic intake at an arbitrary fraction of the inplant guide (or official limit, whichever is more claimant favorable). In the 1950s, the practice at Rocky 

Flats was to monitor workers only if they were expected to be exposed to 10% or more of the limit 

(called tolerance). Later, the goal was to operate at less than 10% of the RCG and to investigate 

conditions if an air sample exceeded 100% of the RCG. A reasonable, though arbitrary, choice of the 

default fraction is 10% of the RCG or official limit, unless worker-specific information is available. The 

default values and assumptions need to be determined.

[To be determined: Table of intake values, based on 100% of the RCG values, once the timeline of 

RCG values is obtained. We also need a concurrence that a default intake at 10% of the limit is 

reasonable.]

5.5.2

Claim files may include event-specific data that should be used to reconstruct internal dose. When 

such data is not available default assumptions may be made.

Internal Dosimetry for Workers Involved in an Incident

[Section 5.5.2 is reserved -- To Be Determined. This issue is worker specific, and likely would need to 

be researched on a case-by-case basis. ]
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REFERENCES

HPS (Health Physics Society), 1995, An American National Standard – Bioassay Programs For 

Uranium, HPS N13.22-1995.

ICRP (International Commission on Radiological Protection), 1979, Limits for Intakes of Radionuclides 

by Workers, ICRP Publication 30, Part 1, Pergamon Press, New York, New York.

ICRP (International Commission on Radiological Protection), 1994, Human Respiratory Tract Model 

for Radiological Protection, ICRP Publication 66, Pergamon Press, Oxford

James, W.H., 1990, “Isotopic Profile of Rocky Flats Plutonium,” interoffice correspondence, February 

28, 1990, EG&G Rocky Flats.

Lawrence, J.N.P., 1984, Uranium Internal Exposure Evaluation Based on Urine Assay Data, LA10246-MS, Los Alamos National Laboratory, U.S. Department of Energy, Los Alamos, New 

Mexico. 

Lederer, C.M., and Shirley, V.S. (eds.), 1978, Table of Isotopes, Seventh Edition, Lawrence Berkeley 

Laboratory, University of California-Berkeley, Berkeley, California, and John Wiley & Sons, 

New York, New York.

Mann, J.R., and Kirchner, R.A., 1967, “Evaluation of Lung Burden Following Acute Inhalation 

Exposure to Highly Insoluble PuO2,” Health Physics, Volume 13, pp. 877-882.

Radiation Safety, 1987, 1986 Annual Report-Radiation Exposures to Rockwell Employees at the 

Rocky Flats Plant, Rockwell International, North American Space Operations, Rocky Flats 

Plant, Golden, CO.

RFETS (Rocky Flats Environmental Technology Center), 1998b, Statement of Work for Analytical 

Measurements, General Bioassay Service Module BA01-B.4, U.S. Department of Energy, 

Golden, CO.

RFETS (Rocky Flats Environmental Technology Center), 2003, “Internal Dosimetry Practices,” 

Internal Dosimetry Technical Basis Documentation, Section 1, Version 8, U.S. Department of 

Energy, Golden, Colorado.

RFETS (Rocky Flats Environmental Technology Center), 1998d, “In-Vitro Radiobioassay,” Internal 

Dosimetry Technical Basis Documentation, Section 2, Version 3, U.S. Department of Energy, 

Golden, Colorado.

RFETS (Rocky Flats Environmental Technology Center), 2000, “In-Vivo Radiobioassay,” Internal 

Dosimetry Technical Basis Documentation, Section 3, Version 8, U.S. Department of Energy, 

Golden, Colorado.

RFETS (Rocky Flats Environmental Technology Center), 2002, “Plutonium and Americium,” Internal 

Dosimetry Technical Basis Documentation, Section 1, Version 6, U.S. Department of Energy, 

Golden, Colorado.
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RFETS (Rocky Flats Environmental Technology Center), 1998a, “Uranium,” Internal Dosimetry 

Technical Basis Documentation, Section 1, Version 3, U.S. Department of Energy, Golden, 

Colorado.

RFETS (Rocky Flats Environmental Technology Center), 1998c, “Other Isotopes,” Internal Dosimetry 

Technical Basis Documentation, Section 1, Version 3, U.S. Department of Energy, Golden, 

Colorado.

DOE (U.S. Department of Energy), 1980, Final Environmental Impact Statement Rocky Flats Site, 

Volume 1, DOE/EIS-0064, Washington, D.C.
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GLOSSARY

alpha particles

Positively charged particles of discrete energies emitted by certain radioactive 

materials; alpha particles usually expend their energy in short distances and will not 

usually penetrate the outer layer of skin; they are a significant hazard only when taken 

into the body where their energy is absorbed by tissues.

curie

A special unit of activity. One curie equals 3.7 x 1010 nuclear transitions per second.

detection limit (lower)

The minimum quantifiable exposure or neutron flux that can be detected.

dosimetry

The science of assessing absorbed dose, dose equivalent, effective dose equivalent, 

etc., from external or internal sources of radiation. 

exposure

As used in the technical sense, exposure refers to a measure expressed in roentgens 

(R) of the ionization produced by photons (i.e., gamma and X-rays) in air. As used in 

internal dosimetry an encounter with uncontained radioactive material.

extremity

That portion of the arm extending from and including the elbow through the fingertips, 

and that portion of the leg extending from and including the knee and patella through 

the tips of the toes.

gamma rays

Electromagnetic radiation (photons) originating in atomic nuclei and accompanying 

many nuclear reactions (e.g., fission, radioactive decay, and neutron capture). 

Physically, gamma rays are identical to X-rays of high energy, the only essential 

difference being that X-rays do not originate in the nucleus. 

isotope

Elements having the same atomic number but different atomic weights; identical 

chemically but having different physical and nuclear properties

maximum permissible lung burden (MPLB)

The occupational limit for plutonium expressed in terms of a quantity of plutonium that 

could be present in the chest at any given time.

minimum detectable activity (MDA)

Limit of radionuclide activity detection for measurements of specific types and energies 

of radiation

photon

A unit or particle of electromagnetic radiation, photons originating from the nucleus or 

extra-nuclear material of an atom are called respectively gamma rays or X rays. 

radiation

Alpha, beta, neutron, and photon radiation. 
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radioactivity

The spontaneous emission of radiation, generally alpha or beta particles, gamma rays, 

and neutrons from unstable nuclei.

radionuclide

A radioactive isotope of an element, distinguished by atomic number, atomic weight, 

and energy state

rem

A unit of dose equivalent equal to the product of the number of rad and the quality 

factor.

whole-body dose

Commonly defined as the absorbed dose at a tissue depth of 1.0 cm (1000 mg/cm2

); 

however, also used to refer to the recorded dose.

X-ray

Ionizing electromagnetic radiation of external nuclear origin.
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ATTACHMENT A 

MINIMUM DETECTABLE ACTIVITY FOR URINALYSIS METHODS AT ROCKY FLATS

TABLE OF CONTENTS

A.1 Introduction...............................................................................................................................37

A.2 MDA Methodology .................................................................................................................... 37

A.3 History of Methods .................................................................................................................... 38

Reference..........................................................................................................................................54
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A.1 INTRODUCTION

Urinalysis has been used at Rocky Flats since the start of operations in 1952 to detect intakes of 

radionuclides by workers who were exposed, or had the potential to be exposed, to plutonium, 

enriched uranium, or depleted uranium. Urinalysis involved the submission of a urine sample by the 

worker, a chemical processing of the sample to isolate the radionuclide of interest (the analyte), and 

measurement and calculation of the quantity (or activity) of the analyte in the sample. The request for 

submission of the urine was either scheduled as part of a routine monitoring program or was specially 

requested following an actual or suspected intake. Routine urine samples were typically 24-hr 

excretions, either one continuous 24-hr period (but not taken at the Rocky Flats site) or two 12-hr 

periods. Special urine samples could be 24-hr samples, overnight samples, or a single voiding. The 

chemical processing of the sample depended on the analyte and the need for specificity and recovery. 

Specificity refers to separation of the desired radionuclide from interferences such as other 

radionuclides. Recovery refers to isolating as much of the analyte in the final medium to be measured 

(counted) as possible. The measurement of the sample typically involved counting the alpha radiation 

from the processed aliquant of the sample and determining the activity of the analyte in the original 

sample. Also involved was the fluorometric measurement of mass of depleted uranium. The 

assessment of the minimum detectable activity (MDA) involves the determination of the activity of the 

analyte in the original urine sample that would be expected to be detected by the methods and 

systems used at Rocky Flats from 1952 to the current time. The analytes of interest are plutonium, 

americium, enriched uranium, and depleted uranium. Also addressed is a category called gross 

alpha, which was a nonspecific analysis used for workers from 1952 to 1971 who were potentially 

exposed to any of the analytes of interest. This attachment focuses on the period from 1952 to 1971, 

for which many of the urinalysis logs have been located and analyzed to obtain the information 

needed to assess the MDA. This also is the period when urinalysis procedures were primitive and

evolving and numerous dosimetrically interesting events and intakes were occurring at Rocky Flats. 

A.2 MDA METHODOLOGY

The general equation for the MDA is Equation 6 in the American National Standard, Performance 

Criteria for Radiobioassay (HPS 1996):

MDA = (1 + ∆K) (2∆BB + 2kso + 3) / KT (5A-1)

where:

B = the total count of the appropriate blank

so = the standard deviation in the net count of a sample with no additional analyte;

so = √ [sB1

2

+ (1/m2) sB0

2

] (5A-2)

where:

sB1 = the standard deviation of the sample, where the sample contains no actual 

analyte above that of the appropriate blank

sB0 = the standard deviation in the unadjusted count of the appropriate blank

m = the adjustment factor for the appropriate blank

K = calibration factor

∆K = the maximum fractional systemic error bound in the calibration factor K

∆B = the maximum expected fractional systemic error bound in the appropriate blank
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k = the abscissa of the standardized normal distribution corresponding to the 0.05 probability level (for 

α = 0.05 and β = 0.05, k = 1.645)

T = the standard counting time for the procedure

Applying this equation to urinalysis methods at Rocky Flats involves determining the value of each 

variable for measurements of the analytes: plutonium, americium, enriched uranium, depleted 

uranium, and gross alpha, as the methods evolved.

A.3 HISTORY OF METHODS

General Information

In the beginning of operations (1952), the Rocky Flats Plant was divided into four distinct sub-plants, 

plus a general support area. The sub-plants were named A Plant, B Plant, C Plant, and D Plant. The 

designations A, B, C, and D are significant because they are also the code names for the materials 

processed in those plants as well for the urinalysis procedures used to analyze those materials. The 

records of the 1950s do not contain the words: depleted uranium, enriched uranium, and plutonium. 

Instead, depleted uranium is A material processed in A Plant (buildings numbered 4##, mainly 

Building 444); enriched uranium is B material processed in B Plant (buildings numbered 8##, mainly 

Building 881); and plutonium is C material processed in C Plant (buildings numbered 7##, mainly 

Building 771). D Plant (buildings numbered 9##, mainly Building 991) handled all materials. A 

nonspecific gross alpha urinalysis method was used for workers in D Plant. (Note: Building numbers 

were two digit numbers until 1968, when the numbers were expanded to three digits, e.g., Building 

771 was originally Building 71.) From 1962 to 1963, the enriched uranium operations were phased 

out at Rocky Flats, although urinalysis monitoring for enriched uranium continued through 1971. 

The Urinalysis Record Card is an important and significant record for the early (1952 to 1969) urine 

data and methods that generated that data for a specific worker. A Urinalysis Record Card was 

established for each monitored worker, on which the result of each urine sample, the date of the 

sample, and the code of the urinalysis method used to generate that result are recorded. The card is 

now in the worker’s Health Physics file, which is the primary Rocky Flats record of dosimetry 

information for a worker. The method codes are:

A Fluorimeter, reported in micrograms/liter (1952-1956); reported in micrograms/24-hr (1957-1964)

B1 Electroplating, reported in disintegrations per minute per 24-hr (dpm/24-hr). (Note: electroplating, in 

Rocky Flats records, more properly should be called electrodeposition.)

B2 Ether extraction, reported in dpm/24-hr

B3 Tributyl phosphate (TBP) extraction (hand written on some cards)

C1 Carrier precipitation, reported in dpm/24-hr

C2 Thenoyl trifluro acetone (TTA) extraction, reported in dpm/24-hr. (Note: On the header of cards for the 

period 1961-1965, the code C2 is “Pu by Radio Autography.” There is no indication that this method 

was implemented at Rocky Flats.)

D TBP extraction

Although there is some correlation of the codes with the sub-plants, there are some exceptions. The 

following summarizes the correlation of the method code and the analyte.

Analyte Method code

Depleted uranium A, B1 (starting 5/1/64)

Enriched uranium B1

Plutonium C1, C2

Gross alpha B2, B3, D
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Tolerance levels were used at Rocky Flats in the 1950s and 1960s as an indicator of the maximum 

permissible amount (activity) of a radionuclide excreted per day in a worker’s urine. The technical 

basis for the values of tolerance levels used at Rocky Flats has not been identified. The significance 

is that urinalysis results less than 10% of the tolerance level were recorded and reported as 

background (BK on the Urinalysis Record Card) or zero, regardless of the underlying sensitivity of the 

method, with some exceptions. The values of the tolerance levels are:

Analyte Tolerance level Reporting level

Depleted uranium 58 μg/24-hr ≥ 5.8 μg/24-hr

Enriched uranium 88 dpm/24-hr ≥ 8.8 dpm/24-hr

Plutonium 8.8 dpm/24-hr ≥ 0.88 dpm/24-hr

Gross alpha 88 dpm/24-hr ≥ 8.8 dpm/24-hr

These reporting (and recording) levels continued through April 1964 for both depleted and enriched 

uranium, through 1961 for plutonium, and through 1963 for gross alpha. After April 1964 through 

1971, the reporting level for depleted and enriched uranium was ≥0.20 to 0.28 dpm/24-hr. After 1963 

for gross alpha, the reporting level was ≥0.9 dpm/24-hr. 

For plutonium, the reporting and recording level was ≥0.20 dpm/24-hr for the period 1962 through 

April 6, 1970. After that date, all results ≥0.00 dpm/24-hr were recorded and reported. Negative 

values were recorded and reported as 0.00 dpm/24-hr. A further exception is that, for some workers, 

the practice implemented on April 7, 1970 was applied retroactively for their plutonium data. The 

retroactive application was variable with respect to how far back it was applied.

In 1963, an analysis specific for 241Am was implemented. The recording and reporting level for 241Am 

was ≥0.24 dpm/24-hr in 1963, ≥0.20 dpm/24-hr from 1964 to 1967, and ≥0.30 dpm/24-hr from 1968 to 

1971.

The general method for data analysis for alpha counting procedures (1952 to 1971) was:

Activity (dpm/24-hr sample) = (C/T – BDet – BBlk) x (V/A) / (ε x R) (5A-3)

where:

C = Total count (c) 

T = Count time (minutes, m)

BDet = Detector background count rate (cpm)

BBlk = Reagent blank count rate (cpm)

V = Sample (or standard) volume (ml)

A = Volume of the aliquant analyzed (or volume of the sample, if the entire sample was 

analyzed) (ml)

ε = Efficiency (geometry) of the detector (cpm per dpm)

R = Recovery, fraction of the analyte in the aliquant or sample that is transferred to the 

planchet or disk to be counted

The detector background count rate was generally tabulated in the urinalysis data logs through 1961. 

After 1961, the value used for the detector background is implicit in the data reduction but is not 

explicitly recorded. The same detectors were used for alpha counting for all analytes.

Reagent blanks were generally processed with each batch of samples, and the value of the blank 

count rate used in the data reduction was generally tabulated in the urinalysis data logs.
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The ratio (V/A) is a volume adjustment factor used for two purposes. If the entire sample was not 

analyzed, this ratio normalizes the result from the volume of the aliquant analyzed to the total sample. 

If the volume of the total sample was less than a minimum specified volume (e.g., 1,000 ml), the 

sample was considered to be less than a 24-hr sample, and the ratio was used to normalize the 

sample result to that for a 24-hr sample. The sample volume was recorded in the urinalysis data log 

for each sample.

The value of ε was the geometry rating of the detector. From 1952 to 1953, ε was 0.45. After that, 

the detectors were called 50% detectors, and ε was 0.50. In 1964, 40% detectors (ε of 0.40) were 

added to the system as a supplement to the 50% detectors.

The value of R was generally a standard value. Depending on the process, spiked samples, samples 

to which a known activity of the analyte was added, were generally processed with each batch of 

samples. The recovery values calculated from the spiked samples were the ratios of the count rate of 

spiked sample to the average count rate of four to six samples deposited on the planchet or plate with 

minimal processing. The recovery values for the spiked samples were not normalized to the activity 

(dpm) deposited. In addition, the recovery values from the spikes usually were not used to customize 

the standard value of R for samples in the batch.

The fraction of absorption of the alpha particles in the residue on the planchet or plate was not 

explicitly incorporated either in the efficiency or recovery.

The term ε x R was frequently combined, especially in the 1950s. In the 1960s, the term 1/( ε x R) 

was occasionally tabulated in the urinalysis data logs as R.F., presumably for recovery factor, and 

was used as a multiplier to convert the net count per minute to activity in the sample.

The general method for the mass measurements of uranium using the fluorimeter (1953 to 1964) was:

Mass (μg/24-hr sample) = (S – BBlk) / K (5A-4)

where:

S = Signal reading of the sample aliquant 

BBlk = Signal reading of the blank 

K = Constant/V (The constant is custom to each process. V = Volume (ml) of the entire 

urine sample. If the sample volume ≤ 1,000 ml, V = 1000 ml.)

The history of the methods is largely based on an interview with the lead chemist for the bioassay 

program from 1961 to 1992 and on a review of the bioassay data logs from 1952 to 1971.

Plutonium

1952–1961

The urine sample was processed using a method called carrier precipitation (also called coprecipitation). The plutonium in the urine sample (plus some americium and thorium) was carried into 

the precipitate with lanthanum fluoride. The precipitate was dissolved, and the solution was 

evaporated on a planchet, which was counted with a gas-flow proportional counter. Typical count 

time was 150 minutes. A spike sample and a reagent blank sample were processed with the workers’ 

samples, sometimes with each batch and sometimes less frequently. The result of the spike sample 

may or may not have been used to establish the value of the recovery of the analyte for the batch. 

Similarly, the result of the blank (counts per minute) may or may not have been used to establish the 
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value of the blank subtracted from the total count rate of the sample. Detector efficiency was stated to 

be 0.50. A volume adjustment factor (1,200/sample volume) was applied as a multiplier to the result if 

the sample volume was less than 1,000 ml. The first evidence of the use of this factor is in 1960. 

1961–1962

Starting on December 13, 1961, a TTA extraction step was added to the carrier precipitation method 

to improve the specificity of the process to isolate plutonium. No other changes were made to the 

previous method.

1963–1978

The ion exchange method replaced the carrier precipitation/TTA extraction method in 1963 and was 

used, with refinements, thereafter. The method was plutonium specific. In addition, americium could 

be recovered separately from the plutonium in the same sample. Evaporation of the analyte on a 

planchet was continued, but that method was gradually phased out and was replaced by 

electrodeposition on a stainless steel disk. About one third of the samples were electrodeposited in 

1964 and one half or more from 1967 through 1971. In 1973, an alpha pulse height analysis counting 

system with surface barrier detectors was started with four detectors. The practice of using internal 

tracers (236Pu or 242Pu) for some plutonium samples was begun concurrently. A batch blank continued 

to be processed, although its use was inconsistent. For example, in 1971, a blank count rate of 0.00 

cpm was used even though the median value of the batch blank was 0.06 cpm. In 1964, detectors 

with an efficiency of 0.40 were used as a supplement to the detectors with 0.50 efficiency 

1978–1993

By 1978, all of the counting system had been converted to the PHA system, and all of the plutonium 

samples were processed with internal tracers. The fraction of the internal tracer recovered for that 

sample was applied in the analysis of the result for that sample. The acceptable range of the 

fractional tracer recovery was 0.10 to 1.10. The result of a sample was invalidated if the recovery was 

outside of the acceptable range. In 1990, the acceptable recovery range was changed to 0.35 to 

1.10. The count time of 720 minutes was used for all samples. A batch blank continued to be 

processed and generally was used in the data analysis unless suspected to have been contaminated 

excessively (a subjective decision). In 1985, the blank method was modified. The value of the blank 

used in the analysis of the result for a sample was the average value of the last 20 valid batch blanks. 

To be valid, a batch blank value was tested using the Dixon outlier test and, if it passed the test, was 

added to the population of the last twenty blanks. In 1988, the blank process was further modified by 

use of the Winsorized trimmed mean of the population of 20 blanks instead of the average value. The 

purpose of these modifications was to minimize the influence of laboratory contamination artifacts, 

which were considered to be nonrandom events that, if incorporated in the blank, would 

inappropriately bias the results of the other samples on the low side. In addition, the reagent blank 

was replaced by a matrix blank, either real or artificial urine. The volume of the sample analyzed 

(aliquant) was 800 ml if the volume of the sample was greater than 800 ml, or the entire sample if the 

volume of the sample was less than 800 ml. The result of the aliquant was divided by the volume 

fraction (800 ml/volume of the sample) if the volume of the sample was ≥800 ml. The efficiency of the 

detectors was typically in the range 0.25 to 0.35.

1993 →

Upgrades to procedures occurred in 1993 in order to achieve a process MDA less than or equal to 

0.020 dpm/sample. Count time was increased to 2,400 minutes. The entire sample was analyzed so 

that the volume fraction was unity for all samples. In addition, a contract was established with a 

commercial bioassay laboratory, with a requirement that a MDA ≤0.020 dpm/sample be achieved. In 

1997, the on-site bioassay laboratory was shut down. 
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Americium (1963 → )

Except for the details of the chemistry, the process for americium was similar to that described for 

plutonium. A solvent extraction process, specific for americium, was first used in 1963. A new 

process (not defined in the data log) was started in November 1965. At some point, not defined in the 

examined data logs, the ion exchange method was implemented for americium.

Enriched Uranium (1952 -1971)

Urine samples were analyzed for enriched uranium according to a process called electroplating. A 

50-ml aliquant of urine was extracted from the 24-hr sample and was chemically processed to 

minimize impurities. The resulting solution was poured into an electrodeposition column, and the 

uranium was deposited on a stainless-steel disk. The disk then was counted for alpha radiation with 

the gas flow proportional counters, as described for plutonium. Counting times used in this period 

were 30, 40, 60, 70, 90, 120, and 150 minutes. 

From 1952 to 1955, one aliquant per sample was used. In 1960, a second aliquant was processed if 

the result of the first aliquant was ≥7 dpm/24-hr sample. If the second result was within a specified 

range of the first result, the average of the two results was recorded and reported. If the second result 

was out of the specified range, a third aliquant was processed, and the average of the two results that 

best confirmed each other was used. If that average was less than the reporting level of 8.8 dpm/24-

hr, the result was recorded and reported as background. From 1961 to 1971, two aliquants routinely 

were processed for each urine sample, with a third aliquant (1961 to 1969) processed if the spread of 

the results of the first set was outside the specified range. The recording and reporting logic was the 

same as that for 1960. From 1964 to 1971, the recording and reporting limit appears to be ≥20 to 28 

dpm/24-hr sample, depending on the volume of the sample. 

Blank data were not used to adjust the sample count rate, except sporadically in 1963 and 1964. 

Detector background usually was subtracted, but not always. Spike samples were processed, 

although it is not obvious how that data were used, if at all. Instead, a constant value of the product of 

the detector efficiency (ε) and the recovery (R) was used: 0.40 (1953 to 1955 and 1971), 0.30 (1960 

to 1970), and 0.24 (1964 to 1970 for detectors with ε = 0.40). 

Enriched uranium operations were phased out at Rocky Flats from 1962 to 1963, although some 

workers were still monitored for enriched uranium intakes through 1971.

Depleted Uranium (1952–1971)

Two methods were used to analyze urine samples for depleted uranium. From 1952 to April 1964, a 

fluorimeter was used to measure the mass (micrograms) of uranium in a 100λ (0.1 ml) aliquant of the 

24-hr urine sample. The result was extrapolated to the total sample and was reported in the unit of 

μg/24-hr sample. A volume adjustment was made if the sample volume was less than 1,000 ml. If 

less than 1,000 ml, the volume was set equal to 1,000 ml. 

Screening was done with one aliquant. A second aliquant was processed if the net reading of the first 

aliquant was greater than or equal to a value in a chart correlated with the volume of the 24-hr urine 

sample. A third aliquant was processed if the results (net readings) of the first two aliquants varied by 

20% or more. The average result of the two aliquants that agreed was converted to μg/24-hr sample 

and reported only if the result was greater than or equal to the reporting level 5.8 μg/24-hr sample. 

Otherwise, the result was reported as background.

After April 1964, the urine sample was analyzed using the electroplating procedure described 

previously for enriched uranium, and the results were reported in the unit of dpm/24-hr sample (or 

background).
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Gross Alpha (1952–1971)

Two methods were used to analyze urine samples for gross alpha counts, from either plutonium or 

uranium. The ether extraction method was used from 1952 to December 12, 1962, and the TBP 

extraction method was used from December 12, 1962, to 1964. The TBP method was replaced by 

the TOPO method. All methods were non-specific in extracting plutonium and uranium, as well as 

americium and natural thorium. 

In all methods, the entire urine sample was processed, and the final extract was evaporated on a 

planchet and counted on the gas flow proportional counter. Counting time was typically 150 minutes, 

although from 1952 to 1955 count times of 55, 60, and 75 minutes, and in 1971 count times of 40 and 

60 minutes, were also used. 

Samples with results ≥0.9 dpm/24-hr sample typically, but not always, were counted using a PHA 

system to determine whether to credit the result to enriched uranium or plutonium, or a portion to 

both. The default condition, through 1963, was to credit the result to enriched uranium unless the 

PHA count indicated otherwise. After 1963 (and enriched uranium operations were phased out), the 

default condition was to credit the result to plutonium. In either case, the results should be considered 

to be upper bounds, because of the non-specificity of the analysis.

Assessment of MDA

The MDA is assessed for plutonium, americium, enriched uranium, depleted uranium, and gross 

alpha, based on equation 5A-1 and the values of parameters for the methods. Some considerations 

are: 

• The probabilities of Type I (false positive) and Type II errors (false negative) are each 5% (α = 

β = 0.05).

• The MDA is assessed for the typical, average, or median condition. Where appropriate, the 

MDA is also assessed for the 5th- or 95th-percentile conditions.

• The MDA is assessed for the methods as they should have been performed, with respect to 

such factors as alpha transmission factor, blank subtraction, recovery fraction, and volume 

adjustment.

• For methods with two or more options in the same period (e.g., evaporation vs. 

electrodeposition, 40% detectors vs. 50% detectors) the option that gives the higher MDA is 

used.

The value of the MDA for the typical, average, or median condition pertains to the process and

indicates the amount or activity in the population of urine samples that would have been detected with 

a 95% probability, given a properly set decision criterion that allows a 5% probability of a Type I error. 

In reality, the decision criterion (and method) at Rocky Flats was not based on the probability of a 

Type I error. Instead, an arbitrary level (10% of the tolerance level or any non-negative value) was 

used as the decision criterion for recording and reporting detected amounts or activities.

The value of the MDA for the 5th- or 95th- percentile conditions pertains to individual samples for 

which the conditions of the sample [e.g., low volume, or conditions of the processing (low recovery, 

high blank, high alpha self-absorption)], were marginal. The conditions of low recovery, low volume, 

and high alpha self-absorption are associated with the calibration factor K and can be incorporated 

either in the value of K or in the value of ∆K. 
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Sample volumes for routine 24-hr urine samples are:

Median: 1,350 ml

5th percentile: 700 ml

95th percentile: 1,750 ml

The values for the parameter values for the processes were obtained through review of the urine data 

logs for the periods from 1952 to 1955 and from 1960 to 1971. For some years in this period, logs for 

only a part of the year were available.

Data for Alpha Counting Systems

The detector background (cpm) for the gas flow proportional counters, based on tabulations in the 

urine data logs from 1952 to 1955 and from 1960 to 1963, for sample count time of 150 minutes, are:

Average 5th percentile Median 95th percentile

1950s 0.060 ± 0.022 0.02 0.06 0.10

1960s 0.054 ± 0.014 0.03 0.05 0.08

Composite 0.056 ± 0.017 0.03 0.05 0.08

No documentation was found concerning the count time used to measure the detector background, 

but the count time is likely to be 150 minutes or longer. For the purpose of assessing the MDA, the 

composite average is used for the value of the detector background count rate, BDet = 0.056 cpm, with 

the standard deviation, sDet = 0.017 cpm, for all alpha counting methods (except for americium) and 

for all sample count times. For americium, the values for the 1960s are used because the americium 

process was not implemented in the 1950s. 

The blank count rate is method specific, and the application of the blank in the data analysis was 

variable between methods and within a method over time. The confounder was the intermittent, but 

persistent, laboratory contamination artifacts introduced into blanks and worker samples. These 

artifacts caused false positives from a worker exposure viewpoint but real positives from a detection 

viewpoint. In practice, high blank values (a subjective decision) were generally ignored, and suspect 

(unexpectedly high) sample results were either confirmed or overruled by recounting, resampling, or 

analyzing another aliquant.

For the purpose of this MDA analysis, the median value of the blank is used to determine the process 

MDA and the 95th-percentile (low to high) value is used to determine the MDA for the more extreme 

conditions. The median and 95th-percentile blank count rates are summarized:

Analyte Period

Blank cpm

Median 95th percentile

Plutonium 1952-1971 0.06 0.28

Enriched uranium 1952-1971 0.05 0.22

Depleted uranium 1964-1971 0.05 0.22

Americium 1963-1971 0.07 0.26

Gross alpha 1952-1971 0.08 0.30

These values are the average of the yearly values extracted from available urine data logs. For each 

of the analytes, the yearly median and 95th-percentile values did not differ enough over the period to 

warrant a separate MDA analysis. The blank values for enriched and depleted uranium are based on

log entries in 1963 and 1964 for cell blank checks for the electrodeposition process.
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The value of the blank count rate, BBlk, is taken from the above table for the given analyte. The 

standard deviation, sBlk, is taken to be the square root of the blank count for the process divided by the 

count time of the process:

sBlk = (1/T )√ (BBlk x T) (5A-5)

The values for B, sB0, sB1, and so in the MDA equation (5A-1) are derived from the detector 

background and blank values:

B = (BDet + BBlk) x T (5A-6)

sB0 = T x √ ( sDet

2 + sBlk2

) (5A-7)

sB1 = √ B (5A-8)

so = √ (sB1

2 + sB02

) (5A-9)

For some analytes (enriched uranium, depleted uranium) and periods, the decision with regard to 

detection was based on the average of two aliquants. In this case, the value of so for the average two 

aliquants is equal to the value of so for one aliquant divided by the square root of two.

The value of ∆B is taken to be zero. This variable may be used to account for high blank values. 

Instead, the effect of high blank values is determined by using the 95th-percentile value of the blank.

The calibration factor K is a combination of the detector efficiency ε, the recovery R, and the volume 

adjustment factor (V/A). Also included is a factor that accounts for absorption of alpha particles in the 

residue of planchets or plates.

Common detectors were used for all alpha-counting methods. The efficiencies of the detectors are:

Period Detector efficiency

1952-1953 0.45

1954-1963 0.5

1964-1971 0.4 and 0.5

For 1964 to 1971, the value of 0.4 is used as the efficiency for the MDA calculation.

The recoveries used in the MDA assessment are taken to be the median recovery and the 

5th-percentile (low to high) value discerned from the spike data for the process:

Analyte Period

Recovery

Median 5th percentile

Plutonium 1952-1962 0.57 0.25

Plutonium 1963-1971 0.67 0.28

Enriched uranium 1952-1971 0.60 0.21

Depleted uranium 1964-1971 0.60 0.21

Americium 1963-1965 0.67 0.29

Americium 1965-1971 0.80 0.26

Gross alpha 1952-1971 0.57 0.24

The recovery values are based on incomplete data sets and involve extrapolations to cover the total 

period. For plutonium from 1952 to 1962, the values are based on data for 1961 and 1962. For 
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plutonium from 1963 to 1971, the values are based on data for 1963 to 1965 and 1969 to 1971. For 

enriched and depleted uranium, recoveries were not calculated for the spiked samples. The median 

value is based on the value used for most of the period. 

The 5th-percentile value is based on the relative standard deviation (0.40) of the average count rate of 

the spiked samples from 1963 to 1966. For americium from 1963 to 1965, the values are based on a 

complete set for that period, ending November 1, 1965. For 1965 to 1971, the values are based on 

data from November 1, 1965, through 1966, and 1968 to 1970. For gross alpha, the values are based 

on data from 1962 to 1969 for the TBP method. In general, values for all the processes are 

remarkably similar, except for americium from 1965 to 1971.

The volume adjustment factor (V/A) is incorporated into the calibration factor K as the reciprocal 

1/(V/A), so it becomes a multiplier with the efficiency and recovery. For convenience, the reciprocal of 

the volume adjustment factor is designated Vf. 

For plutonium, americium, and gross alpha, the median condition is V = A and Vf = 1. The extreme 

condition is a low sample volume normalized to 1,200 ml: V = 1,200 ml, A = 700 ml (the 

5th-percentile volume), and Vf = 0.58. 

For enriched and depleted uranium (for the electrodeposition process), A = 50 ml, the median V = 

1,350 ml, and Vf = 0.037. The extreme condition is a high sample volume: V = 1,750 ml (the 

95th-percentile volume), A = 50 ml, and Vf = 0.029. 

The absorption of the alpha particles in the residue evaporated on the planchets or electrodeposited 

on the plates should be incorporated into the value of the calibration factor. The factor to incorporate 

this effect is the fraction of the alphas emitted by the deposited analyte that successfully escape from 

the residue. Let this factor be designated Fa, where Fa = (1 – fraction of alphas absorbed in the 

residue), and let the fraction of alphas absorbed in the residue be fabs. The values of fabs, based on 

judgments of experienced bioassay chemists, for the extreme (95th-percentile) condition, and the 

corresponding values of Fa are:

Analyte Period

95th percentile

fabs Fa

Plutonium (evaporated) 1952-1962 0.4 0.6

Plutonium (evaporated) 1963-1971 0.3 0.7

Plutonium (electrodeposited) 1963-1971 0.05 0.95

Enriched uranium (electrodeposited) 1952-1971 0.05 0.95

Depleted uranium (electrodeposited) 1964-1971 0.05 0.95

Americium (evaporated) 1964-1971 0.3 0.7

Americium (electrodeposited) 1964-1971 0.05 0.95

Gross alpha (evaporated) 1952-1962 0.1 0.9

Gross alpha (evaporated) 1962-1971 0.3 0.7

From 1963 to 1971, approximately half of the plutonium and americium samples were 

electrodeposited. However, the identities of samples that were electrodeposited are not discernable 

from the databases and reports of urinalysis results that are readily accessible. For the purpose of 

the MDA assessment, use the value of Fa for the evaporation process. 

For the median condition, the value of Fa is taken to be 1, under the assumption that the absorption of 

alphas for the median condition of the planchet or plate was incorporated in the recovery value used 

at the time.
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The calibration factor K is the product of ε, R, Vf, and Fa: 

K = ε R Vf Fa (5A-10)

The values of ∆B and ∆K are considered to be zero.

Data for Fluorometric Mass Measurements

Applying the MDA equation (5A-1) for fluorometric mass measurements involves setting the value of T 

to unity and eliminating the term “3”.

The value of sB0 is the standard deviation of the blank flux readings that are subtracted for the signal 

of the aliquant reading. The value of sB1 is set equal to sB0, and so is equal to the value of sB0 

multiplied by the square root of 2:

so = sB0 √ 2 (5A-11)

The value of sB0 was determined from a review of the urine data logs for 1955 and 1960 to 1962. One 

discontinuity was noted on September 14, 1955. The value of sB0 before the discontinuity was 0.37 

and, after the discontinuity, averaged 0.15.

The calibration factor K converts the fluorimeter net reading to the unit of μg uranium/24-hr sample 

(see equation 5A-4). In 1955, the calibration factor was applied to the uncorrected net reading. In 

1960 and following, the calibration factor was applied to the corrected reading, which was the net 

reading multiplied by the factor 1.15. The factor of 1.15 is incorporated into the value of K starting in 

1960. For the 1950s, the calibration factor for 1955 is used:

Period K

1952-1959 75/V

1960-1964 87/V

For the median condition, the volume V is equal to 1,350 ml. For the extreme condition, the 

95th-percentile volume of 1,750 ml is used.

The values of ∆B and ∆K are considered to be zero.

MDA Values

The value of the MDA is presented to two significant figures, for information purposes. In most cases, 

the value of the MDA should be considered only to one significant figure.

Plutonium

The MDA for plutonium is assessed for the median condition and for the extreme (5th- or 

95-percentile) condition for the blank, the recovery, the volume factor Vf, and the alpha transmission 

factor Fa, individually and in combination. A count time of 150 minutes is used for all assessments.

For the median MDA (dpm/24-hr sample):

Period

Values of the variables

so ε R Vf Fa MDA

1952-1953 5.74 0.45 0.57 1.0 1.0 0.57

1954-1962 5.74 0.50 0.57 1.0 1.0 0.51

1963 5.74 0.50 0.67 1.0 1.0 0.44

1964-1971 5.74 0.40 0.67 1.0 1.0 0.54
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The values of the variables for the extreme (5th- or 95-percentile) conditions and the resulting MDA 

(dpm/24-hr sample) for all of the extreme conditions occurring for the same sample are: 

Period

Values of the variables

so ε R Vf Fa MDA

1952-1953 7.98 0.45 0.25 0.58 0.6 5.0

1954-1962 7.98 0.50 0.25 0.58 0.6 4.5

1963 7.98 0.50 0.28 0.58 0.7 3.4

1964-1971 7.98 0.40 0.28 0.58 0.7 4.3

The value of so incorporates the 95th-percentile value of the blank.

It is unlikely that the four extreme conditions (high blank, low recovery, low volume, and cruddy 

residue on the planchet) all occurred for the same sample. The following tables present the MDA for 

each of the extreme conditions individually, as well as for combinations of two and three extreme 

conditions.

MDA (dpm/24-hr sample) for one extreme condition

Period so R Vf Fa

1952-1953 0.76 1.3 0.98 0.95

1954-1962 0.68 1.2 0.88 0.85

1963 0.58 1.0 0.75 0.62

1964-1971 0.73 1.3 0.94 0.78

MDA (dpm/24-hr sample) for two extreme conditions

Period so, R so, Vf so, Fa R, Vf R, Fa Vf, Fa

1952-1953 1.7 1.3 1.3 2.2 2.2 1.6

1954-1962 1.6 1.2 1.1 2.0 2.0 1.5

1963 1.4 1.0 0.97 1.8 1.5 1.1

1964-1971 1.7 1.3 1.2 2.3 1.9 1.3

MDA (dpm/24-hr sample) for three extreme conditions

Period so, R, Vf so, R, Fa so, Vf, Fa R, Vf, Fa

1952-1953 3.0 2.9 2.2 3.7

1954-1962 2.7 2.6 2.0 3.4

1963 2.4 2.0 1.4 2.6

1964-1971 3.0 2.5 1.8 3.2

Enriched Uranium

The MDA for enriched uranium is assessed for the median condition and for the extreme (5th- or 

95th-percentile) condition for the blank, the recovery, the volume factor Vf, and the alpha transmission 

factor Fa, individually and in combination. A count time of 150 minutes is used for MDA assessments 

from 1952 to 1963. In the period 1964-1969, the count time of 30 minutes is used, and, for 1970 to 

1971, the count time of 40 minutes is used. For 1952 to 1959, the value of so is calculated for one 

aliquant, and, for 1960 to 1971, the value of so is calculated based on the average of two aliquants.
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For the median MDA (dpm/24-hr sample):

Period

Values of the variables

so ε R Vf Fa MDA

1952-1953 5.45 0.45 0.60 0.037 1.0 14

1954-1959 5.45 0.50 0.60 0.037 1.0 13

1960-1963 3.85 0.50 0.60 0.037 1.0 9.4

1964-1969 1.57 0.40 0.60 0.037 1.0 31

1970-1971 1.83 0.40 0.60 0.037 1.0 25

The values of the variables for the extreme (5th- or 95th-percentile) conditions and the resulting MDA 

(dpm/24-hr sample) for all of the extreme conditions occurring for the same sample are: 

Period

Values of the variables

so ε R Vf Fa MDA

1952-1953 6.72 0.45 0.21 0.029 0.95 64

1954-1959 6.72 0.50 0.21 0.029 0.95 58

1960-1963 4.75 0.50 0.21 0.029 0.95 43

1964-1969 2.18 0.40 0.21 0.029 0.95 150

1970-1971 2.48 0.40 0.21 0.029 0.95 120

The value of so incorporates the 95th-percentile value of the blank.

It is unlikely that the four extreme conditions (high blank, low recovery, low volume, and cruddy 

residue on the planchet) all occurred for the same sample. The following tables present the MDA for 

each of the extreme conditions individually, as well as for combinations of two and three extreme 

conditions.

MDA (dpm/24-hr sample) for one extreme condition

Period so R Vf Fa

1952-1953 17 40 18 15

1954-1959 15 36 16 13

1960-1963 11 27 12 9.9

1964-1969 38 88 39 32

1970-1971 31 74 32 27

MDA (dpm/24-hr sample) for two extreme conditions

Period so, R so, Vf so, Fa R, Vf R, Fa Vf, Fa

1952-1953 48 21 18 51 42 19

1954-1959 43 19 16 46 38 17

1960-1963 32 14 12 34 28 13

1964-1969 110 49 40 110 92 41

1970-1971 90 40 33 93 76 34

MDA (dpm/24-hr sample) for three extreme conditions

Period so, R, Vf so, R, Fa so, Vf, Fa R, Vf, Fa

1952-1953 61 50 23 54

1954-1959 55 45 20 48

1960-1963 41 34 15 43

1964-1969 140 120 51 150

1970-1971 120 94 42 120
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Depleted Uranium

The MDA for depleted uranium is assessed for two processes: fluorimetric mass measurements from 

1952 to April 30, 1964; and electrodeposition/alpha counting measurements from May 1, 1964, to 

1971.

For the fluorimetric mass measurements, the MDA is assessed for one aliquant, because the decision 

for detection was based on one aliquant, even though quantification was based on the average of two 

aliquants. The MDA at the extreme condition is based on the 95th-percentile volume.

Period sB0

K MDA (μg/24-hr sample)

Median Extreme Median Extreme

1952-1955 0.37 0.056 0.043 31 40

1955-1959 0.15 0.056 0.043 12 16

1960-1964 0.15 0.064 0.050 11 14

For the electrodeposition/alpha counting measurements, the MDA values tabulated for enriched 

uranium for 1964 to 1971 apply also to depleted uranium.

Americium

The MDA for americium is assessed for the median condition and for the extreme (5th- or 

95th-percentile) condition for the blank, the recovery, the volume factor Vf, and the alpha transmission 

factor Fa, individually and in combination. A count time of 150 minutes is used for assessments from 

1963 to 1970. In 1971, the typical (and minimum) count time is 60 minutes.

Values of the MDA are presented to two significant figures.

For the median MDA (dpm/24-hr sample):

Period

Values of the variables

so ε R Vf Fa MDA

1963 5.82 0.50 0.67 1.0 1.0 0.44

1964-1965 5.82 0.40 0.67 1.0 1.0 0.55

1965-1970 5.82 0.40 0.80 1.0 1.0 0.46

1971 3.51 0.40 0.80 1.0 1.0 0.76

The values of the variables for the extreme (5th- or 95th-percentile) conditions and the resulting MDA 

(dpm/24-hr sample) for all of the extreme conditions occurring for the same sample are: 

Period

Values of the variables

so ε R Vf Fa MDA

1963 9.95 0.50 0.26 0.58 0.7 4.3

1964-1965 9.95 0.40 0.26 0.58 0.7 5.4

1965-1970 9.95 0.40 0.26 0.58 0.7 5.4

1971 5.94 0.40 0.26 0.58 0.7 8.9

The value of so incorporates the 95th-percentile value of the blank.

It is unlikely that the four extreme conditions (high blank, low recovery, low volume, and cruddy 

residue on the planchet) all occurred for the same sample. The following tables present the MDA for 

each of the extreme conditions individually, as well as for combinations of two and three extreme 

conditions.
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MDA (dpm/24-hr sample) for one extreme condition

Period so R Vf Fa

1963 0.68 1.1 0.76 0.63

1964-1965 0.86 1.4 0.95 0.79

1965-1970 0.72 1.4 0.80 0.66

1971 1.2 2.3 1.3 1.1

MDA (dpm/24-hr sample) for two extreme conditions

Period so, R so, Vf so, Fa R, Vf R, Fa Vf, Fa

1963 1.8 1.2 0.98 2.0 1.6 1.1

1964-1965 2.2 1.5 1.2 2.4 2.0 1.4

1965-1970 2.2 1.2 1.0 2.4 2.0 1.1

1971 3.6 2.0 1.7 4.0 3.3 1.9

MDA (dpm/24-hr sample) for three extreme conditions

Period so, R, Vf so, R, Fa so, Vf, Fa R, Vf, Fa

1963 3.0 2.5 1.7 2.8

1964-1965 3.8 3.2 2.1 3.5

1965-1970 3.8 3.2 1.8 3.5

1971 6.2 5.2 2.9 5.7

Gross Alpha

The MDA for gross alpha measurements is assessed for the median condition and for the extreme 

(5th- or 95th-percentile) condition for the blank, the recovery, the volume factor Vf, and the alpha 

transmission factor Fa, individually and in combination. A count time of 55 minutes is used for 1952, 

75 minutes for 1953 to 1959, and 150 minutes for 1960 to 1971 for assessments of the MDA for both 

the median and extreme conditions, except for 1971, when a count time of 40 minutes is also used for 

the extreme condition.

For the median MDA (dpm/24-hr sample):

Period

Values of the variables

so ε R Vf Fa MDA

1952 3.26 0.45 0.57 1.0 1.0 1.0

1953 4.23 0.45 0.57 1.0 1.0 0.88

1954-1959 4.23 0.50 0.57 1.0 1.0 0.79

1960-1963 6.23 0.50 0.57 1.0 1.0 0.55

1964-1971 6.23 0.40 0.57 1.0 1.0 0.69

The values of the variables for the extreme (5th- or 95th-percentile) conditions and the resulting MDA 

(dpm/24-hr sample) for all of the extreme conditions occurring for the same sample are: 

Period

Values of the variables

so ε R Vf Fa MDA

1952 6.09 0.45 0.24 0.58 0.9 7.4

1953 7.12 0.45 0.24 0.58 0.9 6.2

1954-1959 7.12 0.50 0.24 0.58 0.9 5.6

1960-1962 10.27 0.50 0.24 0.58 0.9 3.9

1963 10.27 0.50 0.24 0.58 0.7 5.0

1964-1971 10.27 0.40 0.24 0.58 0.7 6.3

1971 (T=40minutes) 5.18 0.40 0.24 0.58 0.7 13

The value of so incorporates the 95th-percentile value of the blank.
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It is unlikely that the four extreme conditions (high blank, low recovery, low volume, and cruddy 

residue on the planchet) all occurred for the same sample. The following tables present the MDA for 

each of the extreme conditions individually, as well as for combinations of two and three extreme 

conditions.

MDA (dpm/24-hr sample) for one extreme condition

Period so R Vf Fa

1952 1.6 2.5 1.8 1.2

1953 1.4 2.1 1.5 0.98

1954-1959 1.2 1.9 1.4 0.88

1960-1962 0.86 1.3 0.95 0.61

1963 0.86 1.3 0.95 0.79

1964-1971 1.1 1.6 1.2 0.98

1971 (T=40minutes) 2.2 3.4 2.4 2.0

MDA (dpm/24-hr sample) for two extreme conditions

Period so, R so, Vf so, Fa R, Vf R, Fa Vf, Fa

1952 3.9 2.8 1.8 4.3 2.8 2.0

1953 3.3 2.4 1.5 3.6 2.3 1.7

1954-1959 2.9 2.1 1.4 3.2 2.1 1.5

1960-1962 2.0 1.5 0.96 2.3 1.5 1.1

1963 2.0 1.5 1.2 2.3 1.9 1.4

1964-1971 2.6 1.9 1.5 2.8 2.3 1.7

1971 (T=40minutes) 5.2 3.8 3.1 5.8 4.8 3.5

MDA (dpm/24-hr sample) for three extreme conditions

Period so, R, Vf so, R, Fa so, Vf, Fa R, Vf, Fa

1952 6.7 4.3 3.1 4.7

1953 5.6 3.6 2.6 4.0

1954-1959 5.1 3.3 2.4 3.6

1960-1962 3.5 2.3 1.6 2.5

1963 3.5 2.9 2.1 3.1

1964-1971 4.4 3.6 2.6 4.0

1971 (T=40minutes) 9.0 7.5 5.4 8.3

Discussion

The MDA is an a priori concept that may be applied a posteriori to a sample under certain 

circumstances: that the parameter values for the sample (e.g., volume, recovery, detector efficiency, 

count time) are or can be known prior to the processing of the sample result, and that the information 

is used, conceptually, to determine the sub-population of conditions of which that sample is a 

member. Then the a priori MDA value for that subpopulation can be assigned to that sample. The 

sample volume, the characteristics of the detector that is used to count the sample, and the count 

time are all known before the analysis of the sample measurement. In theory, but generally not in 

practice, the recovery could also be known before the analysis of the sample measurement.

The MDA values presented in this report represent an overall process MDA for the median and 

extreme conditions. However, sufficient information is presented to allow the determination of the 

MDA for a specific sample if the sample-specific parameter values were known. The sample-specific 

parameter values, except for recovery, are generally recorded in the urine data logs, but not all of the 

urine data logs have been located and some may not have been archived.
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The recoveries for 1952 to 1971 were determined by batch spikes. Not until 1973 were some 

plutonium samples spiked with an internal tracer (first 236Pu and, later, 242Pu). All plutonium samples 

were spiked with an internal tracer after 1978. Experience has shown that a significant variability of 

recovery can exist within a batch of samples. Therefore, the recovery of a batch spike does not 

necessarily indicate the recovery of each sample in the batch.

Whether to use the median value of the MDA or the extreme value depends on the purpose. By 

definition, the median value implies that half of the samples will have a sample-specific MDA that is 

lower, and half higher. If the purpose is to define a sample-specific conservative bound, the MDA for 

the extreme condition should be considered. In general, the recovery is the variable that had the most 

influence on the sample-specific MDA.

Acknowledgments

The author would like to acknowledge the invaluable insights into the historical bioassay methods and 

procedures at Rocky Flats provided by Dale L. Bokowski, the lead bioassay chemist at Rocky Flats 

from 1961 to 1992. The author also acknowledges the valuable assistance of Michael Prochownik, 

Nancy Chandler, and Kimberly Linnemeyer for their assistance in locating and retrieving archived 

urinalysis data logs so that this analysis could be achieved.
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REFERENCE

HPS (Health Physics Society), 1996, An American National Standard – Performance Criteria for 

Radiobioassay, HPS N13.30-1996.
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ATTACHMENT B 

MINIMUM DETECTABLE ACTIVITY FOR IN VIVO LUNG COUNTS AT ROCKY FLATS

TABLE OF CONTENTS

B.1 Introduction...............................................................................................................................56

B.2 MDA Methodology .................................................................................................................... 56

B.3 History of Counting Systems and Procedures ........................................................................... 57

B.4 Assessment of MDA .................................................................................................................60

B.5 Values of the Variables, 1969 → for NaI(Tl) and Phoswich Detector Systems .......................... 62

B.6 Values of the Variables, 1976 → for Ortec Germanium Detector Systems................................ 63

B.7 Values of the Variables, 1978 → for PGT I Germanium Detector Systems ............................... 64

B.8 Values of the Variables, 1979 → for PGT II Germanium Detector Systems .............................. 65

B.9 Values of the Variables, 1985 → for PGT Organ Pipe Germanium Detector Systems............... 66

B.10 Values of the Variables, 1985 → for EG&G Organ Pipe Germanium Detector Systems............ 66

B.11 MDA for Rocky Flats Plutonium ................................................................................................ 67

References ........................................................................................................................................69
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B.1 INTRODUCTION

In vivo lung counts have been performed at Rocky Flats since 1964 to determine the activity of 

plutonium in the lungs of workers who were exposed, or had the potential to be exposed, to airborne 

plutonium. The method of in vivo lung counts was to place one or more detectors over the chest of 

the subject and count the photons emitted from the plutonium mixture, if any, in the subject’s chest. 

Plutonium was not detected directly because of the low abundance of gamma photons and because 

of the severe attenuation of the more abundant L X-rays. Instead, the 59.5-keV gamma photon from 

241Am was used to detect 241Am, which is present to some extent in all weapons-grade plutonium at 

Rocky Flats. The activity of plutonium was then calculated from the detected 241Am by measuring, 

calculating, or assuming the fraction of the 241Am in the plutonium mixture on the date of the lung 

count. At Rocky Flats, the fraction of the 241Am in the plutonium mixture has historically been 

characterized in terms of parts per million by weight. Direct in vivo measurement of plutonium in the 

lungs, although investigated, was never implemented at Rocky Flats. The Rocky Flats lung counter 

detected 241Am. The assessment of the MDA, therefore, is focused on the MDA for 241Am. The MDA 

for plutonium can then be derived from the 241Am MDA based on the value of the ppm 241Am for the 

plutonium mixture.

B.2 MDA METHODOLOGY

The general equation for the MDA is Equation 6 in the American National Standard, Performance 

Criteria for Radiobioassay (HPS 1996):

MDA = (1 + ∆K) (2∆BB + 2kso + 3) / KT (5B-1)

where:

B = the total count of the appropriate blank

so = the standard deviation in the net sample count of a subject with no additional analyte

so = √ [sB1

2 + (1/m2

) sB0

2

] (5B-2)

where:

sB1 = the standard deviation of the subject, where the subject contains no actual 

analyte above that of the appropriate blank

sB0 = the standard deviation in the unadjusted count of the appropriate blank

m = the adjustment factor for the appropriate blank

K = calibration factor

∆K = the maximum fractional systemic error bound in the calibration factor K

∆B = the maximum expected fractional systemic error bound in the appropriate blank

k = the abscissa of the standardized normal distribution corresponding to the 0.05 probability level 

(for α = 0.05 and β = 0.05, k = 1.645)

T = the standard subject counting time for the procedure

Applying this equation to in vivo lung counting at Rocky Flats involves determining the value of each 

of these variables for the counting systems and procedures used at Rocky Flats as the systems and 

procedures evolved. The MDA for in vivo measurements is necessarily individually specific because 

the detectability of 241Am in the chest is a significant function of the chest wall thickness (CWT) of the 

subject. 
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The MDA may also be determined empirically from replicate measurements on an appropriate blank. 

This approach is used for the current system at Rocky Flats.

B.3 HISTORY OF COUNTING SYSTEMS AND PROCEDURES

The in vivo lung counting systems at Rocky Flats consisted of photon detectors mounted in a shielded 

room (6 inches thick low-background steel lined with layers of lead, tin, and zinc) with electronic 

equipment (amplifiers and multi-channel analyzers) to process and record the data.

There were three counting rooms: 

• Room A, built in 1964, operational in 1965

• Room B, built in 1968, operational in 1969

• Room C, built in 1975, operational in 1976

Each room was equipped with a detector system. When a new detector system was implemented, 

the previous system was usually maintained as a standby, backup system. As a result, end dates for 

a given detector system are not known. In the era of the Ge detector systems, two or more detector 

systems could have been operational simultaneously. In that type of situation, the detector system is 

identified in the record for each lung count.

1964–1968

There was one counting room. The detector system consisted of two NaI(Tl) scintillation detectors 

(there was a third detector used for cesium and potassium measurements), each detector was round 

with a diameter of 4 inches and was 4 millimeters thick with a surface area of 80 square centimeters. 

These detectors were known as the 4x4 detectors. In most situations, the detectors were configured 

with one detector above the left portion of the upper chest and the second detector was over the liver 

and gut region. The chest detector was sometimes placed over the right portion of the upper chest 

instead of the left position. In other cases, both detectors were placed over the chest. The chest 

detector(s) was placed in a framework called a jig to allow a standard and reproducible position for all 

subjects. Count time was either 40 minutes live time (MLT) or 20 MLT. Two backgrounds were used: 

1) room background and 2) matched subject background. The room background was the count rate 

in the empty counting room measured at the start of the day. The matched subject background was 

the count rate of an unexposed subject with matched 137Cs and 40K count rates. Calibration was 

based on 241Am impregnated epoxy lungs in the chest cavity of a water-filled REMCAL phantom, 

manufactured by Alderson Research Laboratories, Inc. No adjustment was made for CWT.

1969–1976 

During this period, two counting rooms were operational with three 4x4 NaI(Tl) scintillation detectors, 

two over the upper chest (right and left portions) and one over the liver/gut region (the liver/gut 

detector was eliminated in 1974). 

Changes:

1. The ROI of the 59.5-keV photopeak of 241Am was expanded.

2. The use of the jig for positioning the detectors was discontinued. Instead, the detectors were 

positioned in light contact with the surface of the chest.

3. The standard count time was changed to 2,000 seconds (1,000 seconds for expedited counts).
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4. The method of the matched subject background based on 137Cs and 40K was replaced by the 

index method. 

The index method, developed by Robert Bistline, had the following features:

1. Subjects were characterized by an index I equal to the ratio of the subject’s weight (W, 

pounds) divided by twice the subject’s height (H, inches). 

2. A population of at least 20 known cold (unexposed) subjects of a diversity of indices was 

counted to generate a data set of net count rate versus index.

3. A curve fit to the data set generated a prediction equation with the index as the variable.

4. The subject’s index was used to determine the predicted net count rate for the subject. 

This approach was done separately for the right chest, the left chest, and the liver/gut.

In 1973, a phoswich detector system was implemented and used intermittently into the 1980’s. The 

NaI(Tl) layer of the phoswich detectors was dimensionally the same as the 4x4 detectors.

This system lacked the stability of the NaI(Tl) detector system and was used mainly as a backup 

system. Use of the phoswich system to detect plutonium directly via the L X-rays was not 

successfully implemented at Rocky Flats.

In about 1972, room background was measured at the start of the day shift, at noon, and at the start 

of the night shift. The value of the room background used was the five-point moving average of the 

last five counts.

Starting in 1969, the ppm 241Am was measured from a representative sample of the plutonium mixture 

associated with incidents with the potential for inhalation exposure of workers. This situation was 

called a PI (for possible inhalation) and refers both to the incident and to the worker involved in the 

incident.

In this period, the use of a lithium-drifted Ge detector system was investigated but was never 

implemented.

1976–1985

This period is the era of the high-purity Ge detector array systems. Three counting rooms were 

operational. When the Ge systems were implemented, most, if not all, quantitative measures were 

accomplished with that system. The NaI(Tl) and phoswich systems were used only as screening 

systems, and later only as backup systems. The Ge systems in this period featured four detectors 

mounted in an array attached to a single cryostat containing liquid nitrogen. The system had two of 

these arrays, one positioned over the upper right chest and the other over the upper left chest. A full 

system consisted of eight detectors. However, occasionally one or more of the detectors failed and 

were electronically eliminated from the system. A minimum system was five detectors, three in the 

right array and two in the left array. To maintain a minimum functional system, a hybrid system 

consisting of two arrays of different characteristics was frequently used.

The Ge system implementation timeline was:

1976 Ortec detectors, 10 cm2 per detector, two arrays
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1977 Princeton Gamma Tech (PGT I) detectors, 15 cm2 per detector, two arrays

1979 First array, Princeton Gamma Tech (PGT II) detectors, 18 cm2 per detector

1980 Second array, Princeton Gamma Tech (PGT II) detectors, and 18 cm2 per detector

Other changes in this period were:

1. The calibration factor for the Ge systems was adjusted for the CWT of the subject. The 

thickness (cm) was equal to twice the index value minus 0.1 (CWT = 2 I – 0.1). 

2. Calibration was accomplished using a masonite phantom from 1976 to 1978.

3. Calibration was accomplished using the Lawrence Livermore Laboratory (LLL) phantom from 

1979 to present.

4. The method of determining the background changed for the Ge systems. Room and subject 

background were determined, as a unit, from the subject’s own spectrum using a ROI in the 

range of 65 to 72 keV.

1985–1995

In this period, Ge detectors in the organ pipe configuration were implemented. Instead of clustering 

four detectors in an array with a common cryostat, each detector was attached to its own cryostat, 

which was tall and slender. The detectors with the cryostats were then clustered in arrays, two to four 

detectors per array, over the right and left portions of the upper chest. If a detector malfunctioned, it 

was physically replaced with a backup, functional detector. A minimum system from 1985 to 1991 

was five detectors, three detectors on the right and two on the left. The full system was seven 

detectors, four detectors on the right and three on the left, although the routine system generally 

consisted of six detectors, either four on the right and two on the left or three on each side. In 1991, 

the full system was six detectors with either four detectors on the right and two on the left or three 

detectors on each side.

The Ge system implementation timeline was:

1985 Princeton Gamma Tech organ pipe detectors, 20 cm2 per detector 

1991 EG&E Ortec organ pipe detectors, 20 cm2 per detector

No other significant changes were made during this period.

1995–Present

In May 1995 the lung counter hardware, software, and detectors were upgraded. The data acquisition 

and analysis were accomplished using the Canberra Industries program ABACOS-Plus. Instead of 

the ROI approach used historically, this program used a peak-search method to detect activity of a 

radionuclide. The value of the MDA was established by replicate measurements on an appropriate 

blank. The Ge detectors were replaced by EG&E Ortec organ pipe detectors with 38 cm2 per 

detector. The standard system was four detectors, two on each side. The minimum system was 

three detectors, two on the right and one on the left.

Another significant change is the equation to determine the CWT. ABACOS-Plus incorporates the 

equation developed by Anderson (1983) at Lawrence Livermore National Laboratory:

CWT (cm) = 1.973 (W/H) – 2.0038 (5B-3)
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where:

W = subject’s weight (pounds), and

H = subject’s height (inches).

The effect of this change is an adjustment factor, given by the equation:

CWT Adjustment Factor = 0.5364 exp(0.635 I) (5B-4)

This adjustment factor is a multiplier to the activity of 241Am, detected via the 59.5-keV gamma, for all 

previous detector systems at Rocky Flats. Equation 5B-4 may also be applied as a divisor to 

historical calibration factors for previous systems at Rocky Flats.

B.4 ASSESSMENT OF MDA

The value of the MDA for 241Am is assessed for each detector system and for each significant change 

in the procedure. It is assessed not only for the typical Rocky Flats male (I = 1.35, CWT = 3.3 cm) but 

also for a reasonable range of statures (I = 0.90, CWT = 1.5 cm and I = 1.80, CWT = 5.1 cm). The 

assessment is also done for the minimally configured system as well as for the standard system and 

for half of the normal count time (for expedited lung counts) as well as the full count time.

Discontinuities, which were significant changes in methods affecting the interpretation of the raw data 

(and hence the MDA), were identified and were incorporated into the value of the calibration factor. 

This process was done starting with the most recent calibration method, assumed to be the most 

accurate. The factors associated with each discontinuity were then applied, as divisors to the 

calibration factor, back through the history of the systems. Alternatively, the product of the factors, for 

the appropriate period, may be used in place of the term (1 + ∆K) in Equation 5B-1. The discontinuity 

factors are:

Year Discontinuity Factor

1995 New CWT method

Index = 0.90 0.95

Index = 1.35 1.26

Index = 1.80 1.68

1979 Calibration using LLL phantom 1.30

1969 Fixed positioning discontinued

ROI for 59.5-keV photopeak increased

1.45

The discontinuity factors for the CWT can be calculated by any index I using:

CWT Discontinuity Factor = 0.5364 exp(0.635 I) (5B-5)

Values of the Variables, 1964–1968

The minimum system was one NaI(Tl) detector positioned over the left chest.

Count time T = 20 MLT or 40 MLT

The appropriate blank, B, was the net man background (after room background was subtracted) 

estimated from matched, unexposed subjects based on 137Cs and 40K measurements.

B = 600 for T = 20 MLT

B = 1,200 for T = 40 MLT
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∆B = 0.2, estimated as the upper bound for this method.

The value of so is calculated from counting statistics, including the total subject count, which is taken 

as the sum of B and the room background R.

R = 500 for T = 20 MLT

R = 1,000 for T = 40 MLT

Because the decision of detection was based on the net subject count rate (after subtraction of room 

background) compared with the predicted net count rate of the appropriate blank, the calculation of 

sB1 includes an extra component of the room background.

sB1

2

 = Total subject count + R = B + 2R 

= 1,600 for T = 20 MLT

= 3,200 for T = 40 MLT

sB0

2

 = B 

= 600 for T = 20 MLT

= 1,200 for T = 40 MLT

m = 1 

so = 44.9 for T = 20 MLT

= 66.3 for T = 40 MLT

The 241Am calibration factor K for two detectors, normalized to the calibration with the Lawrence LLL 

phantom and incorporating the discontinuity factors, equation 5B-5, 1.30, and 1.45, is given by:

K = 55.13 exp(-0.2359 (2 I – 0.1))/exp(0.635 I) (5B-6)

The calibration factor for the system with only one detector over the left portion of the chest is given 

by Equation 5B-6 multiplied by 0.43. This factor is the fraction of the total activity in the calibration 

lungs of the Rocky Flats LLL phantom that is in left portion of the lung. The MDA, therefore, pertains 

to the activity in the total lung based on the detection of activity only in the left portion of the lung.

K = 8.96 for I = 0.90

K = 5.45 for I = 1.35

K = 3.31 for I = 1.80

Because K is normalized to the calibration with the LLL phantom and the discontinuity factors are 

incorporated into K, the value of ∆K is taken to be 0. Since the term (1 + ∆K) in Equation 5B-1 is a 

multiplier to the MDA and since the value of ∆K is estimated based on the professional judgment of the 

analyst, one can easily adjust the values of the MDA in this paper if another analyst has a different 

judgment.

For the standard system of two detectors, over both the right and left portions of the lungs, the counts 

are basically doubled and the values of the variables are:

B = 1,200 for T = 20 MLT

B = 2,400 for T = 40 MLT

∆B = 0.2, estimated as the upper bound for this method

R = 1,000 for T = 20 MLT

R = 2,000 for T = 40 MLT

sB1

2

 = Total subject count + R = B + 2R 
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= 3,200 for T = 20 MLT

= 6,400 for T = 40 MLT

sB0

2

 = B 

= 1,200 for T = 20 MLT

= 2,400 for T = 40 MLT

m = 1 

so = 69.3 for T = 20 MLT

= 93.8 for T = 40 MLT

K = 20.85 for I = 0.90

= 12.67 for I = 1.35

= 7.70 for I = 1.80

B.5 VALUES OF THE VARIABLES, 1969 → FOR NAI(TL) AND PHOSWICH DETECTOR 

SYSTEMS

The standard system was two detectors positioned over the left and right portions of the chest. This is 

also the minimum system.

Count time T = 1,000 seconds or 2,000 seconds

The appropriate blank was the net man background (after room background was subtracted) 

estimated from matched, unexposed subjects based on the subject’s index:

B = 1,100 for T = 1,000 seconds

B = 2,200 for T = 2,000 seconds

∆B = 0 for the NaI(Tl) detector system

∆B = 0.1, estimated for the phoswich detector system, because the system was less stable 

than the NaI(Tl) detector system.

The value of so is calculated from counting statistics, including the total subject count, which is taken 

as the sum of B and the room background R. The value of sB0 is taken to be 10% of the value B, 

based on the typical relative standard deviation of the predicted subject net count rate.

R = 833 for T = 1,000 seconds

R = 1,667 for T = 2,000 seconds

sB1

2

 = Total subject count + R = B + 2R 

= 2,767 for T = 1,000 seconds

= 5,533 for T = 2,000 seconds

sB0

2

 = (0.1B)2

= 12,100 for T = 1,000 seconds

= 48,400 for T = 2,000 seconds

m = 1 

so = 121.9 for T = 1,000 seconds

= 232.2 for T = 2,000 seconds

The 241Am calibration factor K for two detectors, normalized to the calibration with the LLL phantom 

and incorporating the discontinuity factors, Equation 5B-5 and 1.30, is given by:

K = 79.94 exp(-0.2359 (2 I – 0.1))/exp(0.635 I) (5B-7)

K = 30.23 for I = 0.90
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= 18.37 for I = 1.35

= 11.16 for I = 1.80

B.6 VALUES OF THE VARIABLES, 1976 → FOR ORTEC GERMANIUM DETECTOR 

SYSTEMS

The standard system was two arrays, each array with four detectors, positioned over the left and right 

portions of the chest. The minimum system was two arrays with a total of eight detectors.

Count time T = 1,000 seconds or 2,000 seconds

The appropriate blank was the count in the subject’s spectrum (composite for all detectors) in the 

range of 65 keV to 72 keV, divided by eight. The subject, in essence, was his own blank with 

essentially no bias. Room background was no longer assessed separately for Ge systems.

∆B = 0 

m = 8 

For eight detectors:

B = 341 for T = 1,000 seconds (unadjusted by m = 8)

B = 682 for T = 2,000 seconds (unadjusted by m = 8)

For five detectors:

B = 213 for T = 1,000 seconds (unadjusted by m = 8)

B = 427 for T = 2,000 seconds (unadjusted by m = 8)

For the calculation of sB1, the subject background is B/8.

For eight detectors:

T = 1,000 seconds:

sB1 = 6.53 sB0 = 18.5 s0 = 6.93

T = 2,000 seconds:

sB1 = 9.23 sB0 = 26.1 s0 = 9.79

For five detectors:

T = 1,000 seconds:

sB1 = 5.17 sB0 = 14.6 s0 = 5.48

T = 2,000 seconds:

sB1 = 7.30 sB0 = 20.7 s0 = 7.75

The 241Am calibration factor K for two arrays with a total of eight detectors, based on the calibration 

with the LLL phantom and incorporating the discontinuity factors, Equation 5B-5 and 1.30 (for pre1979 systems), is given by:

K = 24.12 exp(-0.3398 (2 I – 0.1))/exp(0.635 I) (5B-8)

and, for Ortec systems 1979 and following:
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K = 31.36 exp(-0.3398 (2 I – 0.1))/exp(0.635 I) (5B-9)

For the minimum system of five detectors, adjust the calibration factor by multiplying by (5/8).

Index

Eight-detector calibration factor (K)

Pre-1979 1979 →

0.90 7.64 9.94

1.35 4.23 5.50

1.80 2.34 3.04

B.7 VALUES OF THE VARIABLES, 1978 → FOR PGT I GERMANIUM DETECTOR 

SYSTEMS

The PGT I Ge system is basically the same as the Ortec Ge system. The main difference is that the 

PGT I detectors had a larger surface area but a poorer resolution. 

∆B = 0 

m = 4

For eight detectors: 

B = 240 for T = 1,000 seconds (unadjusted by m = 4)

B = 480 for T = 2,000 seconds (unadjusted by m = 4)

For five detectors:

B = 150 for T = 1,000 seconds (unadjusted by m = 4)

B = 300 for T = 2,000 seconds (unadjusted by m = 4)

For the calculation of sB1, the subject background is B/4.

For eight detectors:

T = 1,000 seconds:

sB1 = 7.75 sB0 = 15.5 s0 = 8.67

T = 2,000 seconds:

sB1 = 10.95 sB0 = 21.9 s0 = 12.2 

For five detectors:

T = 1,000 seconds:

sB1 = 6.12 sB0 = 12.2 s0 = 6.84

T = 2,000 seconds:

sB1 = 8.66 sB0 = 17.3 s0 = 9.68

The 241Am calibration factor K for two arrays with a total of eight detectors, based on the calibration 

with the LLL phantom and incorporating the discontinuity factors, Equation 5B-5 and 1.30 (for pre1979 systems), is given by:

K = 34.09 exp(-0.3292 (2 I – 0.1))/exp(0.635 I) (5B-10)

and, for PGT I systems 1979 and following:
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K = 44.318 exp(-0.3292 (2 I – 0.1))/exp(0.635 I) (5B-11)

For the minimum system of five detectors, adjust the calibration factor by multiplying by (5/8).

Index

Eight-detector calibration factor (K)

Pre-1979 1979 →

0.90 11.00 14.30

1.35 6.15 7.99

1.80 3.43 4.46

B.8 VALUES OF THE VARIABLES, 1979 → FOR PGT II GERMANIUM DETECTOR 

SYSTEMS

The PGT II Ge system is basically the same as the Ortec and PGT I systems. The main difference is 

that the PGT II detectors, again, had a larger surface area but a poorer resolution. 

∆B = 0 

m = 4

For eight detectors:

B = 273 for T = 1,000 seconds (unadjusted by m = 4)

B = 546 for T = 2,000 seconds (unadjusted by m = 4)

For five detectors:

B = 170 for T = 1,000 seconds (unadjusted by m = 4)

B = 341 for T = 2,000 seconds (unadjusted by m = 4)

For the calculation of sB1, the subject background is B/4.

For eight detectors:

T = 1,000 seconds:

sB1 = 8.26 sB0 = 16.5 s0 = 9.23

T = 2,000 seconds:

sB1 = 11.7 sB0 = 23.4 s0 = 13.1

For five detectors:

T = 1,000 seconds:

sB1 = 6.53 sB0 = 13.1 s0 = 7.31

T = 2,000 seconds:

sB1 = 9.23 sB0 = 18.5 s0 = 10.3

The 241Am calibration factor K for two arrays with a total of eight detectors, incorporating Equation 5B5, is given by:

K = 38.65 exp(-0.3579 (2 I – 0.1))/exp(0.635 I) (5B-12)

For the minimum system of five detectors, adjust the calibration factor by multiplying by (5/8).
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Index

Eight-detector 

calibration factor (K)

0.90 11.88

1.35 6.47

1.80 3.52

B.9 VALUES OF THE VARIABLES, 1985 → FOR PGT ORGAN PIPE GERMANIUM 

DETECTOR SYSTEMS

The PGT organ pipe Ge system is basically the same as the previous Ge array systems. The main 

difference is the ability to maintain a stable, standard configuration with six detectors. 

∆B = 0 

m = 4

T = 1,000 seconds T = 2,000 seconds

B: 215 429

sB1: 7.33 10.4

sB0: 14.7 20.7

s0: 8.20 11.6

The 241Am calibration factor K for two arrays with a total of six detectors, incorporating equation 5B-5, 

is given by:

K = 34.32 exp(-0.2946 (2 I – 0.1)) / exp(0.635 I) (5B-13)

Index

Six-detector 

calibration factor (K)

0.90 11.74

1.35 6.77

1.80 3.90

B.10 VALUES OF THE VARIABLES, 1985 → FOR EG&G ORGAN PIPE GERMANIUM 

DETECTOR SYSTEMS

The EG&G organ pipe Ge system is basically the same as the previous PGT organ pipe Ge array 

system. 

∆B = 0 

m = 4

T = 1,000 seconds T = 2,000 seconds

B: 204 408

sB1: 7.14 10.1

sB0: 14.3 20.2

s0: 7.98 11.3

The 241Am calibration factor K for two arrays with a total of six detectors, incorporating equation 5B-5, 

is given by:

K = 42.36 exp(-0.3708 (2 I – 0.1)) / exp(0.635 I) (5B-14)
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Index

Six-detector 

calibration factor (K)

0.90 12.73

1.35 6.85

1.80 3.69

Values of the Variables, 1995 → 

The MDA for the current system at Rocky Flats was not determined analytically using equation 5B-1. 

Instead, the MDA was determined empirically from replicate measurements on an appropriate blank 

simulating the counts of the average Rocky Flats worker (CWT = 3.36 cm). Therefore, there are no 

values of the variables to be listed here. The value of the MDA for the average Rocky Flats worker 

(CWT = 3.36 cm, I = 1.35) is 0.3 nCi 241Am. 

To extrapolate this value to the range of workers (CWT = 1.15 cm, I = 0.90 to CWT = 5.10 cm, I = 

1.80), the following approach was used to establish the calibration factor equation as a function of 

CWT:

The efficiency equation is:

ε = a1 exp(a2 CWT) (5B-15)

where

ε = count per minute per gamma from 241Am

a1 = 0.045 (factor determined from calibration)

a2 = -0.41 (factor determined from calibration)

The efficiency equation converts to the style of historical calibration equations, using the conversion 

factors of 0.359 gamma photons (59.5 keV) per 241Am nuclear transformation, and 797 gammas per 

minute per nCi 241Am. The derived calibration equation is:

K = 35.9 exp(-0.41 CWT) (5B-16)

The MDA for any value of CWT is then obtained from the product of 0.3 nCi (the MDA for the average 

Rocky Flats worker) and the ratio (9.05/K for the value of the CWT).

B.11 MDA FOR ROCKY FLATS PLUTONIUM

The MDA for Rocky Flats plutonium is derived from the MDA of 241Am, based on the value of the ppm 

241Am in the plutonium mixture at the time of the lung count. To convert the MDA for 241Am to the 

MDA for plutonium (239Pu and 240Pu), the MDA for 241Am is multiplied by the factor:

MDA Conversion Factor = (1x106

 – ppm 241Am)/(48.2 ppm 241Am) (5B-17)

The MDA conversion factors for some typical values of ppm 241Am are:

ppm 241Am

MDA conversion 

factors

100 207

1,000 20.7

10,000 2.05
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The task is to determine the value of the ppm 241Am at the time of the lung count. The practice at 

Rocky Flats was to measure the ppm 241Am in a representative sample of material involved in a 

possible inhalation incident. If a representative sample was not obtained or the origin of the intake 

was not known, a default value of 1,000 ppm 241Am was used and was assigned to the date of the 

intake or to the date of the first positive lung count, if the date of the intake was not known. For 

subsequent lung counts, the value of the ppm 241Am was updated to account for the ingrowth of the 

241Am from the nuclear transformation of 241Pu and for the radioactive decay of the 241Am. The rate of 

ingrowth of 241Am in the plutonium mixture depends on the fraction by weight of the 241Pu in the 

mixture. The initial weight fraction of 241Pu in Rocky Flats plutonium is taken to be 0.0050 in the 

1950s and 1960s and 0.0036 in the 1970s and later. The following table presents values of the ppm 

241Am at times (years) after the intake for initial values of ppm 241Am of 100, 1,000, and 10,000. The 

value of 100 ppm 241Am may be taken as the lower bound and represents freshly purified plutonium.

Years

241Am ppm at Time (years) after Intake

Initial fraction 241Pu = 0.0036 Initial fraction 241Pu = 0.0050

100 1,000 10,000 100 1,000 10,000

1 270 1,200 10,200 340 1,200 10,200

2 430 1,300 10,300 560 1,500 10,400

4 730 1,600 10,600 980 1,900 10,800

6 1,000 1,900 10,800 1,400 2,200 11,100

10 1,500 2,400 11,200 2,000 2,900 11,700

20 2,300 3,200 11,900 3,100 4,000 12,700

30 2,800 3,600 12,200 3,800 4,700 13,200

40 3,000 3,900 12,300 4,200 5,000 13,500

50 3,200 4,000 12,300 4,400 5,200 13,500

The appropriate value of the ppm 241Am should be applied for lung counts that occurred following a 

known or assumed intake.

The equation to calculate the ppm 241Am for any time (years) after the intake is:

A = L1 P0 [exp(-λPu 241T) – exp(-λAm241T)] + 106

 A0 L2 / [A0 L2 + exp(-λPu239T)] (5B-18)

where

A = ppm 241Am at time T (years)

L1 = λPu241 / (λAm241 - λPu241) 

λPu241 = decay constant for 241Pu (half-life = 14.4 years) = 0.0481

λAm241 = decay constant for 241Am (half-life = 433 years) = 0.00160

A0 = initial ppm 241Am

P0 = initial 241Pu ppm = (Initial 241Pu fraction by weight) × (106

 - A0) 

L2 = exp(-λAm241T) / (106

 - A0 ) 

λPu239 = decay constant for 239Pu (half-life = 24,100 years) = 0.0000288

Half-times are from Table of Isotopes, Seventh Edition (Lederer and Shirley 1978).
Effective Date: 01/12/2004 Revision No. 00 Document No. ORAUT-TKBS-0011-5 Page 69 of 103

REFERENCES

HPS (Health Physics Society), 1996, An American National Standard – Performance Criteria for 

Radiobioassay, HPS N13.30-1996.

Lederer, C.M. and Shirley, V.S. (eds.), 1978, Table of Isotopes, Seventh Edition, Lawrence Berkeley 

Laboratory, University of California-Berkeley, Berkeley, California, and John Wiley & Sons, 

New York, New York.
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SUMMARY OF MDA FOR IN VIVO LUNG COUNTS AT ROCKY FLATS

Period Detector system Index

MDA (nCi) for 241Am

Comments

Minimum system Standard system

Half time Full time Half time Full time

1964 - 1968 NaI(Tl) 4x4 0.90 1.7 1.5 1.3 1.2 Full time = 40 minutes live time (MLT).

Minimum system is one detector over the left chest.

Standard system is two detectors, over right and left chests.

1.35 2.8 2.5 2.1 1.9

1.80 4.6 4.1 3.5 3.2

1969 - NaI(Tl) 4x4 0.90 0.80 0.76 Full time = 2000 seconds.

1.35 1.3 1.3 Standard system is two detectors, over right and left chests.

1.80 2.2 2.0

1973 - Phoswich 0.90 1.2 1.2 Full time = 2000 seconds.

Standard system is two detectors, over right and left chests.

NaI sensitive layer is the same as the NaI 4x4.

1.35 2.0 2.0

1.80 3.3 3.2

1976 - 1978 Ortec Arrays

(High-purity Ge)

0.90 0.26 0.18 0.20 0.14 Full time = 2000 seconds.

Standard system is 8 detectors in 2 arrays.

Minimum system is 5 detectors in 2 arrays.

1.35 0.48 0.32 0.37 0.25

1.80 0.86 0.59 0.66 0.45

1979 - Ortec Arrays

(High-purity Ge)

0.90 0.20 0.14 0.16 0.11 Same as previous Ge system.

1.35 0.37 0.25 0.28 0.19

1.80 0.66 0.45 0.51 0.35

1978 - PGT I Arrays

(High-purity Ge)

0.90 0.22 0.15 0.17 0.12 Same as previous Ge systems.

1.35 0.40 0.27 0.31 0.21

1.80 0.71 0.49 0.55 0.38

1979 - PGT I Arrays

(High-purity Ge)

0.90 0.17 0.12 0.13 0.09 Same as previous Ge systems.

1.35 0.31 0.21 0.24 0.16

1.80 0.55 0.38 0.42 0.29

1979 - PGT II Arrays

(High-purity Ge)

0.90 0.22 0.15 0.17 0.12 Same as previous Ge systems.

1.35 0.40 0.28 0.31 0.21

1.80 0.74 0.50 0.57 0.39

1985 - PGT Organ Pipe 

(OP) Ge 

Detectors

0.90 0.15 0.11 Standard system = 6 detectors.

Standard count time = 2000 seconds.

Occasionally, 5 or 7 detectors were used.

1.35 0.26 0.18

1.80 0.46 0.32

1991 - EG&G Organ 

Pipe Ge 

Detectors

0.90 0.14 0.10 Standard system = 6 detectors.

1.35 0.26 0.18 Standard count time = 2000 seconds.

1.80 0.48 0.33

1995 - Ortec 2 Organ 

Pipe Ge 

Detectors

0.90 0.14 Standard system = 4 detectors.

1.35 0.3 Standard count time = 2000 seconds.

1.80 0.6

Note: Starting in 1978, hybrid Ge systems were used that combined two differing Ge arrays or detector types. For hybrid systems, use the higher of the MDA values for the detector 

types involved.

The index is the ratio of the weight (pounds) of the subject divided by twice the height (inches) and is correlated with the CWT. The index of 1.35 represents the typical Rocky Flats 

male subject, with a reasonable range of 0.90 (CWT = 1.5 cm) to 1.80 (CWT = 5.1 cm).

To convert the MDA for 241Am to the MDA for Pu-239,240, multiply the MDA for 241Am by [(1x106

 – ppm 241Am)/(48.2 x ppm 241Am)], where ppm 241Am is the parts per million by weight 

of the 241Am in the plutonium mixture at the time of the lung count. 
Effective Date 01/12/2004 Revision No. 00 Document No. ORAUT-TKBS-0011-5 Page 71 of 103

SUMMARY OF MDA FOR IN VIVO LUNG COUNTS AT ROCKY FLATS (Cont’d.)

241Am grows into the plutonium mixture from the nuclear transformation of Pu241. The initial weight fraction of Pu241 in Rocky Flats plutonium is 0.0050 in the 1950’s and 1960’s and 

0.0036 in the 1970’s and 1980’s. For freshly purified plutonium, with a residual of approximately 100 ppm 241Am, the ppm 241Am would be 270 to 340 after the first year, 430 to 560 

after the second year, 730 to 980 after the 4th year, 1000 to 1400 after the 6th year, 1500 to 2000 after the 10th year, 2300 to 3100 after the 20th year, 2800 to 3800 after the 30th year, 

3000 to 4200 after the 40th year, and 3200 to 4400 after the 50th year. 

Half time was usually used for non-scheduled counts or when a large number of subjects needed to be counted expeditiously.
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THE ATTACHMENT C 

EXAMPLES OF BIOASSAY DATA RECORDS AND REPORTS

TABLE OF CONTENTS

Figure

C-1 Urinalysis Record Card and Health Sciences Data System – Urinalysis Detail (1) 

(first activity date on the HSDS portion: 10-29-54) ................................................................

Page

73

C-2 Urinalysis Record Card and Health Sciences Data System – Urinalysis Detail (2) 

(first activity date on the HSDS portion 8-19-53) .................................................................... 74

C-3 Urinalysis Record Card and Health Sciences Data System – Urinalysis Detail (3) 

(first activity date on the HSDS portion: 1-6-58) .................................................................... 75

C-4 Health Sciences Data System – Urinalysis Detail (1) (first activity date 9-17-58).................... 76

C-5 Health Sciences Data System – Urinalysis Detail (2) (first activity date 3-19-73).................... 77

C-6 Analytical Report – Bioassay Analysis Data 3-15-93.............................................................. 78

C-7 Analytical Report – Bioassay Analysis Data 10-28-93 ............................................................ 79

C-8 Form 1 – Sample Results 1-29-96.......................................................................................... 80

C-9 Rocky Flats Environmental Technology Site (1) 8-27-96 (analytes: U238, U235, 

U234).....................................................................................................................................81

C-10 Rocky Flats Environmental Technology Site (1) 8-8-96 (analyte: Pu239).............................. 82

C-11 Form 1 – Sample Results – Quanterra, Richland 7-31-98 ...................................................... 83

C-12 General Engineering Laboratories, Inc. 6-28-99..................................................................... 84

C-13 Health Sciences Urinalysis Record (with tritium, fecal and nasal smear results) .................... 85

C-14 Health Physics – Body Counter Information 12-8-65.............................................................. 86

C-15 Health Physics – Body Counter Information 5-16-68.............................................................. 87

C-16 Health Physics – Body Counter Information 8-26-68.............................................................. 88

C-17 Health Physics – Body Counter Information 9-16-70.............................................................. 89

C-18 Radiation Dosimetry – Body Count Results 10-3-74............................................................... 90

C-19 Radiation Dosimetry – Body Count Results 5-30-75............................................................... 91

C-20 Radiation Dosimetry – Body Count Results 1-9-78................................................................. 92

C-21 Body Counter Results 12-8-81 ............................................................................................... 93

C-22 Radiation Dose Assessment – Body Count Results 7-22-83.................................................. 94

C-23 Radiation Dose Assessment – Body Count Results 5-18-83.................................................. 95

C-24 Radiation Dose Assessment – Body Count Results 2-21-84.................................................. 96

C-25 Radiation Dose Assessment – Body Count Results 3-22-84.................................................. 97

C-26 Radiation Dose Assessment – Body Count Results 10-10-85 ................................................ 98

C-27 Radiation Dose Assessment – Body Count Results 3-6-89.................................................... 99

C-28 Internal Dosimetry – Lung Count Results 11-23-93.............................................................. 100

C-29 ABACOS-Plus 3-6-96........................................................................................................... 101

C-30 ABACOS-Plus 11-15-01....................................................................................................... 102

C-31 ABACOS-Plus 6-14-01......................................................................................................... 103
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Figure C-1. Urinalysis Record Card and Health Sciences Data System – Urinalysis Detail (1) (first 

activity date on the HSDS portion: 10-29-54). 
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Figure C-2. Urinalysis Record Card and Health Sciences Data System – Urinalysis Detail (2) (first 

activity date on the HSDS portion 8-19-53). 
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Figure C-3. Urinalysis Record Card and Health Sciences Data System – Urinalysis Detail (3) (first 

activity date on the HSDS portion: 1-6-58). 
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Figure C-4. Health Sciences Data System – Urinalysis Detail (1) (first activity date 9-17-58). 
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Figure C-5. Health Sciences Data System – Urinalysis Detail (2) (first activity date 3-19-73). 
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Figure C-6. Analytical Report – Bioassay Analysis Data 3-15-93. 
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Figure C-7. Analytical Report – Bioassay Analysis Data 10-28-93. 
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Figure C-8. Form 1 – Sample Results 1-29-96. 
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Figure C-9. Rocky Flats Environmental Technology Site (1) 8-27-96 (analytes: U238, U235, 

U234). 
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Figure C-10. Rocky Flats Environmental Technology Site (1) 8-8-96 (analyte: Pu239). 
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Figure C-11. Form 1 – Sample Results – Quanterra, Richland 7-31-98. 
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Figure C-12. General Engineering Laboratories, Inc. 6-28-99. 
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Figure C-13. Health Sciences Urinalysis Record (with tritium, fecal and nasal smear results). 
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Figure C-14. Health Physics – Body Counter Information 12-8-65. 
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Figure C-15. Health Physics – Body Counter Information 5-16-68. 
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Figure C-16. Health Physics – Body Counter Information 8-26-68. 
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Figure C-17. Health Physics – Body Counter Information 9-16-70. 
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Figure C-18. Radiation Dosimetry – Body Count Results 10-3-74. 
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Figure C-19. Radiation Dosimetry – Body Count Results 5-30-75. 
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Figure C-20. Radiation Dosimetry – Body Count Results 1-9-78. 
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Figure C-21. Body Counter Results 12-8-81. 
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Figure C-22. Radiation Dose Assessment – Body Count Results 7-22-83. 
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Figure C-23. Radiation Dose Assessment – Body Count Results 5-18-83. 
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Figure C-24. Radiation Dose Assessment – Body Count Results 2-21-84. 
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Figure C-25. Radiation Dose Assessment – Body Count Results 3-22-84. 
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Figure C-26. Radiation Dose Assessment – Body Count Results 10-10-85. 
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Figure C-27. Radiation Dose Assessment – Body Count Results 3-6-89. 
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Figure C-28. Internal Dosimetry – Lung Count Results 11-23-93. 
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Figure C-29. ABACOS-Plus 3-6-96. 
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Figure C-30. ABACOS-Plus 11-15-01. 
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Figure C-31. ABACOS-Plus 6-14-01.