ORAU TEAM

Dose Reconstruction

Project for NIOSH

 

Oak Ridge Associated Universities I Dade Moeller & Associates I MJW Corporation

Page 1 of 122

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Document Title:

Rocky Flats Plant – Occupational Internal Dose

Document Number: ORAUT-TKBS-0011-5 

Revision: 01 

Effective Date: 02/01/2007

Type of Document: TBD

Supersedes: Revision 00 

Subject Expert(s): Roger B. Falk

Site Expert(s): N/A

Approval: Signature on File Approval Date: 01/25/2007

Robert Meyer, Document Owner

Approval: Signature on File Approval Date: 01/19/2007

John M. Byrne, Task 3 Manager

Concurrence: Signature on File Concurrence Date: 01/18/2007

Edward F. Maher, Task 5 Manager

Concurrence: Signature on File Concurrence Date: 01/25/2007

James P. Griffin, Deputy Project Director

Approval: Brant A. Ulsh Signature on File for Approval Date: 02/01/2007

James W. Neton, Associate Director for Science
Document No. ORAUT-TKBS-0011-5 Revision No. 01 Effective Date: 02/01/2007 Page 2 of 123

PUBLICATION RECORD

EFFECTIVE 

DATE

REVISION 

NUMBER DESCRIPTION

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.

12/13/2005 00 PC-1 Approved page change revision 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.

02/01/2007 01 Approved Revision 01 revised to reflect current introductory material. 

Revised in response to Union comments. Change made to Table 5-

9. Constitutes a total rewrite of document. This revision addresses 

Worker Outreach comments as described in CT-0201, CT-0206, and 

CT-0207. Revised to incorporate attribution, per ORAU request. The 

Worker Outreach comments from the June 23, 2004, meeting of the 

United Steelworkers of America Local 8031 and Rocky Flats Security 

Officers Local Union 1 are addressed in Section 5.3.2 and 5.4.2 

regarding internal organ counts; Sections 5.2.1.2, 5.2.2.2, 5.2.3.2., 

and 5.2.4.2 regarding particle size and that as many reports as 

possible have been reviewed in the writing of this section. The 

Worker Outreach comments from the June 23, 2004, meeting of 

Colorado State Building and Construction Trades regarding wholebody counting are addressed in Sections 5.3.2 and 5.2.4. 

Incorporates internal and NIOSH formal review comments. This 

revision results no change to the assigned dose and no PER is 

required. Training required: As determined by Task Manager. 

Initiated by Robert Meyer.
Document No. ORAUT-TKBS-0011-5 Revision No. 01 Effective Date: 02/01/2007 Page 3 of 123

TABLE OF CONTENTS

SECTION TITLE

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

PAGE

7

5.1 Introduction................................................................................................................................9

5.1.1 Purpose......................................................................................................................10

5.1.2 Scope.........................................................................................................................10

5.2 Source Term............................................................................................................................10

5.2.1 Plutonium ...................................................................................................................10

5.2.1.1 Isotopic Composition ..................................................................................... 10

5.2.1.2 Plutonium Solubility and Particle Size............................................................ 11

5.2.2 Americium ..................................................................................................................12

5.2.2.1 Isotopic Composition ..................................................................................... 12

5.2.2.2 Americium Solubility and Particle Size........................................................... 12

5.2.3 Enriched Uranium.......................................................................................................12

5.2.3.1 Isotopic Composition ..................................................................................... 12

5.2.3.2 Enriched Uranium Solubility and Particle Size ............................................... 13

5.2.4 Depleted Uranium.......................................................................................................13

5.2.4.1 Isotopic Composition ..................................................................................... 13

5.2.4.2 Depleted Uranium Solubility and Particle Size ............................................... 13

5.3 Bioassay Data .........................................................................................................................13

5.3.1 Urinalysis Data ...........................................................................................................14

5.3.1.1 Plutonium Urinalysis ...................................................................................... 14

5.3.1.2 Americium Urinalysis ..................................................................................... 16

5.3.1.3 Enriched Uranium Urinalysis ......................................................................... 17

5.3.1.4 Depleted Uranium Urinalysis ......................................................................... 18

5.3.1.5 Gross Alpha Urinalysis .................................................................................. 20

5.3.1.6 Tritium ........................................................................................................... 20

5.3.2 Lung Count Data ........................................................................................................21

5.3.2.1 Americium/Plutonium..................................................................................... 21

5.3.2.2 Thorium/Depleted Uranium............................................................................ 24

5.3.3 Other Bioassay Data ..................................................................................................25

5.3.3.1 Wound Count Data........................................................................................ 25

5.3.3.2 Nasal Smears and Fecal Samples................................................................. 26

5.4 Records and Reports............................................................................................................... 26

5.4.1 Urinalysis Records and Reports .................................................................................27

5.4.1.1 Interpretation of the Urinalysis Record Card .................................................. 27

5.4.1.2 Interpretation of the Health Sciences Data System – Urinalysis 

Detail Report ................................................................................................. 27

5.4.1.3 Interpretation of Other Urinalysis Reports...................................................... 28

5.4.2 Lung Count Records and Reports ..............................................................................30

5.5 Attributions and Annotations.................................................................................................... 33

References ........................................................................................................................................48
Document No. ORAUT-TKBS-0011-5 Revision No. 01 Effective Date: 02/01/2007 Page 4 of 123

Glossary ............................................................................................................................................51

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

A.1 Introduction..............................................................................................................................54

A.2 MDA Methodology...................................................................................................................54

A.3 History of Methods...................................................................................................................55

A.4 Assessment of MDA................................................................................................................61

A.5 MDA Values ............................................................................................................................66

A.6 Discussion ...............................................................................................................................73

A.7 Acknowledgments ...................................................................................................................73

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

B.1 Introduction .............................................................................................................................75

B.2 MDA Methodology...................................................................................................................75

B.3 History of Counting Systems and Procedures.......................................................................... 76

B.4 Assessment of MDA ................................................................................................................79

B.5 Values of the Variables, 1969 – for NaI(Tl) and Phoswich Detector Systems .......................... 81

B.6 Values of the Variables, 1976 – for Ortec Germanium Detector Systems ................................ 82

B.7 Values of the Variables, 1978 – for PGT I Germanium Detector Systems ............................... 84

B.8 Values of the Variables, 1979 – for PGT II Germanium Detector Systems .............................. 85

B.9 Values of the Variables, 1985 – for PGT Organ Pipe Germanium Detector Systems............... 86

B.10 Values of the Variables, 1985 – for EG&G Organ Pipe Germanium Detector Systems............ 87

B.11 MDA for RFP Plutonium .......................................................................................................... 88

Attachment C, Examples of Bioassay Data Records and Reports .....................................................92
Document No. ORAUT-TKBS-0011-5 Revision No. 01 Effective Date: 02/01/2007 Page 5 of 123

LIST OF TABLES

TABLE TITLE

5-1 Weight percent and fraction of alpha activity for weapons-grade plutonium ...........................

PAGE

11

5-2 Weight percent and fraction of alpha activity for ZPPR plutonium .......................................... 11

5-3 Weight percent and fraction of alpha activity for enriched uranium......................................... 12

5-4 Weight percent and fraction of alpha activity for depleted uranium......................................... 13

5-5 Median MDA values for plutonium.......................................................................................... 16

5-6 Median MDA values for americium......................................................................................... 17

5-7 Median MDAs for enriched uranium ....................................................................................... 18

5-8 Median MDAs for depleted uranium from 1952 to April 1964 ................................................. 19

5-9 Median MDAs for depleted uranium from May 1964 to the present ........................................ 19

5-10 Median MDAs for gross alpha measurements........................................................................ 20

5-11 Summary of MDAs for 241Am.................................................................................................. 23

A-1 Method codes ........................................................................................................................55

A-2 Correlation of method code and analyte................................................................................. 56

A-3 Values of tolerance and reporting levels ................................................................................. 56

A-4 Sample volumes for routine 24-hour urine samples................................................................ 61

A-5 Detector background for gas flow proportional counters......................................................... 62

A-6 Median and 95th-percentile blank count rates........................................................................ 62

A-7 Efficiencies of alpha counting detectors ................................................................................. 63

A-8 Recoveries used in MDA assessments .................................................................................. 64

A-9 Fraction of alphas absorbed in residue................................................................................... 65

A-10 Gross alpha calibration factor................................................................................................. 66

A-11 Values of variables and MDA for plutonium for median conditions ......................................... 66

A-12 Values of variables and MDA for plutonium for extreme conditions ........................................ 66

A-13 MDA for plutonium for one, two or three extreme conditions .................................................. 67

A-14 Values of variables and MDA for EU for median conditions.................................................... 67

A-15 Values of variables and MDA for EU for extreme conditions .................................................. 68

A-16 MDA for EU for one, two, or three extreme conditions............................................................ 68

A-17 Values of variables and MDA for fluorimetric measures of DU for median and extreme 

conditions...............................................................................................................................69

A-18 Values of variables and MDA for americium for median conditions ........................................ 69

A-19 Values of variables and MDA for americium for extreme conditions ....................................... 70

A-20 Values of the MDA for americium for one, two, or three extreme conditions........................... 70

A-21 Values of variables and MDA for gross alpha measurements for median conditions .............. 71

A-22 Values of variables and MDA for gross alpha measurements for extreme conditions............. 71

A-23 Values of the MDA for gross alpha measurements for one, two, or three extreme 

conditions...............................................................................................................................71

B-1 Discontinuity factors...............................................................................................................80

B-2 Calibration factors for the Ortec germanium detector system ................................................. 83

B-3 Calibration factors for the PGT I germanium detector system ................................................ 85

B-4 Calibration factors for the PGT II germanium detector system ............................................... 86

B-5 Values of variables for the PGT organ pipe germanium detector system ............................... 86

B-6 Calibration factors for the PGT organ pipe germanium detector system................................. 87

B-7 Values of variables for the EG&G organ pipe germanium detector system ............................ 87

B-8 Calibration factors for the EG&G organ pipe germanium detector system.............................. 87

B-9 MDA conversion factors for values of ppm 241Am ................................................................... 88

B-10 Americium ingrowth in RFP plutonium.................................................................................... 89

B-11 Americium MDA for in vivo lung counts at RFP...................................................................... 90
Document No. ORAUT-TKBS-0011-5 Revision No. 01 Effective Date: 02/01/2007 Page 6 of 123

LIST OF FIGURES

FIGURE TITLE

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

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

PAGE

93

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

activity date on the HSDS portion 8-19-53) ............................................................................94

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

activity date on the HSDS portion: 1-6-58) ............................................................................95

C-4 Health Sciences Data System – Urinalysis Detail (1) (first activity date 9-17-58).................... 96

C-5 Health Sciences Data System – Urinalysis Detail (2) (first activity date 3-19-73).................... 97

C-6 Analytical Report – Bioassay Analysis Data 3-15-93..............................................................98

C-7 Analytical Report – Bioassay Analysis Data 10-28-93 ............................................................99

C-8 Form 1 – Sample Results 1-29-96........................................................................................100

C-9 Rocky Flats Environmental Technology Site (1) 8-27-96 (analytes: 238U, 235U, 234U)............ 101

C-10 Rocky Flats Environmental Technology Site (1) 8-8-96 (analyte: 239Pu) .............................. 102

C-11 Form 1 – Sample Results – Quanterra, Richland 7-31-98 .................................................... 103

C-12 General Engineering Laboratories, Inc. 6-28-99................................................................... 104

C-13 Health Sciences Urinalysis Record (with tritium, fecal, and nasal smear results) ................. 105

C-14 Health Physics – Body Counter Information 12-8-65............................................................ 106

C-15 Health Physics – Body Counter Information 5-16-68............................................................ 107

C-16 Health Physics – Body Counter Information 8-26-68............................................................ 108

C-17 Health Physics – Body Counter Information 9-16-70............................................................ 109

C-18 Radiation Dosimetry – Body Count Results 10-3-74............................................................. 110

C-19 Radiation Dosimetry – Body Count Results 5-30-75............................................................. 111

C-20 Radiation Dosimetry – Body Count Results 1-9-78............................................................... 112

C-21 Body Counter Results 12-8-81 .............................................................................................113

C-22 Radiation Dose Assessment – Body Count Results 7-22-83................................................ 114

C-23 Radiation Dose Assessment – Body Count Results 5-18-83................................................ 115

C-24 Radiation Dose Assessment – Body Count Results 2-21-84................................................ 116

C-25 Radiation Dose Assessment – Body Count Results 3-22-84................................................ 117

C-26 Radiation Dose Assessment – Body Count Results 10-10-85 .............................................. 118

C-27 Radiation Dose Assessment – Body Count Results 3-6-89.................................................. 119

C-28 Internal Dosimetry – Lung Count Results 11-23-93.............................................................. 120

C-29 ABACOS-Plus 3-6-96...........................................................................................................121

C-30 ABACOS-Plus 11-15-01.......................................................................................................122

C-31 ABACOS-Plus 6-14-01.........................................................................................................123
Document No. ORAUT-TKBS-0011-5 Revision No. 01 Effective Date: 02/01/2007 Page 7 of 123

ACRONYMS AND ABBREVIATIONS

AMAD activity median aerodynamic diameter

cm centimeter

cpm counts per minute

CWT chest wall thickness

d day

DOE U.S. Department of Energy

dpm disintegrations per minute

DQO data quality objective

DTPA diethylenetriaminepentaacetate

DU depleted uranium

EDTA ethylenediaminetetraacetate

EEOICPA Energy Employees Occupational Illness Compensation Program Act of 2000

EU enriched uranium

g gram

hr hour

HSDS Health Sciences Data System

ICRP International Commission on Radiological Protection

IMBA Integrated Modules for Bioassay Analysis (computer program)

in. inch

keV kilovolts-electron, 1,000 electron volts

L liter

LLNL Lawrence Livermore National Laboratory

MDA minimum detectable activity

mg milligram

min minute

mL milliliter

MLT minutes live time

mm millimeter

mo month

MPLB maximum permissible lung burden

nCi nanocurie

NIOSH National Institute for Occupational Safety and Health

pCi picocurie

PGT Princeton Gamma Tech

PHA pulse height analysis

ppm parts per million

POC probability of causation
Document No. ORAUT-TKBS-0011-5 Revision No. 01 Effective Date: 02/01/2007 Page 8 of 123

QA Quality Assurance

RCG radioactive concentration guide

RFP Rocky Flats Plant

RFETS Rocky Flats Environmental Technology Site

ROI region of interest

s second

TBD technical basis document

TBP tributyl phosphate

TOPO trioxyl phosphene oxide

TTA thenoyl trifluro acetone

U.S.C. United States Code

WG weapons-grade

yr year

ZPPR Zero Power Physics Reactor

μCi microcurie

μg microgram

μm micrometer

§ section or sections
Document No. ORAUT-TKBS-0011-5 Revision No. 01 Effective Date: 02/01/2007 Page 9 of 123

5.1 INTRODUCTION

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

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

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

dose reconstructions for particular sites or categories of sites. The documents will be revised in the 

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

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

reconstruction.

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

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

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

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

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

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

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

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

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

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

facility encompassed by EEOICPA.

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

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

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

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

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

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

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

determined in accordance with the POC [probability of causation1

As noted above, the statute includes a definition of a DOE facility that excludes “buildings, structures, 

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

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

While this definition contains an exclusion with respect to the Naval Nuclear Propulsion Program, the 

section of EEOICPA that deals with the compensation decision for covered employees with cancer 

[i.e., 42 U.S.C. § 7384n(b), entitled “Exposure in the Performance of Duty”] does not contain such an 

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

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

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

external dosimetry monitoring results are considered valid for use in dose reconstruction. No efforts 

are made to determine the eligibility of any fraction of total measured exposure for inclusion in dose 

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

derived:

] guidelines established under 

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

guidelines (nor the dose reconstruction regulation) define “performance of duty” for DOE employees 

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

 

1 The U.S. Department of Labor is ultimately responsible under the EEOICPA for determining the POC. 
Document No. ORAUT-TKBS-0011-5 Revision No. 01 Effective Date: 02/01/2007 Page 10 of 123

• Radiation from naturally occurring radon present in conventional structures

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

5.1.1

This TBD discusses Rocky Flats Plant (RFP) internal dosimetry data for dose reconstruction and 

includes guidance for the appropriate use of that information. 

Purpose

5.1.2

Workers at RFP had the potential to receive intakes of plutonium, americium, enriched uranium (EU), 

depleted uranium (DU), and tritium, as well as miscellaneous other radionuclides (Daugherty et al. 

2001). Section 5.2 describes the available source term information including isotopic composition, 

solubility, and particle size. Site-specific internal dosimetry information for other radionuclides such as 

thorium, curium, and neptunium, is rare or not available. 

Scope

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) [1]. 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.

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

justification, or clarification of the associated information, are presented in Section 5.5.

5.2 SOURCE TERM

5.2.1

5.2.1.1 Is otopic Compos ition

Plutonium

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

1. The percent by weight of 241Pu, which is needed to calculate the ingrowth of 241Am for the lung 

count data, 

2. The fraction of the activity for each alpha-emitting plutonium isotope, which is needed 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, which is 

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

For weapons-grade (WG) plutonium, which was present at RFP 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 5-1 lists the weight percent and fraction of 

alpha activity for each isotope. 
Document No. ORAUT-TKBS-0011-5 Revision No. 01 Effective Date: 02/01/2007 Page 11 of 123

The Zero Power Physics Reactor (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 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 [2].

Table 5-1. Weight percent and fraction of alpha activity 

for weapons-grade plutonium.

a

Isotope Weight percent

Fraction of 

alpha activity

Pu-238 0.01 0.023

Pu-239 93.79 0.8

Pu-240 5.8 0.18

Pu-241 0.36b –

Pu-242 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 RFP plutonium from July 1976 to July 

1, 1978. This isotopic composition is also typical of plutonium 

metal processed at RFP to 1990 (James 1990).

b. The percent by weight of 241Pu for 1959 to 1977 was 0.49, with a 

range of 0.35 to 0.65 (RFETS 2002, p. 5.1).

Table 5-2. Weight percent and fraction of alpha activity 

for ZPPR plutonium.

a

Isotope Weight percent

Fraction of 

alpha activity

Pu-239 87.6 0.7

Pu-240 10.0 0.3

Pu-241 2.4 –

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

from an undocumented source.

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 byproduct (separated 241Am) from the purification of aged 

plutonium. Starting in 1969, parts per million of 241Am were measured for the plutonium mixture 

involved in significant possible inhalation incidents and were generally recorded on lung count reports 

for workers involved in those incidents. A nominal amount, 100 or 1,000 ppm by mass, of 241Am 

should be assumed if no other data are available. The value of 100 ppm 241Am should be considered 

for workers likely exposed to purified plutonium, such as workers involved in chemical, process, or 

metallurgical operations in Buildings 771, 776, 777, and 707. If the plutonium intake for WG

plutonium is assessed for 239,240Pu, the activity of 241Am in the intake mixture is calculated by 

multiplying the 239,240Pu activity by [48.2 × ppm 241Am ÷ (1 × 106

 - ppm 241Am)]. For ZPPR plutonium, 

the 239,240Pu activity is multiplied by [44.6 × ppm 241Am ÷ (1 × 106

 – ppm 241Am)] to obtain the activity of 

241Am in the intake mixture [3].

5.2.1.2 Plutonium Solubility and Particle Size

Most plutonium in metalworking operations and involved in fires is insoluble (i.e., type S). Exceptions 

such as plutonium metal associated with solvents such as carbon tetrachloride, can be assumed to be 

more soluble (type M) if this is what the data show or if it is more favorable to claimants to do so [4]. 
Document No. ORAUT-TKBS-0011-5 Revision No. 01 Effective Date: 02/01/2007 Page 12 of 123

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

which was strongly retained in the lungs of exposed workers with relatively low transfer to the urine, 

exhibited highly insoluble (super type S) characteristics [5]. 

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

mixture of solubilities. The dose reconstructor should select the material type that is most favorable to 

the claimant [6]. Lung count data in conjunction with urine data can help to determine absorption 

type. 

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

Therefore, dose reconstructions should use the default value of 5-μm activity median aerodynamic 

diameter (AMAD) (NIOSH 2002).

One exception is the plutonium fire on October 15, 1965, in Buildings 776 and 777 (Falk 2004), for 

which Mann and Kirchner (1967) measured a mass median diameter of 0.3 μm (1-μm AMAD) with a 

geometric deviation of 1.83. An approach that is favorable to claimants is to assume 1-μm AMAD for 

all plutonium fires unless the qualifying cancer involves the tissues of the extrathoracic regions [7]. 

5.2.2

5.2.2.1 Is otopic Compos ition

Americium

For the NIOSH Dose Reconstruction Project, the measured americium is 241Am [8]. 

5.2.2.2 Americium Solubility and Particle Size

Americium was present in two forms at RFP, as a purified byproduct of plutonium recovery and as 

atoms formed by the nuclear transformation of 241Pu and embedded in the matrix of the plutonium 

particle. As a purified byproduct, International Commission on Radiological Protection (ICRP) 

Publication 68 specifies americium inhalation absorption as type M (ICRP 1994a, p. 83). For 

embedded atoms in the matrix of an inhaled plutonium particle, the dose reconstructor should use the 

solubility classification described for the plutonium particle in Section 5.2.1.2 (ICRP 1994b, p. 79).

The dose reconstructor should use the default 5-μm AMAD particle size (NIOSH 2002) except for

situations involving fires, where a 1-μm AMAD should be assumed for consistency with Section 

5.2.1.2 above.

5.2.3

5.2.3.1 Is otopic Compos ition

Enriched Uranium

Production at RFP involved EU from 1952 to 1963. Table 5-3 lists the weight percent and fraction of 

alpha activity for each isotope.

Table 5-3. Weight percent and fraction of alpha activity 

for enriched uranium.

a

 

Isotope Weight percent

Fraction of 

alpha activity

U-234 1. 0.97

U-235 93 0.031

U-236 0.39 0.0039

U-238 5.4 0.00028
Document No. ORAUT-TKBS-0011-5 Revision No. 01 Effective Date: 02/01/2007 Page 13 of 123

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

5.2.3.2 Enriched Uranium Solubility and Particle Size

Operations for EU paralleled those for plutonium and included chemical processing and metalworking. 

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

classification is uncertain. The ICRP assigns UO2(NO3)2 (uranyl nitrate) to inhalation type F; 

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

inhalation type S (ICRP 1979, 1994b,c). All of these compounds were involved in the recovery and 

recycle processes for EU in Building 881 (RFETS 2000a).

In many cases, the compound of uranium involved in an intake is not identified. Dose reconstructors 

should use the solubility classification that is most favorable to claimants. 

If site-specific data for particle size of uranium are not available, dose reconstructors should use the 

default particle size value of 5-μm AMAD (NIOSH 2002). 

5.2.4

5.2.4.1 Is otopic Compos ition

Depleted Uranium

DU was present at RFP throughout its production history. Uranium-238 accounts for the majority of 

DU internal dose, but the total uranium alpha activity should be included in the dose reconstruction

(see Table 5-4). 

Table 5-4. Weight percent and fraction of alpha activity 

for depleted uranium.

a

Isotope Weight percent

Fraction of 

alpha activity

U-234 0.00058 0.097

U-235 0.23 0.013

U-238 99.77 0.89

a. These values are derived from data in DOE (1980, Volume 1, 

Table 2.7.2-4, p. 2-172).

5.2.4.2 Depleted Uranium Solubility and Particle Size

Operations with DU involved metalworking including casting, forming, and melting. Likely compounds 

are UO3 and U3O8 (RFETS 2000a). The solubility classification is ambiguous, falling somewhere 

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

reconstructors should use the solubility classification that is most favorable to claimants. 

If site-specific data for particle size of uranium are not available, dose reconstructors should use the 

default particle size value of 5-μm AMAD (NIOSH 2002). 

5.3 BIOASSAY DATA

The primary data for intake assessment at RFP 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 [9].
Document No. ORAUT-TKBS-0011-5 Revision No. 01 Effective Date: 02/01/2007 Page 14 of 123

5.3.1

Attachment A, Minimum Detectable Activity for Urinalysis Methods at RFP, discusses the history of 

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

were implemented at RFP. The following material summarizes, supplements, and expands the 

information in Attachment A.

Urinalysis Data

5.3.1.1 Plutonium Urinalys is

5.3.1.1.1 Methods , Units , Is otopes , and Interferences

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

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

volume or excretion period. Spot urine samples for plutonium were rarely requested and were usually 

associated with a significant incident, especially an incident with chelation (DTPA) treatment followup. 

Assume a 24-hr excretion period unless the record indicates that the actual excretion period was 

different [10]. 

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 favorable to claimants if the dose reconstructor 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 0.0235 [11]. 

Interferences were likely in the period from 1952 to 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 favorable to 

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

can be determined to be an outlier [12].

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 [13].

Another source of interference was contamination of the tracer (236Pu or 242Pu) by the analyte 

isotopes, 239Pu and 240Pu, which was an infrequent occurrence [14].

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

data from within 90 d 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 (HSDS), urine 

data affected by chelation were flagged with a code 1. Code 1 was also used to flag urine data that 
Document No. ORAUT-TKBS-0011-5 Revision No. 01 Effective Date: 02/01/2007 Page 15 of 123

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 [15].

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 RFP tolerance level). For 1962 to April 6, 1970, the minimum reporting level was 0.2 dpm/ 

24-hr sample. Results less than the reporting level were reported as 0.00 dpm/24-hr sample on 

computer-generated reports, such as the HSDS (see Attachment C, Figures C-3 and C-4) or 

background (or some abbreviation; e.g., BK) when manually recorded on the Urinalysis Record Card 

(see Figure C-3). For some workers, results initially reported as background on the Urinalysis Record 

Card were superseded by the report of the actual result in reports of the HSDS, 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 0.00 dpm/24-hr sample 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 disintegrations per minute per sample regardless of the sample 

volume [16].

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 [17]. 

Table 5-5 lists the MDA values for plutonium. The values for 1952 to 1977 are based on examination 

of urinalysis data logs for 1952 to 1971 (see Attachment A). The MDA value for 1971 was 

extrapolated through 1977. The MDA value for 1978 to 1989 is based on matrix blank data (RFETS 

1992) for the routine plutonium urinalysis program for August 1, 1990, to September 27, 1991, using 

blank values with a sample-specific recovery in the range of 0.1 to 1.1 dpm/24-hr sample. 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 1.1 dpm/ 

24-hr sample 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 0.02 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 

onsite laboratory [18]. Note that the value of the sample-specific MDA is included in the urinalysis 

data reports starting in 1990.

Some urine samples could have been processed by an offsite commercial laboratory before 1993. 

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

MDA listed in Table 5-5 should be used [19].

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 min. Because of the difficulty of determining which improvements 

apply to each sample, the MDAs in Table 5-5 do not account for the improvement until the transition 

was completed for all samples (i.e., the MDAs are favorable to claimants). 
Document No. ORAUT-TKBS-0011-5 Revision No. 01 Effective Date: 02/01/2007 Page 16 of 123

Table 5-5. Median MDA values for plutonium.

a,b

Period dpm/24-hr sample

1952–1953 0.57c

1954–1962 0.51c

1963 0.44

1964–1977 0.54

1978–1989 0.24

1990–1992 0.24

1993–present 0.02

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

dpm/sample.

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

starting in 1990, should be used instead of the generic 

MDA values in this table.

c. Note that these values of MDA are lower than the 

reporting level of 0.88 dpm/24-hr sample used at RFP 

through 1961. Many urine results in this period were rereported with the actual value if greater than zero. For 

those re-reported results, these MDA values apply 

instead of the original reporting level. 

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

The reported value was the 2-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 1-sigma standard 

error [20]. 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 A).

5.3.1.2 Americium Urinalys is

5.3.1.2.1 Methods , Units , Is otopes , and Interferences

Attachment A describes the methods through 1971. After 1971, the method for 241Am paralleled that 

for plutonium. 

The units of the results are disintegrations per minute per a 24-hr excretion period through 1989. 

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

volume or excretion period [21]. 

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

those of 241Am and has chemical properties similar to those of 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 with the PHA system, the 228Th 

could not be distinguished from the 241Am [22].

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 (corresponding to a parts-per-million 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) material from the ZPPR project, or (5) waste identified for those operations [23]. 
Document No. ORAUT-TKBS-0011-5 Revision No. 01 Effective Date: 02/01/2007 Page 17 of 123

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

intakes of WG plutonium [24]. 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 dpm/24 hr in 1963, ≥0.2 dpm/24 hr from 1964 to 1967, 

and ≥0.30 dpm/24 hr from 1968 to 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 from 1972 to

1976 has not been determined. Unless it is determined, dose reconstructors should assume that the 

reporting level for 1968 to 1971 was continued through 1976 [25]. 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 [21].

The MDAs for americium (Table 5-6) were determined as described for plutonium (see 

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

Table 5-6. 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

1978–1989 0.31

1990–1992 0.3

1993–present 0.02

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

dpm/sample.

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

record starting in 1990, should be used instead 

of the generic MDA values in this table. 

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

americium urinalysis.

5.3.1.3 Enriched Uranium Urinalys is

5.3.1.3.1 Methods , Units , Is otopes , and Interferences

The units of the results are disintegrations per minute per a 24-hr excretion period for the entire 

period.

Since urine samples analyzed for EU were counted with the air proportional detectors, all of the alphaemitting isotopes of uranium are included in the result.

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

methods for EU is not available. 

It is favorable to claimants to assume that the result is all EU [26].
Document No. ORAUT-TKBS-0011-5 Revision No. 01 Effective Date: 02/01/2007 Page 18 of 123

5.3.1.3.2 EU Reporting Levels , Minimum Detectable Activities , and Uncertainties

Table 5-7 lists the MDAs for EU. The reporting level for EU through 1963 was ≥8.8 dpm/ 

24-hr sample (10% of the RFP 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). It is undetermined when urinalysis for EU was stopped 

at RFP, although the stoppage likely occurred in the early 1970s [27].

Table 5-7. 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 EU were determined as described for plutonium (see Section 5.3.1.1.2 and 

Attachment A). 

Uncertainties for the EU 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 A).

5.3.1.4 Depleted Uranium Urinalys is

5.3.1.4.1 Methods , Units , Is otopes , and Interferences

Attachment A describes the uranium urinalysis methods through 1971. From 1972 to 1979, DU 

samples were chemically processed with the uranium-specific trioxyl phosphene oxide (TOPO) 

extraction procedure, and the extract (electrodeposited) was counted on the gas flow proportional 

counter. From 1980 to 1997, DU 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, 

DU samples have been processed at an offsite commercial laboratory according to provisions of the 

bioassay statement of work (RFETS 1998b).

The units for 1952 to April 1964 are micrograms of uranium per 24-hr excretion period. 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 [28]. 

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

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

measurements. 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 favorable 

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

intakes for the other isotopes [29]. In the 1990s, the urine data reports include the results separately 

for 234U, 235U, and 238U.
Document No. ORAUT-TKBS-0011-5 Revision No. 01 Effective Date: 02/01/2007 Page 19 of 123

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

concentrated pockets, in the terrain near RFP. No adjustments have been made to the reported DU 

urine results for this background, which was highly variable.

5.3.1.4.2 Depleted Uranium Reporting Levels , Minimum Detectable Activities , and 

Uncertainties

The minimum reporting level for DU through April 1964 was 5.8 μg/24-hr sample (10% of the 

tolerance level). From May 1964 to 1971, the minimum reporting level was the same as that for EU, 

20 to 28 dpm/24-hr sample, depending on the volume of the sample. The reporting level for 1972 to

1979 (TOPO procedure) has not been determined. An approach that is favorable to claimants is to 

use the reporting level for 1964 to 1971 [30]. 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 DU for fluorometric measurements were determined as described in Attachment A. 

Median MDAs for DU from 1952 to April 1964 are listed in Table 5-8. For alpha-counting methods, 

the MDAs in the period from April 1964 to 1971 are the same as those for EU in Table 5-6. The MDA 

value for 1972 to 1979 was extrapolated from the value determined for the previous period. The 

MDAs for 1980 to the present were derived in the same manner as that described for plutonium but 

are based on 238U. Table 5-9 lists median MDAs for DU from May 1964 to the present.

Table 5-8. 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 5-9. 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 0.4

1993–present 0.1

a. The MDA value unit starting in 1990 is 

dpm/sample.

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

starting in 1990, should be used instead of the 

generic MDA values in this table.

c. Actual practice is unknown; assume continuation 

of earlier practice.

The discussion of the uncertainty for plutonium in Section 5.3.1.1.2 applies to DU.
Document No. ORAUT-TKBS-0011-5 Revision No. 01 Effective Date: 02/01/2007 Page 20 of 123

5.3.1.5 Gros s Alpha Urinalys is

5.3.1.5.1 Methods , Units , Is otopes , and Interferences

Gross alpha measurement is a nonspecific 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, might also have been monitored 

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

discontinued in the early 1970s, likely in 1973 [31]. The results are reported as dpm/24-hr sample of 

either EU (the default analyte through 1963) or plutonium (the default analyte after 1963). 

Interferences are likely, because the methods were nonspecific. 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 RFP 

tolerance level for EU). 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 [32].) 

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 EU or to plutonium, or a portion to both. The 

default condition through 1963 was to credit the result to EU unless the PHA count indicated 

otherwise. After 1963 (when EU 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 nonspecificity of the analysis [33].

The MDAs for gross alpha listed in Table 5-10 were determined as described in Attachment A. 

Table 5-10. Median MDAs for gross alpha 

measurements. 

Period dpm/24-hr sample

1952 1

1953 0.88

1954–1959 0.79

1960–1962 0.55

1963 0.55

1964–1971 0.69

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 A). This 

uncertainty does not include the effect of interferences, which is a major issue for a nonspecific 

analysis like gross alpha measurement [34].

5.3.1.6 Tritium

Starting in 1973, workers were monitored for possible tritium exposures only for special projects or 

situations. The methods have not been reviewed but likely involved liquid scintillation measurements 

[35]. The urine results are reported as picocurie per 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 1 or 2 times the standard deviation. The sensitivity of the method was 2 to 

3 orders of magnitude better than the significant level of about 1 μCi/L. Although the actual MDA has 
Document No. ORAUT-TKBS-0011-5 Revision No. 01 Effective Date: 02/01/2007 Page 21 of 123

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

to several thousand picocuries per liter [36]. The current MDA for tritium is 600 pCi/L (RFETS 1998c, 

p. 7-3).

5.3.2

In vivo lung counts have been performed at RFP 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 WG plutonium at RFP. 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 (see Section B.11 in Attachment B). At RFP, 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 RFP (Falk et al. 1979). 

Lung Count Data

The RFP 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 DU. This measurement was made possible by the 

improved resolution of the germanium 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 (Berger 1988a). 

Attachment B, Minimum Detectable Activity for In Vivo Lung Counts at RFP, contains more detail. 

Section 5.4 and Attachment C discuss the data and report forms. 

5.3.2.1 Americium/Plutonium

5.3.2.1.1 Methods , Units , Is otopes , 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 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 WG plutonium) and americium (241Am) 

were recorded in nanocuries [37]. In addition, the activity of 241Am was stated as a fraction of the 

MPLB, using 14.7 nCi for the MPLB for 241

Am (Falk 1993). 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 DU operations being mistaken for 241Am in 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 DU operations [38]. The second interference 

is the contribution of count from 241Am not in the lungs (e.g., 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 (especially one that resulted in a 
Document No. ORAUT-TKBS-0011-5 Revision No. 01 Effective Date: 02/01/2007 Page 22 of 123

deposition to the alveolar-interstitial region of the lungs) of the plutonium/americium mixture, 

especially for a lung count in response to an incident [39].

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 [40]. 

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 1 standard 

deviation of 0.00 nCi plutonium [41].

The decision levels varied. From 1965 to 1968, the decision level was two times the uncertainty of 

the matched subject’s net count, although the application of this decision level was inconsistent in this 

period. Starting in 1969, for NaI and phoswich detector systems, the decision level was equal to 3 

times the standard deviation of the net count rate for a set of lung counts for unexposed known cold 

subjects based on the index method (see Attachment B). Results between 2 and 3 sigma were noted 

but not always investigated. For the germanium 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 [42]. 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% (RFETS 2000a). The decision level was used as a reporting level 

from 1995 to early 1997. 

Table 5-11 lists the MDAs for 241Am, which were calculated for the evolution of lung counting systems 

used at RFP as described in Attachment B. 

These values of MDAs are for three indices that represent the median and the approximate 5th- and 

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

reconstructor can calculate the value using the information in Attachment B or interpolate (or 

extrapolate) from the values in Table 5-11 [43]. The worker-specific index is generally stated on lung 

count report forms from 1969 to 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 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 (or index) data for the worker are not 

available [44].

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

241Am by the MDA conversion factor (Equation B-17 in Attachment B), 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 
Document No. ORAUT-TKBS-0011-5 Revision No. 01 Effective Date: 02/01/2007 Page 23 of 123

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

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

Table 5-11. 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) 4 × 4 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) 4 × 4 0.90 – – 0.8 0.76

1.35 – – 1.3 1.3

1.80 – – 2.2 2

1973 → Phoswich 0.90 – – 1.2 1.2

1.35 – – 2.0 2.

1.80 – – 3.3 3.2

1976 – 1978 Ortec Arrays

(High-purity Ge)

0.90 0.26 0.18 0.2 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 0.2 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 0.4 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 0.4 0.28 0.31 0.21

1.80 0.74 0.5 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 0.1

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

available records, an approach that is favorable to claimants is to assume the initial ppm 241Am to be 

100 [45]. 

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 [46]. 

The dose reconstructor must choose the value of the initial mass fraction of 241Pu. At the RFP 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 to June 1976. The fraction 0.0036, based on the isotopic 

composition for RFP stream plutonium in the mid-1970s, should be used for intakes that occurred 

from July 1976 to 1989. For intakes incurred after 1989, the initial fraction of 241Pu should be reduced 

to account for the aging (radioactive decay) of the 241Pu [47].
Document No. ORAUT-TKBS-0011-5 Revision No. 01 Effective Date: 02/01/2007 Page 24 of 123

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

germanium detector systems in 1976. The uncertainty was reported at 1 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 [48]. With the advent of the ABACOS-Plus© software in 1995, the percent error at 1 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 2000a, 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 241Am, 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 [49].

5.3.2.2 Thorium/Depleted Uranium

5.3.2.2.1 Methods , Units , Is otopes , and Interferences

The method to detect DU 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 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 B) and 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); the ratio (59.5-keV gamma/63-keV doublet gamma) is 9.4]. That is, nanocuries of 238U equals 

[(234Th 63-keV net cpm) divided by (241Am calibration factor)] multiplied by 9.4. To calculate the 

activity for DU, 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 reconstructor should use the assumption 

(favorable to claimants) that the 238U activity is all from occupational exposure to DU [50].
Document No. ORAUT-TKBS-0011-5 Revision No. 01 Effective Date: 02/01/2007 Page 25 of 123

5.3.2.2.2 Reporting Levels , Minimum Detectable Activities , and Uncertainties 

Reporting levels were not generally used for DU until 1995 with the implementation of the ABACOSPlus© 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 DU [51]. (As noted in Section 5.3.2.1.2 for americium and plutonium, 

MDA values are reported on report forms for 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 

DU 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 DU metal with a deficiency of 234Th for several 

weeks plus scrap DU with an excess of 234Th (super-equilibrium). The assumption of equilibrium 

when super-equilibrium existed is favorable to claimants [52]. The effect of a deficiency of 234Th (not 

favorable to claimants) is mitigated by the rapid ingrowth of the 234Th into the DU. Fifty-percent

equilibrium occurs after 24 d following a thorium strike, and 90% occurs after 80 d. 

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 DU activity calculated from 

the net count rate, the dose reconstructor can divide the worker-specific MDA by 3.3 [53].

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 RFP 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 (Berger 1988b). The sequence of additional decontamination was washing with soap 

and water, washing with commercial bleach©, scrubbing with commercial bleach©, and excision.

Wound count information is largely irrelevant to dose reconstruction [54]. 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 
Document No. ORAUT-TKBS-0011-5 Revision No. 01 Effective Date: 02/01/2007 Page 26 of 123

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

The profile of the urine data after 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 occurred rarely, if at all. It is reasonable to 

assess any initially positive uranium wound as an acute injection.

5.3.3.2 Nas al Smears and Fecal Samples

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

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). 

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

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

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

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

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

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

The reporting times are the times for the laboratory to analyze the sample and report the results. The 

shorter reporting times indicate an expedited analysis, with the trade-off of a less sensitive analysis (a 

higher MDA).

These MDA values apply to samples starting approximately in 1993 and are specifications for the 

laboratory. (Note: The laboratory MDA does not depend on the time after intake that the sample was 

excreted.) Most reports of fecal sample results do not give the sample-specific MDA but might give 

the decision level (Lc), which is approximately one-half of the sample-specific MDA. 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 [55]. Attachment C, Examples of Records and Reports Used at RFP, contains the 

figures described below.
Document No. ORAUT-TKBS-0011-5 Revision No. 01 Effective Date: 02/01/2007 Page 27 of 123

5.4.1

Figures C-1 to C-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 to 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 HSDS was implemented to record, process, and report urinalysis data 

and the derived fraction of the maximum permissible systemic burden.

Urinalysis Records and Reports

5.4.1.1 Interpretation of the Urinalys is 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 A). 

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

Figure C-1, analysis Code A, 1955). 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) [56]. On some cards, the dose 

reconstructor might 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 HSDS. It is reasonable and favorable to claimants to disregard the 

modified result [57]. 

The analyte code for DU was sometimes transcribed incorrectly from the card to the urinalysis detail 

report as U (see Figure C-1) rather than D (see Figures C-2 and C-3) with the unit of 

dpm/24-hr sample rather than μg/24-hr sample. 

Figures C-4 and C-5 are two versions of urinalysis reports from the HSDS. Both versions report the 

data in the same way but with differences in the headers. Figure C-5 (the newer version) adds a 

column (the uncertainty of the result).

5.4.1.2 Interpretation of the Health Sciences Data Sys tem – Urinalys is 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
Document No. ORAUT-TKBS-0011-5 Revision No. 01 Effective Date: 02/01/2007 Page 28 of 123

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 d of 

a chelation treatment were generally (or should have been) coded as 1 [58]. The use of Code 2 to 

flag the date of a new significant intake occurred 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 DU, from 1952 to 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 [59].

• The BODY BURDEN % or SYSTEM BURDEN is the 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.

5.4.1.3 Interpretation of Other Urinalys is Reports

Figures C-6 and C-7 are examples of urinalysis reports from the onsite bioassay laboratory from 1990 

to the mid-1990s. Figure C-6 is for a special urine sample analyzed for plutonium, and Figure C-7 is 

for a routine urine sample analyzed for plutonium. Both forms have the same format. The first three 

columns are self-explanatory; the remaining columns are:

• Dec Level is the decision level (Lc) in units of dpm/sample.

• 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 HSDS. Codes 1, 3, and 4 indicate a failed analysis and disqualify the result [60].

• 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 DQO, and, if the batch failed, every sample in the batch was 

conditionally failed pending further evaluation [61].

• 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 [62].
Document No. ORAUT-TKBS-0011-5 Revision No. 01 Effective Date: 02/01/2007 Page 29 of 123

• Analyte is self-explanatory.

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

• 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 [63].

• Error is the uncertainty at 1 standard deviation.

Figure C-8 is an example of the urinalysis data report by Quanterra, a commercial offsite 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 [64]. The result header is largely self-explanatory. 

• The primary information is the RESULT and its TOTAL ERROR (at 1 standard deviation) in 

units of dpm/sample (REPORT UNIT). 

• 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.

Figures C-9 and C-10 are examples of the analytical report of the onsite bioassay in the mid-1990s. 

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 C-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, not an occupational intake of DU.

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

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

Figure C-12 is the urinalysis data report for General Engineering Laboratories, Inc. The header 

information is largely self-explanatory. 
Document No. ORAUT-TKBS-0011-5 Revision No. 01 Effective Date: 02/01/2007 Page 30 of 123

• The Date Collected is the sample excretion date. The 24-hr clock time (0600) is also noted; 

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

documented [65].

• 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 have a Data Validation Code = V.

Figure C-13 is an example of the 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 nasal smears. The unit of the 

tritium results is pCi/L. The unit of the fecal sample and nasal smear results is dpm/sample [66].

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

plutonium fire in Buildings 776 and 777 was the first extensive use of the lung counter to detect 

americium and plutonium depositions for RFP workers. 

Lung Count Records and Reports

• The in vivo lung counting system was called the Body Counter. In RFP 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 RFP 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) [67].

• 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) [68].

• 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 the liver or gut or below the sternum rather than over one side of the chest. 

Those data have little dosimetric use [69].

Figure C-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) c/m, and after 

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

in terms of micrograms of plutonium.

Figure C-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. 
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• 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 C-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 B to 

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

any assumed or actual intake date.

Figure C-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 [70]. 

• 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 of 241Am) for a previous 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 C-17).

Figure C-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 c/m [71].

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

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

detectors of the germanium 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 C-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. 

Rather, it reports the cutoff, which is the decision level, and Normal if the DIFFERENCE is less than 

the cutoff [72].

Figure C-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 germanium detector 
Document No. ORAUT-TKBS-0011-5 Revision No. 01 Effective Date: 02/01/2007 Page 32 of 123

system [73]. 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 to 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 C-23 and C-24 are examples of the next generation of reports for the germanium detector 

systems. The innovation is the data capture in 10 ROIs. In Figure C-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 [74]. ROI 5, labeled BKG in Figure C-24, is the common background (divided by a factor) for 

both the 60- and 63-keV photopeaks. 

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

fields.

Figure C-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) [75].

Figure C-27 is an example of a report on which data for the 93-keV photopeak are analyzed and 

presented.

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

detector are tabulated separately, as is 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 

does not report the results in terms of the fraction of the MPLB, an obsolete concept since 1989 [76].

Figure C-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 [77]. 

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

mid-February 1997, when the reporting protocol was changed. The primary change was that 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 C-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 ingrowth), which was

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 [78]. 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 B-18 in Attachment B and the value of the 

initial ppm 241Am tabulated generally on one of the early lung count reports [79].
Document No. ORAUT-TKBS-0011-5 Revision No. 01 Effective Date: 02/01/2007 Page 33 of 123

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 [80].

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

counts (Berger 1988b; RFETS 2000a). 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 of 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 B-18 in Attachment B. The dose reconstructor 

should adjust the activities for the discontinuity factors presented in Attachment B. In general, use of 

the discontinuity factors is favorable to claimants [81]. 

5.5 ATTRIBUTIONS AND ANNOTATIONS

Where appropriate in this document, bracketed callouts have been inserted to indicate information, 

conclusions, and recommendations provided to assist in the process of worker dose reconstruction. 

These callouts are listed here in the Attributions and Annotations section, with information to identify 

the source and justification for each associated item. Conventional References, which are provided in 

the next section of this document, link data, quotations, and other information to documents available 

for review on the Project’s Site Research Database. 

Much of the information in this TBD, including the Attachments, was written by Roger B. Falk and is 

based on his insights, recollections, research and development activities, and administration in the 

radiation dosimetry and health effects programs at the Rocky Flats Plant.

[1] Falk, Roger B. ORAU. Senior Life Scientist. June 2006.

The statements of the primary types of intakes and bioassay data are based on the 

observations by the author during his work at RFP in the internal dosimetry and heath effects 

programs.

[2] Falk, Roger B. ORAU. Senior Life Scientist. June 2006.

This statement is based on the observations of such notations on incident and lung count 

reports related to ZPPR materials.

[3] Falk, Roger B. ORAU. Senior Life Scientist. June 2006.

The multiplier for WG plutonium is the inverse of Equation B-17 in Attachment B. This 

multiplier is modified to apply to ZPPR plutonium based on the ratio of the weighted specific 

activities of the 239Pu and 240Pu for WG and ZPPR plutonium, 0.071 and 0.0767, respectively. 

The ratio of 0.926 times 48.2 results in the value of 44.6 found in the ZPPR multiplying factor. 
Document No. ORAUT-TKBS-0011-5 Revision No. 01 Effective Date: 02/01/2007 Page 34 of 123

[4] Falk, Roger B. ORAU. Senior Life Scientist. June 2006.

Oxides of plutonium metal (air-oxidized and fire-oxidized) are classified as type S and most 

other plutonium compounds as type M by the ICRP (ICRP 1994a). In any case, the dose

reconstructor should use the solubility class that is most favorable to claimants. 

[5] Falk, Roger B. ORAU. Senior Life Scientist. June 2006.

Retention of plutonium in the lungs of workers exposed in the 1965 plutonium fire was 

observed to be more avid than predicted by the default ICRP type S model (ICRP 1994b), 

based on lung counts performed as part of the Former Radiation Worker Medical Surveillance 

Program at RFP, 28 to 38 yr after intake.

[6] Falk, Roger B. ORAU. Senior Life Scientist. June 2006.

Oxides (air-oxidized and fire-oxidized) of plutonium metal are classified as type S and most 

other plutonium compounds as type M by the ICRP (ICRP 1994a). In any case, the dose

reconstructor should use the absorption type that is most favorable to claimants. 

[7] Falk, Roger B. ORAU. Senior Life Scientist. June 2006.

The recommendation is favorable to claimants when intakes are assessed from airborne 

plutonium data and is essentially neutral when intakes are assessed from urine or lung count 

data.

[8] Falk, Roger B. ORAU. Senior Life Scientist. June 2006.

This is a self-evident statement because the source of the americium is only from the decay of 

241Pu. No other americium isotopes are involved.

[9] Falk, Roger B. ORAU. Senior Life Scientist. June 2006.

This statement of how bioassay data were used during operations at RFP does not preclude 

the use of these data for dose reconstructions when and if appropriate.

[10] Falk, Roger B. ORAU. Senior Life Scientist. June 2006.

Spot urine samples for plutonium were rarely requested and then usually associated with a 

significant incident, especially an incident with chelation (DTPA) treatment follow-up. Such 

exceptions should be easily discernable in the documentation of the incident in the worker’s 

health physics file, especially starting in 1990, the period of interest for this recommendation. 

In addition, many of the urine sample result reports have the volume of the sample recorded 

[see Figures C-8 to C-12. (Although some of these examples are for uranium analytes, the 

format of the report is the same for plutonium analytes.)]. These reported volumes can be 

used to normalize the result to a 24-hr sample when appropriate.

[11] Falk, Roger B. ORAU. Senior Life Scientist. June 2006. 

The original 2003 version of this sentence was stated incorrectly. The factor of 1.0264, when 

multiplied by the intake assessed from 239,240Pu urine data, would yield the intake for 

238,239,240Pu, not the 238Pu component of the intake. In addition, the factor was based on a 

slightly different isotopic composition from that stated in Table 5-1. The 238Pu component of 

the intake is obtained correctly by multiplying the intake assessed from 239,240Pu urine data by 

0.0235, a value obtained by dividing the 238Pu fraction of alpha activity stated in Table 5-1 by 

0.98, the sum of the fractions of alpha activity for the 239Pu and 240Pu isotopes.
Document No. ORAUT-TKBS-0011-5 Revision No. 01 Effective Date: 02/01/2007 Page 35 of 123

[12] Falk, Roger B. ORAU. Senior Life Scientist. June 2006.

Interferences that add to the value of the analyte are always favorable to claimants. 

Therefore, the recommendation was made to use the results as found in the record unless the 

dose reconstructor had generic instructions, outside the purview of this TBD, to do otherwise.

[13] Falk, Roger B. ORAU. Senior Life Scientist. June 2006.

This statement is based on the property of the PHA system to separate and count the alphas 

by their energies. The alpha energies of the 239Pu and 240Pu isotopes were sufficiently 

different from the alpha energies of americium and thorium to allow plutonium analyses to be 

unaffected by the presence of americium or thorium, if any.

[14] Falk, Roger B. ORAU. Senior Life Scientist. June 2006.

This statement, based on informal discussions in the 1980s and early 1990s with Dale L. 

Bokowski, the lead bioassay chemist at Rocky Flats starting in 1961, was included here for 

completeness.

[15] Falk, Roger B. ORAU. Senior Life Scientist. June 2006.

The recommendation for the dose reconstructor to be wary of, and generally not use, urine 

data flagged with Code 1 is based on good science and common sense. Although chelationenhanced urine data might be favorable to claimants, use of such data without modification in 

standard models that are based on unenhanced data is not scientifically sound. It is also not 

sound to use data that did not pass quality standards in real time.

[16] Falk, Roger B. ORAU. Senior Life Scientist. June 2006.

The information in this paragraph is based on observations and deductions of the author from 

review of original urine data logs and individual urine data reports in preparation of this TBD 

section and Attachment A, as well as from personal involvement in the development of 

improved urinalysis reporting protocols in real time at RFP in the 1980s and early 1990s.

[17] Falk, Roger B. ORAU. Senior Life Scientist. June 2006.

The recommendation to use the median value of the MDA if a sample-specific MDA could not 

be determined was the outcome of the internal review of a draft of this document, to be 

consistent with the generic approach adopted by the program. This recommendation no 

longer needs to be made.

[18] Falk, Roger B. ORAU. Senior Life Scientist. June 2006.

This statement is based on an undocumented conversation with Steven C. Baker, Manager of 

Radiological Health at RFETS, in the summer of 2003.

[19] Falk, Roger B. ORAU. Senior Life Scientist. June 2006.

The recommendation to use the values in Table 5-19 seemed to be the only viable option.

[20] Falk, Roger B. ORAU. Senior Life Scientist. June 2006.

This information is based on the author’s recollections of the implementation of these 

upgrades in real time at RFP. 

[21] Falk, Roger B. ORAU. Senior Life Scientist. June 2006.

This statement is based on the observation that the same reporting format used for plutonium 

results was used for americium results.
Document No. ORAUT-TKBS-0011-5 Revision No. 01 Effective Date: 02/01/2007 Page 36 of 123

[22] Falk, Roger B. ORAU. Senior Life Scientist. June 2006.

This information is based on informal conversations with in the early 1990s Dale L. Bokowski,

the lead bioassay chemist at Rocky Flats starting in 1961, prior to his retirement from RFP in 

1992.

[23] Falk, Roger B. ORAU. Senior Life Scientist. June 2006.

This credibility check is presented to the dose reconstructor as optional guidance and does not 

preclude the use of americium urine data for dose reconstructions if deemed appropriate by 

the dose reconstructor, even if the data do not pass this credibility test. Note also that the 

maximum ingrowth of americium in virgin WG plutonium (with 0.5% by weight 241Pu) is less 

than 5,000 ppm. Plutonium with 10,000 ppm 241Am or greater would be credible only for a 

process that enhanced the americium concentrations, such as those processes listed.

[24] Falk, Roger B. ORAU. Senior Life Scientist. June 2006. 

This recommendation was based on the plutonium/americium activity ratio, which is 

considerably greater than 2 for WG plutonium, and the problem of the thorium interferences in 

americium urinalyses. These two factors make the plutonium urine data set the better choice 

to determine plutonium intakes.

[25] Falk, Roger B. ORAU. Senior Life Scientist. June 2006. 

This recommendation extrapolates a value favorable to claimants forward to the next point in 

time for which the reporting level was determined.

[26] Falk, Roger B. ORAU. Senior Life Scientist. June 2006.

This assumption states the obvious; that is, if there are interferences that contribute to the 

magnitude of the results, considering those interferences as EU results in a higher than actual 

outcome and, hence, is favorable to claimants. 

[27] Falk, Roger B. ORAU. Senior Life Scientist. June 2006. 

It is not clear in the urine data logs for 1964 to 1971 which electroplated uranium samples 

were for EU and which were for DU. EU operations were discontinued in this period although 

some urine sampling for EU could have occurred for workers involved in decontamination 

activities.

[28] Falk, Roger B. ORAU. Senior Life Scientist. June 2006.

The statements about the units of the reported urine data are based on observations of 

numerous urine data reports.

[29] Falk, Roger B. ORAU. Senior Life Scientist. June 2006.

This recommendation implies dividing the intake assessed from the assumed 238U urine data 

by 0.89 (the 238U fraction of total DU activity) to calculate the total DU intake. This approach is 

favorable to claimants by about 12% if the activities of the 234U and 235U were actually included 

in the reported uranium urine results in the 1980s.

[30] Falk, Roger B. ORAU. Senior Life Scientist. June 2006.

This recommendation extrapolates the earlier practice for a period when the actual practice is 

not known. This approach is favorable to claimants if the earlier values would result in a 

higher intake assessment, as in this case.
Document No. ORAUT-TKBS-0011-5 Revision No. 01 Effective Date: 02/01/2007 Page 37 of 123

[31] Falk, Roger B. ORAU. Senior Life Scientist. June 2006.

No documented date was found about when gross alpha analyses were discontinued. The 

year 1973 was estimated following a review of HSDS urinalysis reports for a sampling of 

workers previously sampled for gross alpha, with a finding of no analysis code G after 1972.

[32] Falk, Roger B. ORAU. Senior Life Scientist. June 2006.

Urinalysis code G was observed in HSDS urinalysis reports to 1972. Code G correlates with 

the gross alpha B2 analysis code on the Urinalysis Record Card (see, for example, Figure C-2).

[33] Falk, Roger B. ORAU. Senior Life Scientist. June 2006.

A better way to indicate the generosity of the nonspecificity of the gross alpha result if applied 

to a specific radionuclide is to use the term “favorable to claimants” rather than “upper 

bounds.”

[34] Falk, Roger B. ORAU. Senior Life Scientist. June 2006.

The effect of interferences is not included in the estimate of the standard deviation because it 

is not really a random variable but rather an intermittent bias of unknown magnitude.

[35] Falk, Roger B. ORAU. Senior Life Scientist. June 2006.

Liquid scintillation was used as the counting method for tritium in the 1980s, based on the 

author’s personal observations. No documentation has been noted about the earlier systems 

at RFP, but it seems reasonable to consider that the same method was used in the 1970s.

[36] Falk, Roger B. ORAU. Senior Life Scientist. June 2006.

Numerous tritium results at these levels have been observed by the author in worker’s health 

physics files at the cited levels.

[37] Falk, Roger B. ORAU. Senior Life Scientist. June 2006.

These modifications were implemented by the author in 1973 at RFP. Figure C-18 is an 

example of the implementation of these modifications.

[38] Falk, Roger B. ORAU. Senior Life Scientist. June 2006.

This interference, though troublesome in real time, is favorable to claimants whose americium 

lung count results were enhanced by count from 234Th.

[39] Falk, Roger B. ORAU. Senior Life Scientist. June 2006.

Interferences, especially contamination on the worker’s chest, occasionally caused a false

positive lung count. This statement was intended to alert the dose reconstructor to this 

possibility. The sentence is modified in recognition of the fact that an intake could have 

occurred without resulting in a deposition in the alveolar-interstitial region of the lung.

[40] Falk, Roger B. ORAU. Senior Life Scientist. June 2006.

These modifications were implemented by the author in 1974 at RFP. 

[41] Falk, Roger B. ORAU. Senior Life Scientist. June 2006.

This modification was implemented by the author in the early 1980s at RFP.

[42] Falk, Roger B. ORAU. Senior Life Scientist. June 2006.

The cutoff, as defined, is based on limiting the probability of a Type I error (false positive) in 

the signal domain to 5%. Figure C-22 is an example of the implementation of this decision 

level.
Document No. ORAUT-TKBS-0011-5 Revision No. 01 Effective Date: 02/01/2007 Page 38 of 123

[43] Falk, Roger B. ORAU. Senior Life Scientist. June 2006.

The worker-specific MDA for the americium in vivo measurement depends on the worker’s 

index and the calibration factor K for that index for the detector system used for the worker’s 

lung count. A relatively easy method is to normalize the MDA value for the 1.35 index by the 

ratios of the calibration factors (given in Attachment B for each detector system) for index 1.35, 

and the worker-specific index is:

MDAworker = MDA1.35 × K1.35/Kworker

An easy method to derive the worker-specific MDA from values listed in Table 5-11 is to use a 

spreadsheet to plot the MDA values for the three indices, for the detector system of interest, 

and to determine the equation for an exponential trend line. This equation, in the form y = 

#.#### e#.#### x, where x is the worker’s index and y is the worker-specific MDA, can then be 

used to calculate the worker-specific MDA.

[44] Falk, Roger B. ORAU. Senior Life Scientist. June 2006.

The recommendation to use the median value of the MDA is consistent with the generic 

approach adopted by the program. Except for some workers with lung counts only in the 

1960s, this situation is expected to be rare.

[45] Falk, Roger B. ORAU. Senior Life Scientist. June 2006.

The recommendation to use an initial value of 100 ppm Am, if the actual value for an intake is 

not documented, is based on freshly purified plutonium (within 0 to 5 mo, depending on the 

efficiency of the purification process).

[46] Falk, Roger B. ORAU. Senior Life Scientist. June 2006.

This guidance is based on the consideration that significant intakes of plutonium at RFP were 

acute intakes, albeit sometimes a set of intermittent acute intakes. After implementation of the 

body counter in 1965 and as the sensitivity of the system improved, the assignment of the 

intake date to newly detected depositions, but not from a new intake, was problematic. The 

dose reconstructor may have generic instructions, especially for efficiency methods, for 

assigning the intake scenario.

[47] Falk, Roger B. ORAU. Senior Life Scientist. June 2006.

This guidance is presented to support the calculation of the ingrowth of 241Am after the date of 

an acute intake or after the start of a chronic intake. After the end of plutonium production 

activities in 1989, the initial fraction of 241Pu in RFP plutonium was a decreasing variable 

based on the age of the plutonium since blending.

[48] Falk, Roger B. ORAU. Senior Life Scientist. June 2006.

These modifications were implemented by the author in 1976 and 1981 at RFP. Figures C-20 

and C-21 are examples of the implementation of these modifications.

[49] Falk, Roger B. ORAU. Senior Life Scientist. June 2006.

The validity of this assumption, listed as an uncertainty, seems to be supported by 

observations that americium lung count measurements for many RFP workers with confirmed 

lung depositions of plutonium-americium mixtures have remained relatively constant or have 

slightly increased at decades after the initial short-term clearance period of several years. 

Recent US Transuranium and Uranium Registries autopsy data for RFP cases also indicate 

the retention of americium in the lungs that is consistent with this assumption.
Document No. ORAUT-TKBS-0011-5 Revision No. 01 Effective Date: 02/01/2007 Page 39 of 123

[50] Falk, Roger B. ORAU. Senior Life Scientist. June 2006.

The basis for this recommendation is the same as that stated in [31].

[51] Falk, Roger B. ORAU. Senior Life Scientist. June 2006.

Dividing the 238U MDA by 0.89 accounts for the contribution to the DU MDA from activities of 

the other uranium isotopes.

[52] Falk, Roger B. ORAU. Senior Life Scientist. June 2006.

If a super-equilibrium situation was operative and the 234Th lung count result was used to 

calculate the DU assuming equilibrium, the calculated DU would be higher than the actual 

activity. Therefore, the situation is favorable to claimants. 

[53] Falk, Roger B. ORAU. Senior Life Scientist. June 2006. 

This method suggested to estimate the uncertainty of the activity from its MDA is the same as 

the method suggested in Sections 5.3.1.1.2, 5.3.1.3.2, and 5.3.1.5.2 and is equally applicable.

[54] Falk, Roger B. ORAU. Senior Life Scientist. June 2006.

Because the dose reconstructor is likely to find numerous wound count reports in files of 

workers assigned to plutonium areas, this statement helps to focus the attention of the dose 

reconstructor on the most relevant data for quantifying internal doses to organs. The relevant 

data are cited in the next sentence. The actual wound count and contamination data might be 

relevant if the cancer site coincided with the site of the wound, an occurrence not yet noted by 

the author.

[55] Falk, Roger B. ORAU. Senior Life Scientist. June 2006.

The interpretations are those of the author, either gleaned from using the data as an internal 

dosimetrist at RFP or as the designer and implementer of the reports as part of the technical 

staff supporting RFP internal dosimetry programs.

[56] Falk, Roger B. ORAU. Senior Life Scientist. June 2006.

The urine data record written on the Urinalysis Record Card preceded the HSDS and was the 

likely source of the urine data loaded into the HSDS and its mainframe database 

predecessors. Because there could have been transcription errors during the preparation of 

the data (punched cards in the 1960s) for loading into the mainframe, the data of the 

Urinalysis Record Cards (the source data) should considered the correct data, as 

recommended.

[57] Falk, Roger B. ORAU. Senior Life Scientist. June 2006.

This recommendation is reasonable because the basis for the change is not known and the 

change was not made in the HSDS. It is favorable to claimants because the original record is 

the higher value.

[58] Falk, Roger B. ORAU. Senior Life Scientist. June 2006.

Exceptions to this practice have been observed in the HSDS urinalysis reports for some 

workers with documented chelation therapy, especially for americium results from analysis of a 

urine sample also analyzed for plutonium (and the plutonium result was correctly coded with 

Code 1).

[59] Falk, Roger B. ORAU. Senior Life Scientist. June 2006.

This circumstance is evident in the example reports in Figures C-2, C-3, and especially C-5.
Document No. ORAUT-TKBS-0011-5 Revision No. 01 Effective Date: 02/01/2007 Page 40 of 123

[60] Falk, Roger B. ORAU. Senior Life Scientist. June 2006.

This is presented for information only. The decision of whether to disqualify the result was the 

call of the laboratory Quality Assurance (QA) officer who reviewed the data and signed the 

report. The Batch Val code V is the primary indicator of a valid result.

[61] Falk, Roger B. ORAU. Senior Life Scientist. June 2006.

This is presented for information only. In practice, the batch evaluation review would have 

occurred before the release of the Analytical Report. As noted in [60], the Batch Val code V is 

the primary indicator of a valid result. 

[62] Falk, Roger B. ORAU. Senior Life Scientist. June 2006.

It seems only prudent not to use a result that was invalidated based on failure to meet quality 

standards.

[63] Falk, Roger B. ORAU. Senior Life Scientist. June 2006.

See narrative for [10]. 

[64] Falk, Roger B. ORAU. Senior Life Scientist. June 2006.

The offsite laboratory sometimes was not provided the date on which the worker excreted the 

urine sample. In such cases, the excretion date was written on the report, as was the case for 

the report in Figure C-8. The guidance for the dose reconstructor is to use the date written on 

the report if such a situation occurs.

[65] Falk, Roger B. ORAU. Senior Life Scientist. June 2006.

This is presented for information only. The time of the end of the excretion period is not critical 

for retrospective dose reconstructions. The Integrated Modules for Bioassay Analysis (IMBA) 

code used by Project dose reconstructors has a default sample time of 12:00 a.m. 

[66] Falk, Roger B. ORAU. Senior Life Scientist. June 2006.

The units for fecal and nasal smear sample results are well known to the author from his 

experience as internal dosimetrist at RFP. In addition, the units are likely to be stated explicitly 

on other reports in the health physics files of affected workers.

[67] Falk, Roger B. ORAU. Senior Life Scientist. June 2006.

The time of day of the count is not critical for retrospective dose reconstruction. Later lung 

count reports usually recorded the time of day in this field and noted the count time only if it 

was different from the standard count for the era, through the 1970s (see Attachment B for the 

standard count times). Electronically generated lung count reports, starting circa 1981, record 

the count time used for that count (see Figures C-21 to C-31). The count time would be useful 

to the dose reconstructor mainly to calculate an MDA for a given lung count, if needed.

[68] Falk, Roger B. ORAU. Senior Life Scientist. June 2006.

The value of the MPLB for plutonium alpha emitters (239Pu and 240Pu) was calculated using 

equation (4) in ICRP Publication 2 (ICRP 1959) for an annual dose of 15 rem (0.3 rem/week), 

organ mass (m) = 1,000 g, f2 = 1, and ε = 53 (from ICRP Publication 2, Table 5, and based on 

a relative biological effectiveness = 10).

[69] Falk, Roger B. ORAU. Senior Life Scientist. June 2006.

The main reason the measurements obtained by the detector over gut/liver/below-sternum 

area are not dosimetrically useful is that a calibration factor was not developed in real time to 

convert the signal to activity. It was a subjective measurement; i.e., was it normal or high?
Document No. ORAUT-TKBS-0011-5 Revision No. 01 Effective Date: 02/01/2007 Page 41 of 123

[70] Falk, Roger B. ORAU. Senior Life Scientist. June 2006.

This ratio was the subjective rule-of-thumb used by the author in real time at RFP as a 

supplemental method to discern possible low-level depositions of the plutonium-americium 

mixtures for lung counts performed with the NaI detector system.

[71] Falk, Roger B. ORAU. Senior Life Scientist. June 2006.

The variability in the background in the L X-ray region of the spectrum prevented the 

establishment of a stable calibration factor for the direct measurement of plutonium via 

L X-rays. The author was directly involved in this effort.

[72] Falk, Roger B. ORAU. Senior Life Scientist. June 2006.

See, for example, Figure C-22.

[73] Falk, Roger B. ORAU. Senior Life Scientist. June 2006.

This statement reflects the practice to perform a follow-up lung count with a better resolution 

detector system (i.e., a germanium detector system) when action levels for a count with a 

poorer resolution, NaI scintillation detector system, such as the phoswich detectors, were 

exceeded. 

[74] Falk, Roger B. ORAU. Senior Life Scientist. June 2006.

See, for example, Figure C-26.

[75] Falk, Roger B. ORAU. Senior Life Scientist. June 2006.

The dose reconstructors are advised to disregard these L X-ray data because the counts were 

unreliable because of low-end electronic noise. Because of this unsolved problem in real time, 

a calibration factor to convert from count of plutonium activity was not established.

[76] Falk, Roger B. ORAU. Senior Life Scientist. June 2006.

DOE Order 5480.11, implemented in 1989, shifted the basis from the ICRP Publication 2 

approach to control the dose to a critical organ to the ICRP Publication 26 and 30 approach of 

assessing the committed dose equivalent to organs from intakes (ICRP 1959, 1977, 1979; 

DOE 1989). The quality factor for alpha radiation was increased from 10 to 20.

[77] Falk, Roger B. ORAU. Senior Life Scientist. June 2006.

This statement is the result of direct observation of information in Figure C-29.

[78] Falk, Roger B. ORAU. Senior Life Scientist. June 2006.

Indeed, not only is the plutonium decision level inoperative because the decision is based on 

the detection of americium, but also the decision level is never operative for follow-up 

measurements of a confirmed deposition – there is no decision to be made. A decision level is 

operative only if the null hypothesis is operative. The null hypothesis is not operative in this 

example. 

[79] Falk, Roger B. ORAU. Senior Life Scientist. June 2006.

This information was provided just in case the dose reconstructor would want to determine the 

value of the parts per million of the 241Am used in the calculation. 

[80] Falk, Roger B. ORAU. Senior Life Scientist. June 2006.

The ABACOS-Plus© software used the date of the count as the default intake date unless an 

intake date was specifically input for the count. This statement should not be interpreted as 

guidance to the dose reconstructor to use that default date in dose reconstructions.
Document No. ORAUT-TKBS-0011-5 Revision No. 01 Effective Date: 02/01/2007 Page 42 of 123

[81] Falk, Roger B. ORAU. Senior Life Scientist. June 2006.

The only exception to this statement found by the author is the CWT adjustment factor 

(Equation B-4 in Attachment B) for low indices. For indices less than 0.98, the CWT 

adjustment is less than 1.00.

[82] Falk, Roger B. ORAU. Senior Life Scientist. July 2006.

D Plant (Building 991) handled “all materials” as a consequence of its function of shipping, 

receiving, and storage of special nuclear and classified materials for RFP, as well as final 

assembly and inspection of plutonium and EU products in the early years. For more 

information, see Historical American Engineering Record Report CO-83-U for Building 991, 

found on the RFP Web site, http://192.149.55.183/HAER/base/Buildings/991.htm. 

[83] Falk, Roger B. ORAU. Senior Life Scientist. July 2006.

The method codes are listed at the top of the Urinalysis Record Card (see Figures C-1 to C-3 

in Attachment C). Units, if not listed on the card, were discerned from the urine data logs.

[84] Falk, Roger B. ORAU. Senior Life Scientist. July 2006.

This information was discerned from comparing records in the urine data logs with entries on 

Urinalysis Record Cards.

[85] Falk, Roger B. ORAU. Senior Life Scientist. July 2006.

The tolerance levels were noted as the “Working MDL” in some early urine data logs. The 

reporting levels were not stated explicitly in the data logs, but rather were discerned from the 

minimum values calculated in the data logs. These minimum values corresponded to 10% of 

the Working MDL.

[86] Falk, Roger B. ORAU. Senior Life Scientist. July 2006.

This change in the reporting level for the gross alpha results corresponded to the change to 

using plutonium as the default analyte rather than EU. The other changes in this paragraph 

were discerned from the lowest values recorded in the urine data logs.

[87] Falk, Roger B. ORAU. Senior Life Scientist. July 2006.

This statement was based on the examination of the urinalysis records of a number of workers 

affected by this practice.

[88] Falk, Roger B. ORAU. Senior Life Scientist. July 2006.

These reporting levels were discerned from the lowest values recorded in the americium urine 

data logs.

[89] Falk, Roger B. ORAU. Senior Life Scientist. July 2006.

This is a description of the general method. How and when the volume adjustments were 

made for each analyte and period are discussed later in the document.

[90] Falk, Roger B. ORAU. Senior Life Scientist. July 2006.

This is a summary of the observations of the use the recovery determined from the batch spike 

versus a standard recovery value, based on calculations to reproduce the result in the urine 

data logs. Additional discussions are provided for the analytes later in the document.

[91] Falk, Roger B. ORAU. Senior Life Scientist. July 2006.

The adjustment of the volume in this manner could have occurred earlier. However, urine data 

logs for 1955 to 1959 were not found.
Document No. ORAUT-TKBS-0011-5 Revision No. 01 Effective Date: 02/01/2007 Page 43 of 123

[92] Falk, Roger B. ORAU. Senior Life Scientist. July 2006.

This information was provided by Dale L. Bokowski, the lead bioassay chemist at Rocky Flats 

starting in 1961, in an interview with the author in 1992.

[93] Falk, Roger B. ORAU. Senior Life Scientist. July 2006.

Notations indicating detectors with 40% efficiency started to appear in the urine data logs in 

August 1964.

[94] Falk, Roger B. ORAU. Senior Life Scientist. July 2006.

This statement is based on the author’s direct experience and on discussions with Dale L. 

Bokowski prior to his retirement in 1992, who was the lead bioassay chemist at Rocky Flats 

starting in 1961 and also bioassay laboratory manager during the cited period.

[95] Falk, Roger B. ORAU. Senior Life Scientist. July 2006.

The author was directly involved with making this change.

[96] Falk, Roger B. ORAU. Senior Life Scientist. July 2006.

The author was directly involved with a committee of Radiological Health and Analytical Lab 

personnel in 1993 to implement these changes. Upon further review, the author observed that 

the count time was increased to 1,440 min (24 hr) rather than the stated 2,000 min. This 

correction is now made.

[97] Falk, Roger B. ORAU. Senior Life Scientist. July 2006.

This information was provided by Dale L. Bokowski, the lead bioassay chemist at Rocky Flats 

starting in 1961, in an interview with the author in 1992 prior to his retirement from RFP, and 

was verified during examination of the urine data logs for the early 1960s.

[98] Falk, Roger B. ORAU. Senior Life Scientist. July 2006.

These values were based on observations made in the urine data logs concerning when the 

count results were converted to the activity of EU in the sample. Apparently, that decision was 

based on the count uncorrected by volume, for which the minimum reported activity was 

20 dpm/24-hr sample. When a volume adjustment was made, higher minimum reported 

values up to 28 dpm/24-hr sample were observed.

[99] Falk, Roger B. ORAU. Senior Life Scientist. July 2006.

These values were obtained through calculations by the author to duplicate the results stated 

in the urine data logs.

[100] Falk, Roger B. ORAU. Senior Life Scientist. July 2006.

The urine data logs for 1964 to 1971 do not distinguish explicitly which samples were for 

workers in EU areas versus those for workers in DU areas.

[101] Falk, Roger B. ORAU. Senior Life Scientist. July 2006.

This information was discerned by the author from examination of the urine data logs for 

fluorimetric measurements.

[102] Falk, Roger B. ORAU. Senior Life Scientist. July 2006.

This information was discerned by the author from examination of the urine data logs for 

electroplating measurements.
Document No. ORAUT-TKBS-0011-5 Revision No. 01 Effective Date: 02/01/2007 Page 44 of 123

[103] Falk, Roger B. ORAU. Senior Life Scientist. July 2006.

This information was provided by Dale L. Bokowski, the lead bioassay chemist at Rocky Flats 

starting in 1961, in an interview with the author in 1992, prior to his retirement from RFP. 

[104] Falk, Roger B. ORAU. Senior Life Scientist. July 2006.

A better way to indicate the generosity of the nonspecificity of the gross alpha result if applied 

to a specific radionuclide is to use the term “favorable to claimants” rather than “upper 

bounds.”

[105] Falk, Roger B. ORAU. Senior Life Scientist. July 2006.

To have a coherent data set, only background count data for samples counted for 150 min 

were extracted from the urine data logs.

[106] Falk, Roger B. ORAU. Senior Life Scientist. July 2006.

The composite value was used because the detector background appeared to be reasonably 

stable in the 1950s and 1960s, as observed in the previous table. 

[107] Falk, Roger B. ORAU. Senior Life Scientist. July 2006.

This exception was made for americium because the detector backgrounds for the 1950s did 

not apply.

[108] Falk, Roger B. ORAU. Senior Life Scientist. July 2006.

This statement is a summary of the observations of the author during the review of the urine 

data logs.

[109] Falk, Roger B. ORAU. Senior Life Scientist. July 2006.

This approach is consistent with the consideration stated in the fourth bullet in the section 

headed, “Assessment of MDA.”

[110] Falk, Roger B. ORAU. Senior Life Scientist. July 2006.

This subjective observation of the similarities of the recovery values in the preceding table was 

interesting to the author but was not used to determine any MDA value.

[111 Falk, Roger B. ORAU. Senior Life Scientist. July 2006.

a,b] The distribution of volumes for routine 24-hr urine samples was determined from the volumes 

recorded in the urine data logs for gross alpha analyses for 1967 and 1971, a data set of 1,437 

values. The author chose the gross alpha samples as the sample set least likely to include 

special samples that could have had an excretion period of less than 24 hr.

[112] Falk, Roger B. ORAU. Senior Life Scientist. July 2006.

Electrodeposited plutonium and americium samples were marked in the data logs with an “E.” 

No similar designation has been observed by the author in any reports of these urinalysis 

results.

[113] Falk, Roger B. ORAU. Senior Life Scientist. July 2006.

No evidence of a systematic bias in the background or the calibration factor was discerned by 

the author. Therefore, ∆B and ∆K were set equal to zero.

[114] Falk, Roger B. ORAU. Senior Life Scientist. July 2006.

This information was discerned from the urinalysis data logs by the author.
Document No. ORAUT-TKBS-0011-5 Revision No. 01 Effective Date: 02/01/2007 Page 45 of 123

[115] Falk, Roger B. ORAU. Senior Life Scientist. July 2006.

The decision by the author to assess the MDA based on one aliquant was based on the 

observation of the data logs that the decision of detection for the overwhelming majority of the 

samples was based on only one aliquant. Occasionally, the decision was based on the 

average of two aliquants. Because the MDA for one aliquant is higher than that for two 

aliquants, this decision is consistent with the consideration stated at the beginning of the 

“Assessment of MDA” section, fourth bullet.

[116] Falk, Roger B. ORAU. Senior Life Scientist. July 2006.

This statement is based on the personal observations of the author when the recovery of an 

internal tracer was determined for each urine sample in a batch, even in modern times (post1993) for batches of samples processed by commercial laboratories with rigorous QA 

procedures and controls.

[117] Falk, Roger B. ORAU. Senior Life Scientist. July 2006.

This introduction summarizes the information presented in more detail in the body of this TBD. 

Most of this information is based on the direct experience of the author, who provided 

technical support to the operations and developments of the in vivo lung counting systems at 

RFP starting in 1970 and extending into the mid-1980s and also from 1989 to 1992. The 

author also provided second-level management of the dosimetry programs from 1986 to 1989. 

[118] Falk, Roger B. ORAU. Senior Life Scientist. July 2006.

The dates for the start of routine operations of the three counting rooms were determined from 

the references [Room A (Boss and Mann 1967) and Room C (Falk et al. 1979)] or from the 

author discerning the year that the room was first recorded on a worker’s lung count report 

(Room B). 

[119] Falk, Roger B. ORAU. Senior Life Scientist. July 2006.

The information summarized in this section was discerned by the author from (1) notations on 

lung count reports for workers counted in that era, (2) a transition briefing from Robert W. 

Bistline, the physicist providing technical support to the Rocky Flats in vivo measurement 

systems 1966-1969, and (3) reports cited in the reference section. 

[120] Falk, Roger B. ORAU. Senior Life Scientist. July 2006.

The information in this section describes the lung counting program as found by the author in 

1970. The transitions in 1969 were verified by the author in preparation of this report by 

observations of notations in workers’ lung count reports. The transitions did not generally take 

place exactly on January 1, 1969, but usually sometime in 1968, which was a transition year. 

However, for the purpose of determining MDAs, the old practice was considered to be 

extended through the year, and no credit was taken for the new practice until it was in effect 

for the entire year. 

[121] Falk, Roger B. ORAU. Senior Life Scientist. July 2006.

The date of the first entry in the logbook for routine ppm 241Am determinations for incident 

samples was observed by the author to be January 3, 1969. This statement does not preclude 

earlier special ppm 241Am determinations. Indeed, a special ppm 241Am determination was 

made for the October 15, 1965, plutonium fire incident. 
Document No. ORAUT-TKBS-0011-5 Revision No. 01 Effective Date: 02/01/2007 Page 46 of 123

[122] Falk, Roger B. ORAU. Senior Life Scientist. July 2006.

The information in this section describes the systems based on the direct involvement of the 

author and verified by the author from observations of the body count result reports as 

needed.

[123] Falk, Roger B. ORAU. Senior Life Scientist. July 2006.

The information in this section describes the systems as observed by the author, either directly 

or indirectly.

[124] Falk, Roger B. ORAU. Senior Life Scientist. July 2006.

The upgrades in this period were implemented by Eugene Potter and Milan Gadd, technical 

staff supporting the Internal Dosimetry program at Rocky Flats in the 1990s into the 2000s, 

with discussions with the author, who at that time was a customer of the in vivo measurements

program as the internal dosimetrist for the medical monitoring program for former radiation 

workers at RFP. This era is well documented in RFETS (2000b).

[125] Falk, Roger B. ORAU. Senior Life Scientist. July 2006.

The count time, which is the duration of the lung count, was observed by the author from lung 

count reports for workers in this era.

[126] Falk, Roger B. ORAU. Senior Life Scientist. July 2006.

For this equation and for subsequent equations of the calibration factor K, the calibration 

factors were normalized to the Lawrence Livermore National Laboratory (LLNL) phantom [also 

called the Lawrence Livermore Torso Phantom and described in RFETS (2000b)]. 

Normalizing to this phantom is consistent with the approach described in the second 

paragraph in Section B.4 of Attachment B because this phantom was used in “the most recent 

calibration method.”

[127] Falk, Roger B. ORAU. Senior Life Scientist. July 2006.

See section headed 1969 – 1976 and its attribution [4].

[128 Falk, Roger B. ORAU. Senior Life Scientist. July 2006.

a,b] The author personally determined the resolution of the Ortec and Princeton Gamma Tech 

(PGT) detectors for the 59.5-keV photopeak of the 241Am gamma and noted the degradation in 

the resolution. This was expected because the collection efficiency of the charge induced in 

the active part of the detector diminishes with an increasing volume of that active part (based 

on the author’s recollection of solid-state physics).

[129] Falk, Roger B. ORAU. Senior Life Scientist. July 2006.

See Table 5-1 in the body of this TBD. The value of 0.0049 had been historically used at the 

RFP body counter as the rounded value of 0.005.

[130] Falk, Roger B. ORAU. Senior Life Scientist. July 2006.

The method of determining the ppm 241Am from samples representative of the plutonium 

mixtures involved in possible inhalation incidents, starting in 1969, involved the ratio of the 

L X-ray photopeaks and the 241Am 59.5-keV photopeak as measured by a NaI(Tl) detector. 

The ppm 241Am determined by this method was highly uncertain for values less than 100 ppm 

and greater than 10,000 ppm because of the counting statistics. Although one might consider 

zero ppm 241Am to be the true lower bound for freshly purified plutonium, a zero value is not 

practical to use in Equation B-17 (i.e., division by zero is not allowed). The value of 100 ppm 

241Am is also supported by its rank at the 10th percentile in the low-to-high ranking of 442 
Document No. ORAUT-TKBS-0011-5 Revision No. 01 Effective Date: 02/01/2007 Page 47 of 123

values of the incident ppm recorded in the logbook for January 1969 to September 1972. The 

value also represents the ingrowth of americium in freshly purified plutonium (within 0 to 5 mo, 

depending on the efficiency of the purification process).
Document No. ORAUT-TKBS-0011-5 Revision No. 01 Effective Date: 02/01/2007 Page 48 of 123

REFERENCES

Berger, C. D., 1988a, In-Vivo Assessment of Plutonium, Americium, and Uranium at the Rocky Flats 

Plant, IT Corporation/Radiological Sciences Laboratory, Oak Ridge , Tennessee, July 27.

Berger, C. D., 1988b, Assessment of Radioactivity in Wounds at the Rocky Flats Plant, IT 

Corporation/Radiological Sciences Laboratory, Oak Ridge, Tennessee, July 27.

Bistline, R. W., 1968, A Method for Predicting Subject Background for In Vivo Plutonium 

Measurements, RFP-893, The Dow Chemical Company, Rocky Flats Division, Golden, 

Colorado, November 19. 

Boss, M. R., and J. R. Mann, 1967, “Body Counter Evaluations of Plutonium Exposures,” Health 

Physics, Volume 13, pp. 259-266.

Daugherty, N. M., R. B. Falk, F. J. Furman, J. M. Aldrich, and D. E. Hilmas, 2001, “Former Radiation 

Worker Medical Surveillance Program at Rocky Flats,” Health Physics, June 2001, Volume 80, 

Number 6, pp. 544–551. [SRDB Reference ID 22040] 

DOE (U.S. Department of Energy), 1980, Final Environmental Impact Statement Rocky Flats Site, 

Volume 1, DOE/EIS-0064, Washington, D.C.

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

5480.11, Washington, D.C.

Falk, R. B., 1993, Guidelines for Dose Reconstruction for Plutonium Internal Exposures at Rocky 

Flats, Draft Report, EG&G Rocky Flats, Golden, Colorado.

Falk, R. B., 2004, Conversion from Mass Median Diameter to Activity Median Aerodynamic Diameter 

for Plutonium Oxide, Unpublished Report, February 3.

Falk, R. B., W. H. Tyree, C. B. Wood, and C. R. Lagerquist, 1979, “A System of High-Purity 

Germanium Detectors for the Detection and Measurement of Inhaled Radionuclides,” in 

Advances in Radiation Protection Monitoring, IAEA-SM-229/54, pp. 445-457, International 

Atomic Energy Agency, Vienna. 

HPS (Health Physics Society), 1995, An American National Standard – Bioassay Programs For 

Uranium, HPS N13.22-1995.

HPS (Health Physics Society), 1996, An American National Standard – Performance Criteria for 

Radiobioassay, HPS N13.30-1996.

ICRP (International Commission on Radiological Protection), 1959, Report of ICRP Committee II on 

Permissible Dose for Internal Radiation" Publication 2, Pergamon Press, New York, New York.

ICRP (International Commission on Radiological Protection), 1977, Recommendations of the 

International Commission on Radiological Protection, Publication 26, Pergamon Press, Oxford, 

England.

ICRP (International Commission on Radiological Protection), 1979, Limits for Intakes of Radionuclides 

by Workers, Publication 30, Part 1, Pergamon Press, Oxford, England. 
Document No. ORAUT-TKBS-0011-5 Revision No. 01 Effective Date: 02/01/2007 Page 49 of 123

ICRP (International Commission on Radiological Protection), 1994a, Dose Coefficients for Intakes of 

Radionuclides by Workers, Publication 68, Pergamon Press, Oxford, England. 

ICRP (International Commission on Radiological Protection), 1994b, Human Respiratory Tract Model 

for Radiological Protection, Publication 66, Pergamon Press, Oxford, England. 

James, W. H., 1990, “Isotopic Profile of Rocky Flats Plutonium,” interoffice correspondence, EG&G 

Rocky Flats, Golden, Colorado, February 28.

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 V. S. Shirley (editors), 1978, Table of Isotopes, Seventh Edition, Lawrence 

Berkeley Laboratory, University of California, Berkeley, California, and John Wiley & Sons, 

New York, New York.

Mann, J. R., and R. A. Kirchner, 1967, “Evaluation of Lung Burden Following Acute Inhalation 

Exposure to Highly Insoluble PuO2,” Health Physics, Volume 13, pp. 877–882.

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

Implementation Guideline, Rev. 0, Office of Compensation Analysis and Support, Cincinnati, 

Ohio, August.

RFETS (Rocky Flats Environmental Technology Site), 1992, ALPHA QCBLK2.LIS;1, Computer Data 

Report 16-JAN-1992, U.S. Department of Energy, Golden, Colorado.

RFETS (Rocky Flats Environmental Technology Site), 1998a, “Uranium,” Internal Dosimetry Technical 

Basis Documentation, Section 1, Version 3, U.S. Department of Energy, Golden, Colorado.

RFETS (Rocky Flats Environmental Technology Site), 1998b, Statement of Work for Analytical 

Measurements, General Bioassay Service Module BA01-B.4, U.S. Department of Energy, 

Golden, Colorado. 

RFETS (Rocky Flats Environmental Technology Site), 1998c, “Other Isotopes,” Internal Dosimetry 

Technical Basis Documentation, Section 1, Version 3, U.S. Department of Energy, Golden, 

Colorado.

RFETS (Rocky Flats Environmental Technology Site), 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 Site), 2000a, Report on the Flow of Recycled 

Uranium at the Department of Energy’s Rocky Flats Plant 1953—1993, U.S. Department of 

Energy, Golden, Colorado, June 30.

RFETS (Rocky Flats Environmental Technology Site), 2000b, “In-Vivo Radiobioassay,” Internal 

Dosimetry Technical Basis Documentation, Section 3, Version 8, U.S. Department of Energy, 

Golden, Colorado.
Document No. ORAUT-TKBS-0011-5 Revision No. 01 Effective Date: 02/01/2007 Page 50 of 123

RFETS (Rocky Flats Environmental Technology Site), 2002, “Plutonium and Americium,” Internal 

Dosimetry Technical Basis Documentation, Section 1, Version 6, U.S. Department of Energy, 

Golden, Colorado.
Document No. ORAUT-TKBS-0011-5 Revision No. 01 Effective Date: 02/01/2007 Page 51 of 123

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 × 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. 

L X-rays 

Low-energy X-rays produced during radioactive decay. 

maximum permissible lung burden (MPLB)

The occupational limit for plutonium expressed in terms of a quantity of plutonium that could 

be present in the pulmonary lungs at any given time.

minimum detectable activity (MDA)

Limit of radionuclide activity detection for measurements of specific types and energies of 

radiation. 
Document No. ORAUT-TKBS-0011-5 Revision No. 01 Effective Date: 02/01/2007 Page 52 of 123

photon

A unit or particle of electromagnetic radiation; photons originating from the nucleus or extranuclear material of an atom are called respectively gamma rays or X-rays. 

radiation

Alpha, beta, neutron, and photon radiation. 

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 (1,000 mg/cm2

); 

however, also used to refer to the recorded dose.

X-ray

Ionizing electromagnetic radiation of external nuclear origin.
Document No. ORAUT-TKBS-0011-5 Revision No. 01 Effective Date: 02/01/2007 Page 53 of 123

Roger B. Falk

Senior Life Scientis t

Oak Ridge As s ociated Univers ities

Originally s ubmitted as draft on J une 5, 2003

TABLE OF CONTENTS

A.1 Introduction..............................................................................................................................54

A.2 MDA Methodology...................................................................................................................54

A.3 History of Methods...................................................................................................................55

A.4 Assessment of MDA................................................................................................................61

A.5 MDA Values ............................................................................................................................66

A.6 Discussion ...............................................................................................................................73

A.7 Acknowledgments ...................................................................................................................73

ATTACHMENT A

MINIMUM DETECTABLE ACTIVITY FOR URINALYSIS METHODS AT ROCKY FLATS

Page 1 of 20
Document No. ORAUT-TKBS-0011-5 Revision No. 01 Effective Date: 02/01/2007 Page 54 of 123

A.1 INTRODUCTION

Urinalysis has been used at RFP 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, EU, or 

DU. 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 RFP 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 as possible in the final medium to be measured (counted). 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 DU. The assessment of the 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 RFP from 1952 to the present. The analytes of interest are plutonium, americium, 

EU, and DU. 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 RFP. 

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 + 2k so + 3) ÷ KT (A-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

] (A-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

ATTACHMENT A

Minimum Detectable Activity for Urinalysis Methods at Rocky Flats

Page 1 of 20

ATTACHMENT A

MINIMUM DETECTABLE ACTIVITY FOR URINALYSIS METHODS AT ROCKY FLATS

Page 2 of 20
Document No. ORAUT-TKBS-0011-5 Revision No. 01 Effective Date: 02/01/2007 Page 55 of 123

K = calibration factor

∆K = the maximum fractional systematic error bound in the calibration factor K 

∆B = the maximum expected fractional systematic 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 counting time for the procedure.

Applying this equation to urinalysis methods at RFP involves determining the value of each variable 

for measurements of the analytes: plutonium, americium, EU, DU, and gross alpha, as the methods 

evolved.

A.3 HISTORY OF METHODS

General Information

In the beginning of operations (1952), RFP was divided into four distinct subplants plus a general 

support area. The subplants 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, DU 

is A material processed in A Plant (buildings numbered 4##, mainly Building 444); EU 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 [82]. 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 EU operations were phased out at RFP, although urinalysis monitoring for EU 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 those data for a specific worker. A Urinalysis Record Card was 

established for each monitored worker and included 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 RFP record of dosimetry information for a 

worker. Table A-1 lists the method codes [83].

Table A-1. Method codes.

Code Meaning

A Fluorimeter, reported in μg/L (1952–1956); reported in μg/24 hr (1957–1964)

B1 Electroplating, reported in dpm/24 hr. (Note: Electroplating, in RFP records, more properly should be 

called electrodeposition.)

B2 Ether extraction, reported in dpm/24 hr

B3 TBP extraction (hand-written on some cards)

C1 Carrier precipitation, reported in dpm/24 hr

C2 TTA extraction, reported in dpm/24 hr. (Note: On the header of cards for 1961–1965, the code C2 is 

“Pu by Radio Autography.” There is no indication that this method was implemented at RFP.)

D TBP extraction

ATTACHMENT A

MINIMUM DETECTABLE ACTIVITY FOR URINALYSIS METHODS AT ROCKY FLATS

Page 3 of 20
Document No. ORAUT-TKBS-0011-5 Revision No. 01 Effective Date: 02/01/2007 Page 56 of 123

Although there is some correlation of the codes with the subplants, there are some exceptions. 

Table A-2 summarizes the correlation of the method code and the analyte [84].

Table A-2. Correlation of method code 

and analyte. 

Analyte Method code

Depleted uranium A, B1 (starting 5/1/64)

Enriched uranium B1

Plutonium C1, C2

Gross alpha B2, B3, D

Tolerance levels were used at RFP 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 RFP 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. Table A-3 lists the values of the tolerance levels [85]:

Table A-3. Values of tolerance and reporting levels

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 DU and EU, through 

1961 for plutonium, and through 1963 for gross alpha. From May 1964 through 1971, the reporting 

level for DU and EU was ≥20 to 28 dpm/24 hr. After 1963, the reporting level for gross alpha was 

≥0.9 dpm/24 hr [86]. 

For plutonium, the reporting and recording level was ≥0.2 dpm/24 hr for 1962 to 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. This retroactive 

application was variable with respect to how far back it was applied [87].

In 1963, a specific analysis for 241Am was implemented. The recording and reporting level for 241Am 

was ≥0.24 dpm/24 hr in 1963, ≥0.2 dpm/24 hr from 1964 to 1967, and ≥0.3 dpm/24 hr from 1968 to 

1971 [88].

The general method for data analysis for alpha counting procedures (1952 to 1971) was:

Activity (dpm/24-hr sample) = (C/T – BDet – BBlk) × (V/A) / (ε × R) (A-3)

where:

 C = Total count 

T = Count time (min) 

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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.

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 [89]. The sample volume was recorded in the urinalysis data 

log for each sample.

The value of ε was the geometry rating of the detector. In 1952 and 1953, ε was 0.45. After that, the 

detectors were called 50% detectors, and ε was 0.50. In 1964, 40% detectors (ε = 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 [90].

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 × R was frequently combined, especially in the 1950s. In the 1960s, the term 1/(ε × 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 (A-4)

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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 = 1,000 mL.]

The history of these urinalysis methods is largely based on an interview with the lead chemist for the 

bioassay program from 1961 to 1992, Dale L. Bokowski, in 1992 prior to his retirement from RFP, and 

on a review of the bioassay data logs from 1952 to 1971.

Plutonium

1952 to 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 min. 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 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 have been used to establish the 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 [91]. 

1961 to 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 [92]. No other changes were made to the 

previous method.

1963 to 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 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 to 1971. In 1973, an alpha PHA 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 0.4 were 

used as a supplement to the detectors with 0.5 efficiency [93]. 

1978 to 1993

By 1978, all of the counting systems 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 

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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 the acceptable range [94]. In 1990, the acceptable recovery range was changed to 0.35 to 

1.10 [95]. The count time of 720 min 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 20 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 of 0.25 to 0.35.

1993 and beyond

Upgrades to procedures occurred in 1993 in order to achieve a process MDA less than or equal to 

0.020 dpm/sample [96]. Count time was increased to 1,400 min. 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 an MDA ≤0.02 dpm/sample be achieved. In 

1997, the onsite bioassay laboratory was shut down. 

Americium (1963 and beyond)

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 [97]. 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 to 1971)

Urine samples were analyzed for EU according to a process called electroplating. A 50-mL aliquant 

of urine was extracted from the 24-hr sample and 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 min. 

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 

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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 [98]. 

Blank data were not used to adjust the sample count rate, except sporadically in 1963 and 1964. 

Detector background was usually subtracted, but not always. Spike samples were processed, 

although it is not obvious how those 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) [99]. 

EU operations were phased out at RFP from 1962 to 1963, although some workers were still 

monitored for EU intakes through 1971 [100].

Depleted Uranium (1952 to 1971)

Two methods were used to analyze urine samples for DU. 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 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 [101].

After April 1964, the urine sample was analyzed using the electroplating procedure described above 

for EU, and the results were reported in dpm/24-hr sample (or background) [102].

Gross Alpha (1952 to 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 nonspecific in extracting plutonium and uranium as well as 

americium and natural thorium [103]. 

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 min, 

although from 1952 to 1955 count times of 55, 60, and 75 min, and in 1971 count times of 40 and 60 

min, 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 EU or to plutonium, or a portion to each. The 

default condition, through 1963, was to credit the result to EU unless the PHA count indicated 

otherwise. After 1963 (when EU operations were phased out), the default condition was to credit the 

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result to plutonium. In either case, the results should be considered upper bounds because of the 

nonspecificity of the analysis [104].

A.4 ASSESSMENT OF MDA

General Considerations

The MDA is assessed for plutonium, americium, EU, DU, and gross alpha, based on Equation A-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. If 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 regard 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 RFP 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. 

Table A-4 lists sample volumes for routine 24-hr urine samples. 

Table A-4. Sample volumes for 

routine 24-hour urine samples.

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 1960 to 1971. For some years in these periods, logs for 

only a part of the year were available.

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Data for Alpha Counting Systems

Table A-5 lists 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 a sample count time 

of 150 min [105]:

Table A-5. Detector background for gas flow proportional counters.

Detector background (cpm)

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 about the count time used to measure the detector background, but the 

count time is likely to be 150 min 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 [106]. For americium, the values for the 1960s are used because the americium 

process was not implemented in the 1950s [107]. 

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 [108].

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. Table A-6 summarizes the median and 95th-percentile blank count rates. 

Table A-6. Median and 95th-percentile blank count rates.

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 (as 

reviewed by R. Falk, 2003). 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 EU 

and DU 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 Table A-6 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 × T) (A-5)

The values for B, sB0, sB1, and so in the MDA equation (A-1) are derived from the detector background 

and blank values:

B = (BDet + BBlk) × T (A-6)

sB0 = T × √ ( sDet

2 + sBlk2

) (A-7)

sB1 = √ B (A-8)

so = √ (sB1

2 + sB02

) (A-9)

For some analytes (EU, DU) and periods, the detection decision was based on the average of two 

aliquants. In this case, the value of so for the average of two aliquants is equal to the value of so for 

one aliquant divided by the square root of 2.

The value of ∆B is taken to be zero. This variable could 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. Table A-7 lists the efficiencies of the 

detectors (as noted above):

Table A-7. Efficiencies of alphacounting detectors.

Period Detector efficiency

1952–1953 0.45

1954–1963 0.50

1964–1971 0.40 and 0.50

For 1964 to 1971, the value of 0.40 is used as the efficiency for the MDA calculation [109].

Table A-8 lists the recoveries used in the MDA assessment, which are taken to be the median 

recovery and the 5th-percentile (low to high) value discerned from the spike data for the process. 

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Table A-8. Recoveries used in MDA assessments.

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 

plutonium from 1963 to 1971, the values are based on data for 1963 to 1965 and 1969 to 1971. For 

EU and DU, 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, to 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 [110].

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 [111a]. 

For EU and DU (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 [111b]. 

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. Table A-9 lists the values of 

fabs, based on judgments of experienced bioassay chemists, for the extreme (95th-percentile) 

condition, and the corresponding values of Fa. 

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Table A-9. Fraction of alphas absorbed in residue.

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 [112]. 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).

The calibration factor K is the product of ε, R, Vf, and Fa: 

K = ε R Vf Fa (A-10)

The values of ∆B and ∆K are considered to be zero [113].

Data for Fluorimetric Mass Measurements

Applying the MDA equation (A-1) to fluorimetric 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 (A-11)

The value of sB0 was determined from a review by Roger Falk 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 U/24-hr sample (see 

Equation A-4). In 1955, the calibration factor was applied to the uncorrected net reading. In 1960 and 

later, the calibration factor was applied to the corrected reading, which was the net reading multiplied 

by the factor 1.15 [114]. 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, as listed in Table A-10.

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Table A-10. Gross alpha 

calibration factor

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.

A.5 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 

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 min is used for all assessments.

Table A-11 lists the values of the variables and the median MDA (dpm/24-hr sample).

Table A-11. Values of variables and MDA for plutonium for median 

conditions.

Period

Values of the variables

MDA

(dpm per 24-hr 

so ε R Vf Fa sample)

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

Table A-12 lists 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. 

Table A-12. Values of variables and MDA for plutonium for extreme 

conditions.

Period

Values of the variables

MDA

(dpm per 24-

so ε R Vf Fa hr sample)

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.

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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. Table A-13 lists the MDA for each of the 

extreme conditions individually, as well as for combinations of two and three extreme conditions.

Table A-13. MDA for plutonium for one, two, or 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 EU 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 min is used for MDA assessments from 1952 to

1963. For 1964 to 1969, the count time of 30 min is used and, for 1970 to 1971, the count time of 40 

min 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.

Table A-14 lists the values of the variables and the median MDA (dpm/24-hr sample).

Table A-14. Values of variables and MDA for EU for median 

conditions 

Period

Values of the variables MDA

(dpm per 24-

so ε R Vf Fa hr sample)

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

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Table A-15 lists 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. 

Table A-15. Values of variables and MDA for EU for extreme 

conditions

Period

Values of the variables

MDA

(dpm per 24-

so ε R Vf Fa hr sample)

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. Table A-16 lists the MDA for each of the 

extreme conditions individually, as well as for combinations of two and three extreme conditions.

Table A-16. MDA for EU for one, two, or 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 DU 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 [115]. In Table A-17, the MDA at the extreme condition is based on the 95th-percentile 

volume.

Table A-17. Values of variables and MDA for fluorimetric 

measurements of DU for median and extreme conditions.

Period

Values of the variables MDA (μg/24-hr sample)

sB0

Median 

K

Extreme 

K 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 and alpha counting measurements, the MDA values tabulated for EU for 

1964 to 1971 apply also to DU.

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 min is used for assessments from 1963 

to 1970. In 1971, the typical (and minimum) count time is 60 min.

Table A-18 lists the MDA to two significant figures.

Table A-18. Values of variables and MDA for americium for 

median conditions.

Period

Values of the variables MDA

(dpm per 24-hr 

so ε R Vf Fa sample)

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

Table A-19 lists 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. 

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Table A-19. Values of variables and MDA for americium for 

extreme conditions

Period

Values of the variables MDA

(dpm per 24-hr 

so ε R Vf Fa sample)

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. Table A-20 lists the MDA for each of the 

extreme conditions individually, as well as for combinations of two and three extreme conditions.

Table A-20. Values of the MDA for americium for one, two, 

or three extreme conditions. 

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 min is used for 1952, 

75 min for 1953 to 1959, and 150 min 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 min is also used for the 

extreme condition. See Table A-21.

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Table A-21. Values of variables and MDA for gross alpha 

measurements for median conditions.

Period

Values of the variables MDA

(dpm per 24-hr 

so ε R Vf Fa sample)

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

Table A-22 lists 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. 

Table A-22. Values of variables and MDA for gross alpha 

measurements for extreme conditions 

Period

Values of the variables MDA

(dpm per 24-hr 

so ε R Vf Fa sample)

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 = 40 min) 5.18 0.40 0.24 0.58 0.7 13

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. Table A-23 lists the MDA for each of the 

extreme conditions individually, as well as for combinations of two and three extreme conditions.

Table A-23. Values of the MDA for gross alpha 

measurements for one, two, or 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 = 40 min) 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

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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 = 40 min) 5.2 3.8 3.1 5.8 4.8 3.5
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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 = 40 min) 9.0 7.5 5.4 8.3

A.6 DISCUSSION

The MDA is an a priori concept that can 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 before the processing of the sample result, and that the information 

is used conceptually to determine the subpopulation 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 in this attachment represent overall process MDAs 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 are known. The sample-specific parameter 

values, except recovery, are generally recorded in the urine data logs, but not all of the urine data logs 

have been found and some might not have been archived.

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 or extreme 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 fraction was the 

variable that had the most influence on the sample-specific MDA.

A.7 ACKNOWLEDGMENTS

The author would like to acknowledge the invaluable insights into the historical bioassay methods and 

procedures at RFP provided by Dale L. Bokowski, the lead bioassay chemist at RFP from 1961 to 

1992. The author also acknowledges the valuable assistance of Michael Prochownik, Nancy 

Chandler, and Kimberly Linnemeyer, manager and staff of the Radiological Records group at RFP, for 

their assistance in finding and retrieving archived urinalysis data logs so this analysis could be 

achieved.

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Roger B. Falk

Senior Life Scientist

Oak Ridge Associated Universities

Originally submitted as draft on January 9, 2003

TABLE OF CONTENTS

B.1 Introduction .............................................................................................................................75

B.2 MDA Methodology...................................................................................................................75

B.3 History of Counting Systems and Procedures.......................................................................... 76

B.4 Assessment of MDA................................................................................................................79

B.5 Values of the Variables, 1969 – for NaI(Tl) and Phoswich Detector Systems .......................... 81

B.6 Values of the Variables, 1976 – for Ortec Germanium Detector Systems ................................ 82

B.7 Values of the Variables, 1978 – for PGT I Germanium Detector Systems ............................... 84

B.8 Values of the Variables, 1979 – for PGT II Germanium Detector Systems .............................. 85

B.9 Values of the Variables, 1985 – for PGT Organ Pipe Germanium Detector Systems............... 86

B.10 Values of the Variables, 1985 – for EG&G Organ Pipe Germanium Detector Systems............ 87

B.11 MDA for RFP Plutonium .......................................................................................................... 88

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B.1 INTRODUCTION [117] 

In vivo lung counts have been performed at RFP 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 (Boss and 

Mann 1967). Plutonium was not detected directly because of the low abundance of gamma photons 

and because of the severe attenuation of the more abundant, low-energy L X-rays. Instead, the 

59.5-keV gamma photon from 241Am was used as a surrogate. The isotope 241Am was present to 

some extent in all WG plutonium at RFP. 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 RFP, 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 RFP. The RFP 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 (B-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

] (B-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 systematic error bound in the calibration factor K

∆B = the maximum expected fractional systematic 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 RFP involves determining the value of each of these 

variables for the counting systems and procedures used at RFP as the systems and procedures 

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evolved. The MDA for in vivo measurements is necessarily individual-specific because the 

detectability of 241Am in the chest is a significant function of the CWT of the subject. 

The MDA can also be determined empirically from replicate measurements on an appropriate blank. 

This approach is used for the systems starting in 1995 at RFP.

B.3 HISTORY OF COUNTING SYSTEMS AND PROCEDURES

The in vivo lung counting systems at RFP consisted of photon detectors mounted in a shielded room 

(6-in.-thick low-background steel lined with layers of lead, tin, and zinc) with electronic equipment 

(amplifiers and multichannel 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 [118] 

Each room was equipped with a detector system. When a new detector system was implemented, 

the previous system was usually maintained as a backup system. As a result, end dates for use of a 

given detector system are not known. In the era of the germanium 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 to 1968 [119] 

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 in. and was 4 mm thick with a surface area of 80 cm2

. These detectors were 

known as the 4×4 detectors. In most situations, the detectors were configured with one detector 

above the left portion of the upper chest; the second detector was over the liver and gut region. The 

chest detector was sometimes placed over the right portion of the upper chest rather than the left 

position. In other cases, both detectors were placed over the chest. The chest detectors were placed 

in a framework called a jig to allow a standard and reproducible position for all subjects. Count time 

was either 40 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 REMAB phantom, manufactured by Alderson Research 

Laboratories, Inc. No adjustment was made for CWT.

1969 to 1976 [120]

During this period, two counting rooms were operational with three 4- by 4-in. 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). 

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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 s (1,000 s for expedited counts).

4. The method of the matched subject background based on 137Cs and 40K was replaced by the 

index method. 

The index method had the following features (Bistline 1968): 

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 [a detector with a primary scintillation Nal(TI) layer backed by a 

Csl layer for coincidence counting] was implemented and used intermittently into the 1980s. The 

NaI(Tl) layer of the phoswich detectors was dimensionally the same as the 4- by 4-in. 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 plutonium L X-rays was not 

successfully implemented at RFP.

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 [121], the ppm 241Am was measured routinely 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 germanium detector system was investigated but was never 

implemented.

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1976 to 1985 [122]

This period is the era of the high-purity germanium detector array systems. Three counting rooms 

were operational. When the germanium 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 germanium 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 germanium system implementation timeline was:

1976 Ortec detectors, 10 cm2 per detector, two arrays

1977 PGT I detectors, 15 cm2 per detector, two arrays

1979 First array, PGT II detectors, 18 cm2 per detector

1980 Second array, PGT II detectors, 18 cm2 per detector

Other changes in this period were:

1. The calibration factor for the germanium systems was adjusted for the CWT of the subject. 

The thickness (centimeters) 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 LLNL phantom starting in 1979.

4. The method of determining the background changed for the germanium 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 to 1995 [123] 

In this period, germanium detectors in an 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 germanium system implementation timeline was:

1985 PGT organ pipe detectors, 20 cm2 per detector 

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1991 EG&G Ortec organ pipe detectors, 20 cm2 per detector

No other significant changes were made during this period.

1995 to 2005 [124]

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 that was 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 germanium detectors were replaced by EG&G 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 (RFETS 2000b) was the equation to determine the CWT. ABACOS-Plus

incorporates the equation developed at LLNL: 

CWT (cm) = 1.973 (W/H) – 2.0038 (B-3)

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) (B-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 RFP. Equation B-4 can also be applied as a divisor to historical 

calibration factors for previous systems at RFP.

B.4 ASSESSMENT OF MDA

The value of the MDA for 241Am is assessed here for each detector system and for each significant 

change in the procedure. It is assessed not only for the typical RFP 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 by the author of this attachment via review of available records 

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. As an alternative, the product of the factors, for the appropriate period, can be used in place 

of the term (1 + ∆K) in Equation B-1. Table B-1 lists the discontinuity factors. 

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Table B-1. Discontinuity factors.

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 LLNL 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) (B-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 [125] 

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

∆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 comparison of the net subject count rate (after 

subtraction of room background) 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 LLNL phantom 

[126] and incorporating the discontinuity factors (Equation B-5, 1.30, and 1.45) is given by:

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K = 55.13 exp(-0.2359 (2 I – 0.1)) ÷ exp(0.635 I) (B-6)

The calibration factor for the system with only one detector over the left portion of the chest is given 

by Equation B-6 multiplied by 0.43. This factor is the fraction of the total activity in the calibration 

lungs of the RFP LLNL phantom that is in the 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 LLNL phantom and the discontinuity factors are 

incorporated into K, the value of ∆K is taken to be zero. Because the term (1 + ∆K) in Equation B-1 is a 

multiplier to the MDA and because 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 attachment 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 

= 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 s or 2,000 s [127] 

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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 s 

B = 2,200 for T = 2,000 s 

∆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 s 

R = 1,667 for T = 2,000 s 

sB1

2

= Total subject count + R = B + 2R 

= 2,767 for T = 1,000 s 

= 5,533 for T = 2,000 s 

sB0

2

= (0.1B)2

= 12,100 for T = 1,000 s 

= 48,400 for T = 2,000 s 

m = 1 

so = 121.9 for T = 1,000 s 

= 232.2 for T = 2,000 s 

The 241Am calibration factor K for two detectors, normalized to the calibration with the LLNL phantom 

and incorporating the discontinuity factors (Equation B-5 and 1.30) is given by:

K = 79.94 exp(-0.2359 (2 I – 0.1)) ÷ exp(0.635 I) (B-7)

K = 30.23 for I = 0.90

= 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 s or 2,000 s 

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 germanium systems. 

∆B = 0 

m = 8 

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For eight detectors:

B = 341 for T = 1,000 s (unadjusted by m = 8)

B = 682 for T = 2,000 s (unadjusted by m = 8)

For five detectors:

B = 213 for T = 1,000 s (unadjusted by m = 8)

B = 427 for T = 2,000 s (unadjusted by m = 8)

For the calculation of sB1, the subject background is B/8.

For eight detectors:

T = 1,000 s: 

sB1 = 6.53 sB0 = 18.5 s0 = 6.93

T = 2,000 s: 

sB1 = 9.23 sB0 = 26.1 s0 = 9.79

For five detectors: 

T = 1,000 s:

sB1 = 5.17 sB0 = 14.6 s0 = 5.48

T = 2,000 s:

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 LLNL phantom and incorporating the discontinuity factors [Equation B-5 and 1.30 (for pre1979 systems)] is given by:

K = 24.12 exp(-0.3398 (2 I – 0.1)) ÷ exp(0.635 I) (B-8)

and, for Ortec systems 1979 and following:

K = 31.36 exp(-0.3398 (2 I – 0.1)) ÷ exp(0.635 I) (B-9)

For the minimum system of five detectors, adjust the calibration factor by multiplying by (5/8).

Table B-2 lists the calibration factors for the Ortec germanium detector system.

Table B-2. Calibration factors for the Ortec 

germanium detector system.

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

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B.7 VALUES OF THE VARIABLES, 1978 – FOR PGT I GERMANIUM DETECTOR SYSTEMS

The PGT I germanium system is basically the same as the Ortec germanium system. The main 

difference is that the PGT I detectors had a larger surface area but a poorer resolution [128a]. 

∆B = 0 

m = 4

For eight detectors:

B = 240 for T = 1,000 s (unadjusted by m = 4)

B = 480 for T = 2,000 s (unadjusted by m = 4)

For five detectors:

B = 150 for T = 1,000 s (unadjusted by m = 4)

B = 300 for T = 2,000 s (unadjusted by m = 4)

For the calculation of sB1, the subject background is B/4.

For eight detectors:

T = 1,000 s: 

sB1 = 7.75 sB0 = 15.5 s0 = 8.67

T = 2,000 s:

sB1 = 10.95 sB0 = 21.9 s0 = 12.2 

For five detectors:

T = 1,000 s:

sB1 = 6.12 sB0 = 12.2 s0 = 6.84

T = 2,000 s:

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 LLNL phantom and incorporating the discontinuity factors [Equation B-5 and 1.30 (for pre1979 systems)] is given by:

K = 34.09 exp(-0.3292 (2 I – 0.1)) ÷ exp(0.635 I) (B-10)

and, for PGT I systems 1979 and following:

K = 44.318 exp(-0.3292 (2 I – 0.1)) ÷ exp(0.635 I) (B-11)

For the minimum system of five detectors, adjust the calibration factor by multiplying by (5/8).

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Table B-3 lists calibration factors for the PGT I germanium detector system.

Table B-3. Calibration factors for the PGT I 

germanium detector system 

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 germanium 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 [128b]. 

∆B = 0 

m = 4

For eight detectors:

B = 273 for T = 1,000 s (unadjusted by m = 4)

B = 546 for T = 2,000 s (unadjusted by m = 4)

For five detectors:

B = 170 for T = 1,000 s (unadjusted by m = 4)

B = 341 for T = 2,000 s (unadjusted by m = 4)

For the calculation of sB1, the subject background is B/4.

For eight detectors:

T = 1,000 s: 

sB1 = 8.26 sB0 = 16.5 s0 = 9.23

T = 2,000 s:

sB1 = 11.7 sB0 = 23.4 s0 = 13.1

For five detectors:

T = 1,000 s:

sB1 = 6.53 sB0 = 13.1 s0 = 7.31

T = 2,000 s:

sB1 = 9.23 sB0 = 18.5 s0 = 10.3

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The 241Am calibration factor K for two arrays with a total of eight detectors (incorporating 

Equation B-5), is given by:

K = 38.65 exp(-0.3579 (2 I – 0.1)) ÷ exp(0.635 I) (B-12)

For the minimum system of five detectors, adjust the calibration factor by multiplying by (5/8).

Table B-4 lists calibration factors for the PGT II germanium detector system.

Table B-4. Calibration factors for 

the PGT II germanium detector 

system. 

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 germanium system is basically the same as the previous germanium array

systems. The main difference is the ability to maintain a stable, standard configuration with six 

detectors. 

∆B = 0 

m = 4

Table B-5 lists the values of variables for the PGT organ pipe germanium detector system.

Table B-5. Values of variables for the PGT 

organ pipe germanium detector system 

T = 1,000 s T = 2,000 s

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 B-5) 

is given by:

K = 34.32 exp(-0.2946 (2 I – 0.1)) ÷ exp(0.635 I) (B-13)

Table B-6 lists calibration factors for the PGT organ pipe germanium detector system.

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Table B-6. Calibration factors 

for the PGT organ pipe 

germanium detector system.

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 germanium system is basically the same as the previous PGT organ pipe 

germanium array system. 

∆B = 0 

m = 4

Table B-7 lists the values of variables for the EG&G organ pipe germanium detector system. 

Table B-7. Values of variables for the EG&G 

organ pipe germanium detector system.

T = 1,000 s T = 2,000 s

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 B-5, 

is given by:

K = 42.36 exp(-0.3708 (2 I – 0.1)) ÷ exp(0.635 I) (B-14)

Table B-8 lists calibration factors for the EG&G organ pipe GE detector system.

Table B-8. Calibration factors 

for the EG&G organ pipe 

germanium detector system.

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 system at RFP was not determined analytically using Equation B-1. Instead, the 

MDA was determined empirically from replicate measurements on an appropriate blank that simulated 

the counts of the average RFP worker (CWT = 3.36 cm). Therefore, there are no values of the 

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variables to be listed here. The value of the MDA for the average RFP 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) (B-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 γ/min per nCi 

241Am. The derived calibration equation is:

K = 35.9 exp(-0.41 CWT) (B-16)

The MDA for any value of CWT is then obtained from the product of 0.3 nCi (the MDA for the average 

RFP worker) and the ratio (9.05/K for the value of the CWT).

B.11 MDA FOR RFP PLUTONIUM

The MDA for RFP 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 = (1 × 106

 – ppm 241Am) ÷ (48.2 ppm 241Am) (B-17)

Table B-9 lists MDA conversion factors for some typical values of ppm 241Am. 

Table B-9. MDA conversion 

factors for values of ppm 241Am.

ppm 

Am-241

MDA conversion 

factors

100 207

1,000 20.7

10,000 2.05

The task is to determine the value of the ppm 241Am at the time of the lung count. The practice at 

RFP 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 

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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 RFP plutonium is taken to be 0.005 in the 1950s and 1960s and

0.0036 in the 1970s and later [129]. Table B-10 lists 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 can 

be taken as the lower bound [130] and represents freshly purified plutonium.

Table B-10. Americium ingrowth in RFP plutonium.

Years

Am-241 ppm at time (yr) after intake

Initial fraction Pu-241 = 0.0036 Initial fraction Pu-241 = 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)] (B-18)

where

A = ppm 241Am at time T (years)

L1 = λPu241 ÷ (λAm241 – λPu241) 

λPu241 = decay constant for 241Pu (half-life = 14.4 yr) = 0.0481

λAm241 = decay constant for 241Am (half-life = 433 yr) = 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).

Table B-11 summarizes the americium MDAs for RFP in vivo lung counts.

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Table B-11. Americium MDA for in vivo lung counts at RFP.

Period Detector system Index

MDA (nCi) for Am-241

Comments

Minimum system Standard system

Half time Full time Half time Full time

1964–1968 NaI(Tl) 4×4 0.90 1.7 1.5 1.3 1.2 Full time = 40 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) 4×4 0.90 0.80 0.76 Full time = 2,000 s.

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 = 2,000 s.

Standard system is two detectors, over right and left chests.

NaI sensitive layer is the same as the NaI 4×4.

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 = 2,000 s.

Standard system is eight detectors in two arrays.

Minimum system is five detectors in two 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 germanium 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 germanium 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 germanium 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 germanium 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 = six detectors.

Standard count time = 2,000 s. 

Occasionally, five or seven 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 = six detectors.

1.35 0.26 0.18 Standard count time = 2,000 s. 

1.80 0.48 0.33

1995– Ortec 2 Organ 

Pipe Ge 

Detectors

0.90 0.14 Standard system = four detectors.

1.35 0.3 Standard count time = 2,000 s. 

1.80 0.6

Note: Starting in 1978, hybrid germanium systems were used that combined two different germanium arrays or detector types. For hybrid systems, use the higher of the MDA values

for the detector types involved.

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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 RFP 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 239,240Pu, multiply the MDA for 241Am by [(1 × 106

 – ppm 241Am) ÷ (48.2 × 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. 

Americium-241 grows into the plutonium mixture from the nuclear transformation of 241Pu. The initial weight fraction of 241Pu in RFP plutonium is 0.0050 in the 1950s and 1960s and 

0.0036 in the 1970s and 1980s. For freshly purified plutonium, with a residual of approximately 100 ppm 241Am, the ppm 241Am would be 270 to 340 after the 1st year, 430 to 560 after 

the 2nd year, 730 to 980 after the 4th year, 1,000 to 1,400 after the 6th year, 1,500 to 2,000 after the 10th year, 2,300 to 3,100 after the 20th year, 2,800 to 3,800 after the 30th year, 

3,000 to 4,200 after the 40th year, and 3,200 to 4,400 after the 50th year. 

Halved count times were usually used for nonscheduled counts or when a large number of subjects needed to be counted expeditiously.

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MINIMUM DETECTABLE ACTIVITY FOR IN VIVO LUNG COUNTS AT ROCKY FLATS

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LIST OF FIGURES

FIGURE TITLE

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

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

PAGE

93

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

activity date on the HSDS portion 8-19-53) ............................................................................94

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

activity date on the HSDS portion: 1-6-58) ............................................................................95

C-4 Health Sciences Data System – Urinalysis Detail (1) (first activity date 9-17-58).................... 96

C-5 Health Sciences Data System – Urinalysis Detail (2) (first activity date 3-19-73).................... 97

C-6 Analytical Report – Bioassay Analysis Data 3-15-93..............................................................98

C-7 Analytical Report – Bioassay Analysis Data 10-28-93 ............................................................99

C-8 Form 1 – Sample Results 1-29-96........................................................................................100

C-9 Rocky Flats Environmental Technology Site (1) 8-27-96 (analytes: 238U, 235U, 234U)............ 101

C-10 Rocky Flats Environmental Technology Site (1) 8-8-96 (analyte: 239Pu) .............................. 102

C-11 Form 1 – Sample Results – Quanterra, Richland 7-31-98 .................................................... 103

C-12 General Engineering Laboratories, Inc. 6-28-99................................................................... 104

C-13 Health Sciences Urinalysis Record (with tritium, fecal, and nasal smear results) ................. 105

C-14 Health Physics – Body Counter Information 12-8-65............................................................ 106

C-15 Health Physics – Body Counter Information 5-16-68............................................................ 107

C-16 Health Physics – Body Counter Information 8-26-68............................................................ 108

C-17 Health Physics – Body Counter Information 9-16-70............................................................ 109

C-18 Radiation Dosimetry – Body Count Results 10-3-74............................................................. 110

C-19 Radiation Dosimetry – Body Count Results 5-30-75............................................................. 111

C-20 Radiation Dosimetry – Body Count Results 1-9-78............................................................... 112

C-21 Body Counter Results 12-8-81 .............................................................................................113

C-22 Radiation Dose Assessment – Body Count Results 7-22-83................................................ 114

C-23 Radiation Dose Assessment – Body Count Results 5-18-83................................................ 115

C-24 Radiation Dose Assessment – Body Count Results 2-21-84................................................ 116

C-25 Radiation Dose Assessment – Body Count Results 3-22-84................................................ 117

C-26 Radiation Dose Assessment – Body Count Results 10-10-85 .............................................. 118

C-27 Radiation Dose Assessment – Body Count Results 3-6-89.................................................. 119

C-28 Internal Dosimetry – Lung Count Results 11-23-93.............................................................. 120

C-29 ABACOS-Plus 3-6-96...........................................................................................................121

C-30 ABACOS-Plus 11-15-01.......................................................................................................122

C-31 ABACOS-Plus 6-14-01.........................................................................................................123

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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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EXAMPLES OF BIOASSAY DATA RECORDS AND REPORTS

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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: 238U, 

235U, 234U). 

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Figure C-10. Rocky Flats Environmental Technology Site (1) 8-8-96 (analyte: 239Pu). 

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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. 

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